Text

Proposed Tech Specs,Deleting &/Or Relocating Addl primary- to-secondary Leak Rate Limits & Enhanced Leakage Monitoring Requirements Imposed Following 1987 SG Tube Rupture Event
	ML20206G836
Person / Time
Site:	North Anna
Issue date:	05/03/1999
From:	; VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	;
Shared Package
ML20206G830	List: ML20206G827; ML20206G836;
References
	NUDOCS 9905100158
	Download: ML20206G836 (41)
	v • d • e

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Attachment 2 Mark-up of Technical Specifications Changes T

North Anna Power Station Units 1 and 2 Virginia Electric and Power Company 9905100158 990503 PDR ADOCK 05000338 P PDR l

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. 12-12-88 REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION -

3.4.6.2 Reactor Coolant System leakage shall be limited to: l

a. No PRESSURE BOUNDARY LEAKAGE, l

b. 1 GPM UNIDENTIFIED LEAKAGE, ,

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c. 1 GPM total primary-to-secondary leakage through all steam f generators-nct isolated 'rce the % ::ter C 01:r,t Sy: tem-and 500 gallons per day through any one steam generator, e+4-

- i; lated frc the hatter C Clar,t Systs ,^-

d. 10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System,

e. 30 CPM CONTROLLEO LEAKAGE at a Reactor Coolant System pressure of 2235 20 psig, and ,

f. Leakage for the Reactor Coolant System Pressure Isolation Valves specified in Table 3.4-1.

APPLICABILITY: MODES 1, 2, 3 and 4.

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l ACTION. '

a. With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANOSY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLO SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . l

b. With any Reactor Coolant System leakage greater than any one i

of the above limits, excluding PRESSURE BOUNDARY LEAKAGE and leakage from the Reactor Coolant System Pressure Isolation Valves, reduce the leakage rate to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ur be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTOOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months , l

c. With any Reactor Coolant System Pressure Isolation Valve leakage greater than the above limit, be in at least HOT STANOBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUT 00WN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

&nen in t de 1 e eve 507 pccer, preei:ica: cf Sp::ificatica 3.'.5.3 0FF17- -

NORTH ANNA - UNIT 1 3/4 4-17 $t/At AAt#4 /4/20/37 Amendeent No.109

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4-20-81 REACTOR C00LANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.6.2.1 Reactor Coolant System leakages shall be demonstrated to be -

l2 within,each of the above limits by:

a. Monitoring the containment atmosphere particulate radioactivity j

monitoring at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Monitoring the containment sump inventory and discharge at least l b.

once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

c. Measurement of the CONTROLLED LEAKAGE to the reactor coolant pump l seals when the Raaetor Coolant Systes pressure is 2235 2 20 psig l.

at least once per 31 days with the modulating valve fully open,

d. Performance of a Reactor Coolant System water inventory balance at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months during steady state operationT"and I e. Monitoring the reactor head flange leakoff temperature at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

4.4.6.2.2 Each Reactor Coolant System Pressure Isolation Valve specified in Table 3.4.1 shall be individually demonstrated OPERA 8LE by verifying l

leakage" to be within its limit:

a. Prior to entering 20E 2 after each refueling,

b. Prior to entering 20E 2 whenever the plant has been in COLD SHUT-DOWN for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or more and if leakage testing has not been performed in the previous 9 months, and

c. Prior to returning the valve to service following maintenance, repair or replacement work on the valve.

"To satisfy ALARA requirements, leakage may be sensured indirectly

~ (as fros the performance of pressure indicators) if accomplished in accordance with approved procedures and supported by computations showing that the method is capable of demonstrating valve compliance with the leakage criteria. q -

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i MORTH AMMA - LMIT 1 3/4 4-18

12-12-88 I

MEACTORCOOLANTSYSTEM P ARY TO SECONDARY LEAKAGE LIMI G CONDITION FOR OPERATION /

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3.4.6.3 imary to secondary leakage shall be limited to:

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a. Tota leakage from all steam generators of 300 gpd,

b. Leakage rom an individual steam generator of 100 gpd,

c. Total leak e increase of 60 gpd between surveillance i ervals, and

d. An increasing rend based on the latest surveillance hat indicates 100 gpd would n t be exceeded on an individual ste generator within

) 90 minutes.

i APPLICABILITY _: MODE I above 0% power.*

ACTION: l

a. If the total leakage limit from all st m generators or the leakage linit from any individual st am gener tor is exceeded, be in HOT I STANOBY within the next 6 hou and cold shutdown within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months ,

b. If the increase in total leaka f m all steam generators exceeds j 60 gpd between surveillance terval , reduce power below S0% rated j j

thermal power within 90 mi tes, 4

c. If an increasing trend dicates that the limit of 100 gpd per steam generator is going to e exceeded within 9 minutes, reduce power to below 50% rated ther al power within 90 minu es, be in HOT STAN08Y t within the next 6 urs and cold shutdown wit n the following l 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . l

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0nce the limiting con tion for operation has been exceede the corresponding l action must be folio ed to completion. l SURVEILLANCE REQU EMENTS \

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Prim y to secondary leakage shall be demonstrated to be wi in each 4.4.6.3 .

of he above limits by:

a. rimary to secondary leakage will be recorded and trended at lea t <

once during each 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months interval (e.g., 00:00-04:00, 04:00-08:00, 08:00-12:00, 12:00-16:00, 16:00-20:00, 20:00-24:00) from each OPERAB r l

N-16 continuous readout and alarm radiation monitoring system and the condenser air ejector exhaust continuous readout and alarm radiation monitor. l I

3/4 4-18b Amendment No. 109 NORTH ANNA - UNIT I l

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12-12-88 litAS.TQRCOOLANTSYSTEM SURVEIL IREMENTS

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b. Primary to secon leakage will be determine om a condenser air ejector grab sample a st every 24 ho .

c. Primary to secondary leaka determined from steam generator and reactor coolant samples at t every 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

d. If the surveillance operations cannot be pe d as specified, miting conditions for operation and associated ac statements ,

of Specification 3.4.6.4 shall apply. l l

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l 3/4 4-18e Amendment No. 109 NORTH ANNA - UNIT 1

1 12-12-88 EACTOR COOLANT SYSTEM PR RY TO SECONDARY LEAKAGE DETECTION SYSTEMS LIMIT CONDITION FOR OPERATION '

1 3.4.6.4 following primary to secondary leakage detection systems sh I be OP ABLE:

a. One o the two N-16 radiation monitoring systems (either th N-16 continu s readout and alarm radiation monitors on each eam line, or the N- 6 continuous readout and alarm radiation moni r on the main steam eader),

b. The condenser ir ejector exhaust continuous readou and alarm radiation monit r,

c. The capability to btain and analyze a condens air ejector exhaust l grab sample, and l

-d. The capability to obta and analyze a li id sample from each steam l

generator and from the R S. ;

APPLICABILITY: MODE 1 above 50% poa r.

ACTION:

a. If both the N-16 radiation moni ing system on each steam line and the N-16 radiation monitoring yste on the main steam header are INOPERABLE, increase the fre Jency o the condenser air ejector grab sample required by Specific tion 4.4. 3b to at least once during each 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months interval (e.g , 00:00-04:00, 04:00-08:00, 08:00-12:00, 12:00-16:00, 16:00-20:00 20:00-24:00) an return at least one of the systems to operation w hin seven days or duce power to less than 50% within the next f ur hours.

b. If the condenser a'r ejector exhaust continuous eadout and alarm radiation monitor is IN0PERABLE, provided at leas one of the N-16 l monitoring syst as is OPERABLE, increase the frequ cy of the condenser air jector grab sample required by Speci ' cation 4.4.6.3b to at least nce during each 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months interval (e.g., 0 00-04:00, 04:00-08:0 , 08:00-12:00, 12:00-16:00, 16:00-20:00, 20: 0-24:00) and retur the system to operation within seven days or duce power to less han 50% within the next four hours.

c. If t capability to obtain and analyze a condenser air eject r grab sam e is lost, provided at least one of the N-16 monitoring s tems i OPERABLE and the condenser air ejector exhaust continuous rea ut nd alarm radiation monitor is OPERABLE, restore the capability wi hin seven days or reduce power to less than 50% within four hours, s

MORTH ANNA - UNIT I 3/4 4-18d Amendment No. 109

-o . 12-12-88

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REACTOR COOLANT SYSTEM LIM M CONDITION FOR OPERATION /

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d. If b h N-16 monitoring systems are INOPERABLE and either e t conden air ejector exhaust continuous readout and al radiation l monitor i INOPERABLE or the capability to obtain an analyze a condenser at ejector exhaust grab sample is lost reduce power to less than 50% ' hin the next 90 minutes.

e. if the condenser air ector exhaust con nuous readout and alarm I radiation monitor is I ERABLE and t capability to obtain and analyze a condenser air e tor ex ust grab sample is lost, reduce power to less than 50% withi next 90 minutes.

f. If the capability to obta and an ze a liquid sample from each ,

steam generator and th CS is lost, crease the frequency of j performance of the water inventory b ante in T.S. 4.4.6.2.ld to once every 24 ours.

SURVEILLANCE REQUIRC NTS x

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4.4.6.4 Th .-16 monitors and air ejector exhaust radiation mon ring tion channels shall be demonstrated OPERABLE by the per mance of i instrume the l the C

NEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST dur' l M00 and at the frequencies shown in Tables 4.4-2a and 4.3-14, respective. .

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l 3/4 4-18e /cendment No.109 s NORTH ANNA - UNIT 1

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12-12-88 3/4.4 REACTOR COOLANT SYSTEM

- BASES 3/4.4.5 STEAM GENERATORS ;

The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be maintained. The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1. 3 Inservice inspection of steam generator tubing is essential in order to f maintain surveillance of the conditions of the tubes in the event that j there is evidence of mechanical damage or progressive degradation due to !

design, manufacturing errors, or inservice conditions that lead to corrosion.

Inservice inspection of steam generator tubing also provides a means of "

characterizing the nature and cause of any tube degradation so that correc-tive measures can be taken. ;

1 The plant is expected to be operated in a manner such that the secondary f coolant will be maintained within those parameter limits found to result in negligible corrosion of the steam generator tubes. If the secondary coolant chemistry is not maintained within these parameter limits, localized corro-sion may likely result in stress corrosion cracking. The extent of cracking during plant operation would be limited by the limitation of steam generator tube leakage between the primary coolant system and the secondary coolant j system (primary-to-secondary leakage = 500 gallons per day per steam ;

generator). Cracks having a primary-to-secondary leakage less than this i limit during operation will have an adequate margin of safety to withstand "

the loads imposed during nonnal operation and by postulated accidents. j Operating plants have demonstrated that primary-to-secondary leakage of 500 gallons per day per steam generator can readily be detected by radiation i monitors of steam generator blowdown. i

\ It has been determined, however, that certain conditions within the steam M ator may produce limited displacement fluideiastic instability in the hat may result in fatigue fai. lure of a tube. Modifications h een e

bun accompi d in all steam generators consisting of installation o owncomer ;

resistance p bs and r eventive plugging of potentially susc tble tuber.

Even thougn thes ieasures are consider ^ed to have been.v conservative and i highlf effective in ducing the probabilirty 6f fati induced tube rupture, l enhanced leakage monitoH and more stringent 1 rate limits have been established. Leakage is no imited'to 100 (rather than 500 gpd) per steam generator when operating at gre tgr the % poyer. CycTic.lfferanalysis of fatigue induced tube cracks has sh hat, assuming a post-modification maximum stress amplitude of 7 ksi, a le rate up to 500 gpd would be reached some 90 minutes prior to tube r re. Therefo the'100 gpd leak rate limit is bounding since a, the .gpd limit would be detected.well in ance of reachiiig gpd, '

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b. the time required for leak rate' detection and power red ion to less than 50% s expected to be less than 90 minutes, and 1

NORTH ANNA - UNIT 1 B 3/4 4-3 Amendment No. 32, 88 109 !

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12-12-88 REACTOR COOLANT SYSTEM BASES I N .

the maximum stress amplitude is anticipated to lie in the 5 ksi ra .

! w would allow for much earlier leak before break warning l would e r in the assumed 7 ksi case, l j

These assumptions als clude an appropriate all ce for measurement uncertainty. (

References:

Vir a Electric an wer Co., " North Anna Unit 1 July 15,1987 Steam Generator Tube R re Report, Revision 1, September 15, 1987, and Westinghouse WCAP-il601, "No Unit 1 Steam Generator Tube Rupture and Remedial Actions Tec al Evaluat September 1987").

This limit, alon the enhanced monitoring syste , ould provide sufficient notifi on to permit orderly shutdown prior to a p tial tube l rupture eve .

Leakage in excess of any of these limits will requir nt l l shutdo nd an unscheduled inspection, during which the leaking tubes w i b cated and plugged. ;

Wastage-type defects are unlikely with the all volatile treatment (AVT) of secondary coolant. However, even if a defect of similar type should develop in service, it will be found during scheduled inservice steam generator tube examinations. Plugging will be required of all tubes with imperfections l

exceeding the plugging limit which, by the definition of Specification 4.4.5.4.a is 40% of the tube nominal wall thickness. Steam generator tube inspections of l operating plants have demonstrated the capability to reliably detect degradation that has penetrated 20% of the original tube wall thickness.

Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these results will be promptly reported to the l Commission pursuant to Section 50.72 to 10 CFR Part 50 with a follow up I

report pursuant to Section 50.73 to 10 CFR Part 50. Such cases will be considered by the Commission on a case-by-case basis and may result in a requirement for analysis, laboratory examinations, tests, additional eddy-current inspection, and revision of the Technical Specifications, if necessary.

l NORTH ANNA - UNIT 1 B 3/4 4-3a Amendment No.109

6 05-22-97 REACTOR COOLANT SYSTEM BASES 3/4.4,6 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.6.1 LEAKAGE DETECTION SYSTEMS The RCS leakage detection systems required by this specification are provided to monitor and detect leakage from the Reactor Coolant Pressure Boundary during normal plant operations and after seismic events to provide prompt and quantitative information to the operators to permit immediate corrective actions should a Reactor Coolant Pressure Boundary leak be detrimental to the safety of the facility.

These detection systems are generally consistent with the recommendations of Regulatory Guide 1.45," Reactor Coolant Pressure Boundary Leakage Detection Systems," May 1973. The containment atmospheric particulate and gaseous radioactivity monitoring system is not fully seismically qualified. Consistent with RG 1.45 these monitors can perform their intended function during normal plant operations. To ensure ;h: safety function of detecting reactor coolant pressure boundary leakage is maintained after a seismic event the operability of these monitors is required to be verified immediately following a seismic event or the affected units will be shut down and cooled down to COLD SHUTDOWN.

3/4.4.6.2 OPER ATIONAL LEAKAGE Industry experience has shown that while a limited amount of leakage is expected from the RCS, the unidentified portion of this leakage can be reduced to a threshold value ofless than 1 GPM. This threshold value is sufficiently low to ensure early detection of additional leakage.

The 10 GPM IDENTIFIED LEAKAGE limitation provides allowance for a limited amount of leakage from known sources whose presence will not interfere with the detection of UNIDENTIFIED LEAKAGE by the leakage detection systems.

The CONTROLLED LEAKAGE limitation restricts operation when the total flow supplied to the reactor coolant pump seals exceeds 30 GPM with the modulating valve in the supply line fully open at a nominal RCS pressure of 2235 psig. This limitation ensures that in the event of a LOCA, the safety injection flow will not be less than assumed in the accident analyses.

The total steam generator tube leakage limit of 1 GPM for all steam generators not isolated from the RCS ensures that the dosage contribution from the tube leakage will be limited to a small fraction of Part 100 limits in the event of either a steam generator tube rupture or steam line break. The 1 GPM limit is consistent with the assumptions used in the analysis of these ,

accidents.4 genera!, for p! ant cperatien at er be!c"> 50% pc's/er,Ihe 500 gpd leakage limit per steam generator ensures that steam generator tube integrity is maintained in the event of a main steam line rupture or under LOCA conditions.

NORTH ANNA - UNIT 1 B 3/4 4-4 Amendment No. 409, REVISED BY NRC LETTER DATED 5/22/97

12-12-88 REACTOR COOLANT SYSTEM l BASES en operating at greater than 50% power, more stringent primary to secondar limits of 300 ga er day (GPD) total from all three steam generators and i d from an individual steam generato e event that a fatigue ve been imposed. These limits ensurejtha induced crack were to occur in on more generators, the re thig leak would be detected in l sufficient time to conduct an orderly shut rior t astrophic tube failure. The limits on an increase in leakage of 60 gpd between surveil ei als and for an increasing trend indicating that 100 gpd would be exceeded wit minutes ensure t - 'n the event of fatigue crack initiation, power can be redu o a level below which propagation not occur. In the latter case, the limit also es for orderly shutdown since the 100 gpd limit is approached.

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These leak ates are conservative with regard to dosage contribution in that they a ss than the nously analyzed total amount of 1 GPM and 500 GPD for any single steam generator.

x PRESSURE BOUNDARY LEAKAGE of any magnitude is unacceptable since it may be indicative of an impending gross failure of the pressure boundary. Therefore, the presence of any PRESSURE BOUNDARY LEAKAGE requires the unit to be promptly placed in COLD SHUTDOWN.

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f NORTH ANNA - UNIT I B 3/4 44a Amendment No.109

i Unit 2 a

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y 12-12-88

-REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE j

-LIMITING CONDITION FOR OPERATION ;

i 3.4.6.2 Reactor Coolant System leakage shall be limited to: l l

a. No PRESSURE BOUNDARY LEAKAGE, ' f

b. 1 GPM UNIDENTIFIED LEAKAGE, ;

c. 1 GPM total primary-to-secondary leakage through aT1 steam generators net f:01sted 're" the Peeete- Ceelant Sy:ter and 500 gallons per day i through any one steam generator n:t i::1:t:d fr; th; neactor Cocient g l Sy;ter,'* I i

d. 10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System,

e. 30 GPM CONTROLLED LEAKAGE at a Reactor Coolant System pressure of 4 2235 1 20 psig, and.

f. Leakage for the Reactor Coolant System Pressure Isolation Valves specified in Table 3.4-1.*

APPLICASILITY: MODES 1, 2, 3 and 4.

ACTION:

a. With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANOBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTOOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months ,

b. With any Reactor Coolant System leakage greater than any one of the.

above limits, excluding PRESSURE BOUNDARY LEAKAGE and leakage from the Reactor Coolant System Pressure Isolation Valves, reduce the leakage rate to kithin limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUT 00WN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months ,

c. With any Reactor Coolant System Pressure Isolation Valve leakage greater than the above limit,- be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTOOWN within the folloking 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

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The leakage-limit for any RHR system isolation valve shown in Table 3.4-1 j shall be 5 GPM.

' ",en in Mode 1 ebeve 50t pe.,er, previsiens of Specification 2.5.0.3 epply, l

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Amendment No. 95 j NORTH ANNA - UN:T 2 '3/4 4-17 1

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11-17-90 1

l REACTOR COOLANT SYSTEM -

SURVE!LLANCE REQUIREMENTS 4.4.6.2.1 Reactor Coolant System leakages shall be demonstrated to be within each of the above limits by:

a. Monitoring the containment atmosphere particulate radioactivity monitor at/least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ,

b. Monitoring the containment sump inventory and discharge at least l once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

c. Measurement of the CONTROLLED LEAKAGE to the reactor coolant pump seals when the Reactor Coolant System pressure is 2235 + 20 psi at least once per 31 days with the modulating valve fulIy open.s The provisions of Specification 4.0.4 are not applicable for entry into MODE 4. j

d. Perfomance of a Reactor Coolant System water inventory balance at least once per 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months sy sm % m any sem oevaama

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e. Mcnitoring the reactor heac flange leakoff temperature at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . -

4.4.6.2.2 Each Reactor Coolant System Pressure !sclation Valve specified in Table 3.4-1 shall be demonstrated OPERABLE pursuar.: to Specification 4.0.5, except that in lieu of any leakage testing requirec by Specification 4.C.5, each valve shall be demonstrated 00ERABLE by verifying leakage to be within its limit:

a. At least once per 18 months,

b. Prior to entering MODE 2 whenever the plant has been in COLO SHUTOCWS for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or more and if leakage testing has not been perfome in the previous 9 months.

c. Prior to returning the valve to service following maintenance, repair or replacement work on the valve.

d. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following valve actuation due to automatic or manual action or. flow through the valve.

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7 NORTH ANNA - UNIT 2 3/4 4-18 /. e. 6e .: S: . 2. ,

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l 12-12-88 1

I kEACTORCOOLANTSYSTEM PRIMARY TO SECONDARY LEAKAGE Il, l LIMIT CONDITION FOR OPERATION f J.4.6.3 P mary to secondary leakage shall be limited to:

a. Tota leakage from all steam generators of 300 gpd, I

b. Leakage rom an individual steam generator of 100 gpd,

c. Total leaka e increase of 60 gpd between surveillance intervals, and

d. Anincreasing\rendbasedonthelatestsurveilla e that indicates 100 gpd would n: he exceeded on an individual s eam generator within 90 minutes.

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I APPLICABILITY: MODE 1 above % power.*

ACTION: ,

a. If the total leakage limit rom all s eam generators or the leakage limit from any individual st m gen rator is exceeded, be in HOT cold shutdown within"the following

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STANDBY within the next 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months a 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . ,

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b. Iftheincreaseintotalleapgef m all steam generators exceeds 60 god between surveillance interval reduce power below 50% rated thermal power within 90 m utes,

c. If an increasing trend ndicates that the limit of 100 gpd per steam generator is g ng to be exceeded wi in 90 minutes, reduce power to below 50% ated thermal power withi 90 minutes, be in HOT STANDBY withi the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and cold s utdown within the following 30 hou s.

0nce the limiting co ition for operation has been exceede the corresponding action must be foll ed to completion.

SURVEILLANCE REQU EMENTS \

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4.4.6.3 Pri ry to secondary leakage shall be demonstrated to be wi in each of he above limits by:

a. rimary to secondary leakage will be recorded and trended at lea t once during each 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months interval (e.g., 00:00-04:00, 04:00-08:00,-

08:00-12:00, 12:00-16:00, 16:00-20:00, 20:00-24:00) from each OPERABLE N-16 continuous readout and alarm radiation monitoring system and the condenser air ejector exhaust continuous readout and alam radiation monitor. ,

liORTH ANNA - UNii 2 3/4 4-18b Amendment No. 95

12-12-88 R COOLANT SYSTEM SURVE!LL IREMENTS

~

/

b. Primary to secon leakage will be determin ejector grab sampie at rom a condenser air ry 24 b .

c. Primary to secondary leaka determined from steam generator and reactor coolant I samples at t every 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

d. If the ab surveillance operations cannot be pe sp med as ed, the limiting conditions for operation and a iated ction statements of Specification 3.4.6.4 shall apply.

NORTH ANN' . UNIT 2 3/4 4-18c Amendment Mc. 95

1 12-12-88 (ACTOR COOLANT SYSTEM PR .)RY TO SECONDARY LEAKAGE DETECTION SYSTEMS LIMIT Q CONDITION FOR OPERATION

/

3.4.6.4 following primary to secondary leakage detection systems shall be opt BLE: /

a. One o the two N-16 radiation monitoring systems (either he N-16 continu s readout and alarm radiation monitors on eac steam line, or the M- 6 continuous readout and alarm radiation mo itor on the main steam eader),

b. The condenser ir ejector exhaust continuous rea cut and alarm radiation monit ,

c. The capability to tain and analyze a cond nser air ejector exhaust grab sample, and

d. The capability to obtal and analyze a iquid sample from each ,

steam generator and from the RCS.

APPLICABILITY: MODE I above 50% pow .

ACTION: )

a. If both the N-16 radiation ionito 'ng system on each steam line and the N-16 radiation mo 1toring s tem on the main steam header are INOPERABLE, increas the frequen of the condenser air ejector grab sample re utred by Specif ation 4.4.6.3b to at least once during each 4 h r interval (e.g., 0:00-04:00, 04:00-08:00, 08:00-12:00, 12:00- 6:00, 16:00-20:00, 20. 0-24:00) and return at least one of t e systems to operation w hin seven days or reduce power to ess than 50% within the nex four hours,

b. If the conde er air ejector exhaust continuous eadout and alarm radiation nitor is INOPERABLE, provided at leas one of the N-16 monitorin systems is OPERABLE, increase the frequ cy of the condens air ejector grab sample required by Speci cation 4.4.6.3b to at east once during each 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months interval (e.g., 0 00-04:00, 04:0 -08:00, 08:00-12:00, 12:00-16:00, 16:00-20:00, 20 (:0 -24:00) an return the system to operation within seven days or duce power t less than 50% within the next four hours.

c. If the capability to obtain and analyze a condenser air eject grab sample is lost, provided at least one of the N-16 monitoring sys ens is 0PERABLE and the condenser air ejector exhaust continuous read t and alarm radiation monitor is OPERABLE, restore the capability within seven days or reduce power to less than 50% within four hours.

i l

NORTil ANNA - UNIT 2 3/4 4-18d Amendment No. 95

12-12-88 REACTOR COOL ANT SYSTEM LIM CONDITION FOR OPERATION

x

d. If oth N-16 monitoring systems are INOPERABLE and either airejeh ondenser exhaust continuous readout and alarm radia dn monitor is l INOPERABL the capability to obtain and analyze ondenser air (jector exh grab sample is lost, reduce powe c less than 50% ;

within the next minutes.

e. If the condenser air ector exhaust co nuous readout and alarm radiation monitor is INO RABLE and e capability to obtain and analyze a condenser air eje or e aust grab sample is lost, reduce power to less than 50% within e next 90 minutes.

f. If the capability to obt n and ana e a liquid sample from each steam generator and t RCS is lost, i ease the frequency of performance of the 5 water inventory ba nce in T.S. 4.4.6.2.1d to once every 2 ours.

SURVEILLANCE RE0'JI .ENTS s

4.4.6.4 T N-16 monitors and air ejector exhaust radiation moni ing '

instrum ation channels shall be demonstrated OPERABLE by the perfo nce of the nNNEL CHECX, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST d0rin the M tS and at the frequencies shown in Tables 4.4-2a and 4.3-13 respective 1 1

NORTH ANNA - UNIT 2 3f4 4-I8e Amendment Nc. 95

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12-12-88 3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator i tubes be ensure that the structural integrity of this portion of the RCS will maintained.

The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1.

Inservice inspection of steam generator tubing is essential in order to maintain surveillance of the conditions of the tubes in the event tha there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead _to corrosion.

Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that correc-tive measures can be taken.

coolant will be maintained within those parameter lim negligible corrosion of the steam generator tubes. If the secondary coolant chemistry is not maintained within these parameter 1.imits, localized corro-sion may likely result in stress corrosion cracking.

during plant operation would be limited by the limitation of steam generat tube leakage between the primary coolant system and the secondary coolant system generator).(primary-to-secondary leakage = 500 gallons per day per steam Cracks having a primary-to-secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents.

Operating plants have demonstrated that primary-to-secondary leakage of 500 gallons per day per steam generator can readily be detected by-radiation monitors of steam generator blowdown.

\ .

NsteIt has been determined, however, that certain conditions within the I the tub enerator may produce limited displacement fluidelastic instabili in have been a undle that may result in fatigue failure of a tube. Modi ations downcomer resis mplished in all steam generators consisting of i lation of tubes. Even thoug ce plates and preventive plugging of pote ally susceptible tive and highly effect ese measures are considered to h been very conserva-in reducing the probabili of fatigue induced tube I rupture, enhanced leakage toring and more s been established. Leakage is n limited t ngent Teak rate limits have l

hteam generator when operating at r t 0 gpd (rather than 500 gpd) per i analysis of fatigue induced tube er than 50% power. Cyclic life modification maximum stress amp s shown that, assuming a post-ude of 1, a leak rate of up to 500 gpd ;

would be reached some 90 mi es prior to tube spd leak rate limit is tu re'. Therefore, the 100 nding'since

a. the 10 pd limit would be detected well in advan I 50 pd, * ' of reaching l j

' the time required for leak rate detection and power reduction less than 50% is expected to be less than 90 minutes, and

' {

1 NORTH ANNA - UNIT C B 3/4 4-3 Amendment No. 4 , 95

l l

l 12-12-88 REACTOR COOLANT SYSTEM )

8ASES )

" the maximum stress amplitude is anticipated to lie in the 5 ksi nge

, h would allow for much earlier leak before break warn han would uccur - tne assumed 7 ksi case.

These assumptions a include an appropriate wance for measurement uncertainty. (

References:

(V1TRnia Electric ower Co., " North Anna Unit 1 July 15, 1987 Steam Generator Tube % tur ent Report, Revision 1 September 15, 1987, and Westinghouse WCAP-11601, " na Unit 1 Steam Generator Tube Rupture and Remedial Actions T cal Evalua , September 1987").

This limit, alo th the enhanced monitoring sys should provide sufficient noti tion to permit orderly shutdown prior to tential tube rupture ev . Leakage in excess of any of these limits will requ' lant shut and an unscheduled inspection, during which the leaking tubes be l ated and plugged. ]

l Wastage-type defects are unlikely with all volatile treatment (AVT) of secondary' coolant. However, even if a defect.of similar type should develop 4n service, it will be found during scheduled inservice steam generator tube examinations. Plugging will be required of all tubes with imperfections exceeding the plugging limit which, by the definition of Specification 4.4.5.4.a is 40% of the tube nominal, wall thickness. Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect degradatici that has penetrated 20% of the original tube wall thickness. )

Whenever the results of any steam generator tubing inservice inspection oli into Category C-3, these results will be promptly reported to the Comission pursuant to Section 50.72 to 10 CFR Part 50 with a follow up report puesuant to Section 50.73 to 10 CFR Part 50. Such cases will be considered by the Commission on a case-by-case basis and may result in a requirement for analysis, laboratory examinations, tests, additional eddy-current inspection, and revision of the Technical Specifications, if necessary.

NORTH ANNA - UNIT 2 S 3/4 4-3a Amendment No. 95

. 05-22-97 REACTOR COOLANT SYSTEM BASES 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE j 1

3/4.4.6.1 LEAKAGE DETECTION SYSTEMS '

The RCS leakage detection systems required by this specification are provided to monitor and detect leakage from the Reactor Coolant Pressure Boundary during normal plant operations and after seismic events to provide prompt and quantitative information to the operators to permit immediate corrective actions should a Reactor Coolant Pressure Boundary leak be tigrimental to the safety of the facility.

These detection systems are generally consistent with the recommendations of Regulatory Guide 1.45," Reactor Coolant Pressure Boundary Leakage Detection Systems," May !

1973. The containment atmospheric particulate and gaseous radioactivity monitoring system is not fully seismically qualified. Consistent with RG 1.45 these monitors can perform their intended function during normal plant operations. To ensure the safety function of detecting reactor coolant pressure boundary leakage is maintained after a seismic event the operability of these monitors is required to be verified immediately following a seismic event or the affected units will be shut down and cooled down to COLD SHUTDOWN.

3/4.4.6.2 OPERATIONAL LEAKAGE

. Industry experience has shown that while a limited amount ofleakage is expected from the RCS, the unidentified portion of this leakage can be reduced to a threshold value ofless than 1 I GPM. This threshold value is sufficiently low to ensure early detection of additional leakage.

The 10 GPM IDENTIFIED LE AKAGE limitation provides allowance for a limited amount of leakage from known sources whose presence will not interfere with the detection of UNIDENTIFIED LEAKAGE by the leakage detection systems. l The surveillance requirements for RCS Pressure Isolation Valves provide added assurance of valve integrity thereby reducing the probability of gross valve failure and consequent )

intersystem LOCA. Leakage from the RCS Pressure Isolation Valves is IDENTIFIED LEAKAGE and will be considered as a portion of the allowed limit.

The CONTROLLED LEAKAGE limitation restricts operation when the total flow supplied to the reactor coolant pump seals exceeds 30 GP.M with the modulating valve in the supply line fully open at a nominal RCS pressure of 2235 psig. This limitation ensures that in the event of a LOCA, the safety injection flow will not be less than assumed in the accident analyses.

The total steam ge.nerator tube leakage limit of 1 GPM for all steam generators not isolated

_ from the RCS ensures that the dosage contribution from the tube leakage will be limited to a small fraction of Part 100 limits in the event of either a steam generator tube rupture or steam line break.

The 1 GPM limit is consistent with the assumptions,used in the analysis of these accidents.--Ie--

g; accel, for pbn: cp;r::ic: :: :r 5:!: >' 50% p :::r,)he 500 gpd leakage limit per steam generator ensures that steam generator tube integrity is maintained in the event of a main steam line rupture or under LOCA conditions.

NORTH ANNA - UNIT 2 B 3/4 4-4 Amendment No. 95, REVISED BY NRC LETTER DATED 5/22/97

12-12-88 REACTOR COOLANT SYSTEM BASES

\ erating at greater than 50% power, more stringent primary to second eakage limits of 300 gallo gr day (GPD) total from all three steam generators an gpd from an .

individual steam generator 7rav been imposed.These limits ensure n the event that a fatigue ,

induced crack were to occur in one ore generators, the ing leak would be detected in '

sufficient time to conduct an orderly shut o rio atastropnic tube failure. The limits on an increase in leakage of 60 gpd between surv ' ce i s and for an increasing trend indicating that 100 gpd would be exceeded w' hr 0 minutes ensure t ,' the event of fatigue crack initiation, power can be re to a level below which propagation s ot occur. In the latter case, the limit also ides for orderly shutdown since the 100 gpd limit is a proached.

These lea - rates are conservative with regard to dosage contribution in that they ar s than ti eviously analyzed total a' mount of 1 GPM and 500 GPD for any single steam generator.

PRESSURE BOUNDARY LEAKAGE of any magnitude is unacceptable since it may be indicative of an impending gross failure of the pressure boundary.Therefore, the presence of any PRESSURE BOUNDARY LEAKAGE requires the unit to be promptly placed in COLD 1

SHUTDOWN.

l l

1 l

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l NORTH ANNA - UNIT 2 B 3/4 4-4a Amendment No. 95

Attachment 3 Proposed Technical Specifications Changes i

l l

North Anna Power Station

- Units 1 and 2 Virginia Electric and Power Company

9 1

1 l

l l

Unit 1 i 1

1

REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System leakage shall be limited to: l l

a. No PRESSURE BOUNDARY LEAKAGE,

b. I GPM UNIDENTIFIED LEAKAGE,

c. 1 GPM total primary-to-secondary leakage through all steam generators and 500 gallons per day through any one steam generator, d.10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System, I

e. 30 GPM CONTROLLED LEAKAGE at a Reactor Coolant System pressure of 2235 20 psig, and l )

f. Leakage for the Reactor Coolant System Pressure Isolation Valves specified in Table 3.4- 1.

APPLICABILITY: MODES 1,2,3 and 4.

ACTION:

a. With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE and leakage from the Reactor

) Coolant System Pressure Isolation Valves, reduce the leakage rate to within limits t

within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in a at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

c. With any Reactor Coolant System Pressure Isolation Valve leakage greater than the above limit, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

NORTH' ANNA - UNIT 1 3/4 4-17 Ordct dated 4/20/87 Amendment No. 409, l-

l REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.6.2.1 Reactor Coolant System leakages shall be demonstrated to be within each of the above limits by:

a. Monitoring the containment atmosphere particulate radioactivity monitoring at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ,

b. Monitoring the containment sump inventory and discharge at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

c. Measurement of the CONTROLLED LEAKAGE to the reactor coolant pump seals when the Reactor Coolant System pressure is 2235 i 20 psig at least once per 31 days with the modulating valve fully open,

d. Performance of a Reactor Coolant System water inventory balance at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months during steady state operation **, and l )

i

e. Monitoring the reactor head flange leakoff temperature at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

4.4.6.2.2 Each Reactor Coolant System Pressure Isolation Valve specified in Table 3.4.1 shall be individually demonstrated OPERABLE by verifying leakage

to be within its limit:

a. Prior to entering MODE 2 after each refueling,

b. Prior to entering MODE 2 whenever the plant has been in COLD SHUTDOWN for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or more and ifleakage testing has not been performed in the previous 9 months, and

c. Prior to returning the valve to service following maintenance, repair or replacement work on the valve.

To satisfy ALARA requirements, leakage may be measured indirectly (as from the performance of pressure indicators) if accomplished in accordance with approved procedures and supported by computations showing that the method is capable of demonstrating valve compliance with the leakage criteria.

- Primary-to-secondary leakage not required below 50% power. l NORTH ANNA - UNIT 1 3/4 4-18 Amendment No.

i 3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be maintained. The program for inservice ;

inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision

1. Inservice inspection of steam generator tubing is essential in order to maintain surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion.

Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken.

The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those parameter limits found to result in negligible corrosion of the steam generator tubes. If the secondary coolant chemistry is not maintained within these parameter limits, localized corrosion may likely result in stress corrosion cracking. The extent of cracking during plant operation would be limited by the limitation of steam generator tube leakage between the primary coolant system and the secondary coolant system (primary-to-secondary leakage = 500 gallons per day per steam generator). Cracks having a primary-to-secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents. Operating plants have demonstrated that j primary-to-secondary leakage of 500 ' gallons per day per steam generator can readily be detected I by radiation monitors of steam generator blowdown.

1 Wastage-type defects are unlikely with all volatile treatment (AVT) of secondary coolant.

However, even if a defect of similar type should develop in service, it will be found during scheduled inservice steam generator tube examinations. Plugging will be required of all tubes with imperfections exceeding the plugging limit which, by the definition of Specification 4.4.5.4.a is 40% of the tube nominal wall thickness. Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect degradation that has penetrated 20% of the original tube wall thickness.

Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these resuhs will be promptly reported to the Commission pursuant to Section 50.72 to 10 CFR Part 50 with a follow up report pursuant to Section 50.73 to 10 CFR Part 50. Such cases will be considered by the Commission on a case-by-case basis and may result in a requirement for analysis, laboratory examinations, tests, additional eddy-current inspection, and revision of the Technical Specifications, if necessary. .

l i

NORTH ANNA - UNIT 1 B 3/4 4-3 Amendment No. 32,63,109,

l REACTOR COOLANT SYSTEM B ASES -

1 i

3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.6.1 LEAKAGE DETECTION SYSTEMS The RCS leakage detection systems required by this specification are provided to monitor and detect leakage from the Reactor Coolant Pressute Boundary during normal plant operations l

and after seismic events to provide prompt and quantitative information to the operators to permit l immediate corrective actions should a Reactor Coolant Pressure Boundary leak be detrimental to the safety of the facility.

These detection systems are generally consistent with the recommendations of Regulatory Guide 1.45,." Reactor Coolant Pressure Boundary Leakage Detection Systems," May 1973.The containment atmospheric particulate and gaseous radioactivity monitoring system is not i fully seismically qualified. Consistent with RG 1.45 these monitors can perform their intended function during normal plant operations. To ensure the safety function of detecting reactor coolant pressure boundary leakage is maintained after a seismic event the operability of these monitors is required to be verified immediately following a seismic event or the affected units will be shut i down and cooled down to COLD SHUTDOWN.

i 3/4.4.6.2. OPERATIONAL LEAKAGE I 4

l Industry experience has shown that while a limited amount ofleakage is expected from the RCS, the unidentified portion of this leakage can be reduced to a threshold value ofless than i GPM. This threshold value is sufficiently low to ensure early detection of additional leakage.

l The 10 GPM IDENTIFIED LEAKAGE limitatien provides allowance for a limited amount ofleakage from known sources whose presence will not interfere with the detection of UNIDENTIFIED LEAKAGE by the leakage detection systems.

The CONTROLLED LEAKAGE limitation restricts operation when the total flow supplied to the reactor coolant pump seals exceeds 30 GPM with the modulating valve in the l supply line fully open at a nominal RCS pressure of 2235 psig. This limitation ensures that in the event of a LOCA, the safety injection flow will not be less than assumed in the accident analyses.

The total steam generator tube leakage limit of 1 GPM for all steam generators not l isolated from the RCS ensures that the dosage contribution from the tube leakage will be limited l

to a small fraction of Part 100 limits in the event of either a steam generator tube rupture or steam line break. The 1 GPM limit is consistent with the assumptions used in the analysis of these accidents. The 500 gpd leakage limit per steam generator ensures that steam generator tube l integrity is maintained in the event of a main steam line rupture or under LOCA conditions.

NORTH ANNA - UNIT I B 3/4 4-4 Amendment No. M9; Lctici Dated 5/22!97, l

REACTOR COOLANT SYSTEM BASES PRESSURE BOUNDARY LEAKAGE of any magnitude is unacceptable since it may be indicative of an impending gross failure of the pressure boundary.Therefore, the presence of any PRESSURE BOUNDARY LEAKAGE requires the unit to be promptly placed in COLD l

SHUTDOWN.

I 4

NORTH ANNA - UNIT I B 3/4 44a Amendment No. 409,

Unit 2

REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION

! l l 3.4.6.2 Reactor Coolant System leakage shall be limited to: !

l

a. No PRESSURE BOUNDARY LEAKAGE, I b. I GPM UNIDENTIFIED LEAKAGE, I c.1 GPM total primary-to-secondary leakage through all steam generators and !

500 gallons per day through any one steam generator, i d.10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System,

e. 30 GPM CONTROLLED LEAKAGE at a Reactor Coolant System pressure of 2235 i 120 psig, and j

f. Leakage for the Reactor Coolant System Pressure Isolation Valves specified in Table 3.4- 1.*

APPLICABILITY: MODES 1,2,3 and 4. I ACTION:

a. With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANDBY ,

within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

I

b. With any Reactor Coolant System leakage greater than any one of the above limits, l excluding PRESSURE BOUNDARY LEAKAGE and leakage from the Reactor l Coolant System Pressure Isolation Valves, reduce the leakage rate to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT STANDB Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD '

SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

c. With any Reactor Coolant System Pressure Isolation Valve leakage greater than the above limit, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

The leakage limit for any RHR system isolation valve shown in Table 3.4-1 shall be 5GPM.

I

' NORTH ANNA - UNIT 2 3/4 4-17 Amendment No. 95,

E l l a s .-

7 REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.6.2.1 Reactor Coolant System leakages shall be demonstrated to be within each of the above limits by:

a. Monitoring the containment atmosphere particulate radioactivity monitor at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b. Monitoring the containment sump inventory and discharge at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

c. Measurement of the CONTROLLED LEAKAGE to the reactor coolant pump seals i when the Reactor Coolant System pressure is 2235 t 20 psig at least once per 31 ,

days with the modulating valve fully open.The provisions of Specification 4.0.4 are not applicable for entry into MODE 4.

d. Performance of a Reactor Coolant System water inventory balance at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months during steady state operation *. l 1

e. Monitoring the reactor head flange leakoff temperature at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

l 4.4.6.2.2 Each Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1 shall be demonstrated OPERABLE pursuant to Specification 4.0.5, except that in lieu of any leakage testing required by Specification 4.0.5, each valve shall be demonstrated OPERABLE by verifying leakage to be within its limit:

a. At least once per 18 months.

b. Prior to entering MODE 2 whenever the plant has been in COLD SHUTDOWN for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or more and if leakage testing has not been performed in the previous 9 months.

c. Prior to returning the valve to service following maintenance, repair or replacement work on the valve.

d. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following valve actuation due to automatic or manual action or flow through the valve.

Primary-to-secondary leakage not required below 50% power. l NORTH ANNA - UNIT 2 3/4 4-18 Amendment No. +,

.s

1 3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be maintained. The program for inservice inspection of steam generatortubes is based on a modification of Regulatory Guide 1.83, Revision

1. Inservice inspection of steam generator tubing is essential in order to maintain surveillance of 3

the conditions of the tubes in the event that there is evidence of mechanical damage or progressive i degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion.

Inservice inspection of steam generator tubing also provides a means of characterizing the nature !

and cause of any tube degradation so that corrective measures can be taken. I The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those parameter limits found to result in negligible corrosion of the steam generator tubes. If the secondary coolant chemistry is not maintained within these parameter limits, ;

localized corrosion may likely result in stress corrosion cracking. The extent of cracking during plant operation would be limited by the limitation of steam generator tube leakage between the primary coolant system and the secondary coolant system (primary-to-secondary leakage = 500 gallons per day per steam generator). Cracks having a primary-to-secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents. Operating plants have demonstrated that primary-to-secondary leakage of 500 gallons per day per steam generator can readily be detected by radiation monitors of steam generator blowdown.

Wastage-type defects are unlikely with all volatile treatment (AVT) of secondary coolant.

He.vever, even if a defect of similar type should develop in service, it will be found during scheduled inservice steam generator tube examinations. Plugging will be required of all tubes with imperfections exceeding the plugging limit which, by the definition of Specification 4.4.5.4.a is 40% of the tube nominal wall thickness. Steam generator tube inspections of operating plants have

- demonstrated the capability to reliably detect degradation that has penetrated 20% of the original tube wall thickness.

Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these results will be promptly reported to the Commission pursuant to Section 50.72 to 10 CFR Part 50 with a follow up report pursuant to Section 50.73 to 10 CFR Part 50. Such cases will be considered by the Commission on a case-by-case basis and may result in a requirement for analysis, laboratory examinations, tests, additional eddy-current inspection, and revision of the Technical Specifications,if necessary.

NORTH ANNA - UNIT 2 B 3/4 4-3 Amendment No. 4'h95,

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l REACTOR COOLANT SYSTEM BASES l I

l 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE l 3/4.4.6.1 LEAKAGE DETECTION SYSTEMS The RCS leakage detection systems required by this specification are provided to monitor l and detect leakage from the Reactor Coolant Pressure Boundary during normal plant operations

) and after seismic events to provide prompt and quantitative information to the operators to pennit immediate corrective actions should a Reactor Coolant Pressure Boundary leak be detrimental to !

the safety of the facility.

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These detection systems are generally consistent with the recommendations of Regulatory Guide 1.45," Reactor Coolant Pressure Boundary Leakage Detection Systems," May 1973. The containment atmospheric particulate and gaseous radioactivity monitoring system is not fully l seismically qualified. Consistent with RG 1.45 these monitors can perform theirintended function I

during normal plant operations. To ensure the safety function of detecting reactor coolant pressure boundary leakage is maintained after a seismic event the operability of these monitors is required i l to be verified immediately following a seismic event or the affected units will be shut down and l l cooled down to COLD SHUTDOWN.

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1 3/4.4.6.2 OPERATIONAL LEAKAGE l

Industry experience has shown that while a limited amount of leakage is expected from the RCS, the unidentified portion of this leakage can be reduced to a threshold value ofless than i GPM. This threshold value is sufficiently low to ensure early detection of additional leakage. )

The 10 GPM IDENTIFIED LEAKAGE limitation provides allowance for a limited amount l

ofleakage from known sources whose presence will not interfere with the detection of )

UNIDENTIFIED LEAKAGE by the leakage detection systems.

The surveillance requirements for RCS Pressure Isolation Valves provide added assurance of valve integrity thereby reducing the probability of gross valve failure and consequent intersystem LOCA. Leakage from the RCS Pressure Isolation Valves is IDENTIFIED LEAKAGE and will be considered as a portion of the allowed limit.

l The CONTROLLED LEAKAGE limitation restricts operation when the total flow supplied to the reactor coolant pump seals exceeds 30 GPM with the modulating valve in the l supply line fully open at a nominal RCS pressure of 2235 psig. This limitation ensures that in the event of a LOCA, the safety injection flow will not be less than assumed in the accident analyses. ;

The total steam generator tube leakage limit of 1 GPM for all steam generators not isolated from the RCS ensures that the dosage contribution from the tube leakage will be limited to a small fraction of Part 100 limits in the event of either a steam generator tube rupture or steam line break.

The i GPM limit is consistent with the assumptions used in the analysis of these accidents. The l 500 gpd leakage limit per steam generator ensures that steam generator tube integrity is maintained in the event of a main steam line rupture or under LOCA conditions.

1 NORTH ANNA - UNIT 2 B 3/4 4-4 Amendment No. 95 LETFER DATED 5/2/97

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. . . r, REACTOR COOLANT SYSTEM

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BASES I I

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PRESSURE BOUNDARY LEAKAGE of any magnitude is unacceptable since it may be indicative of an impending gross failure of the pressure boundary. Therefore, the presence of any PRESSURE BOUNDARY LEAKAGE requires the unit to be promptly placed in COLD SHUTDOWN.

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l NORTH ANNA UNIT 2 B 3/4 4-4a Amendment No. 95,

. ,, i Attachment 4 Significant Hazards Consideration Determination North Anna Power Station Units 1 and 2 Virginia Electric and Power Company I

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SIGNIFICANT HAZARDS CONSIDERATION On July 15,.1987, North Anna Unit i experienced a steam generator tube rupture event due to fatigue caused by flow-induced vibration. As part of the corrective actions, very conservative corrective measures were implemented to reduce the probability of fatigue-induced tube rupture. These measures included the installation of downcomer flow resistance plates to reduce the source of loads associated with the fatigue mechanism in .the U-bend area and the preventive plugging of potentially susceptible tubes. Additionally, enhanced leakage monitoring instrumentation was installed and more stringent leak rates were established for operation in Mode 1 above 50% power in the unlikely . event that the downcomer modification and preventive plugging were unsuccessful in preventing occurrence of a similar fatigue failure.

Since the SGTR event, new steam generators have been installed for Unit 1 during the 1993 outage and Unit 2 during the 1995 outage with excellent performance. As a result, the more stringent leak rate limits and enhanced leakage monitoring are no longer deemed necessary. The existing Technical Specification requirements to limit Reactor Coolant System (RCS) operating leakage as specified in LCO 3.4.6.2, associated Action Statements, and Surveillance Requirements 4.4.6.2.1 and 4.4.6.2.2 are bounded by the existing accidents analyzed and will be retained.

The proposed changes retain the original Technical Specification requirements and delete the more stringent primary-to-secondary leakage limits and increased monitoring requirements of Technical specifications 3.4.6.3 and 3.4.6.4 that were imposed when operating above 50% power following the Unit 1 steam generator tube rupture event. Portions of the enhanced leakage monitoring requirements for the N-16 radiation monitoring systems will also be relocated to the Technical Requirements Manual. l Virginia Electric and Power Company has reviewed the proposed Technical Specification changes against the requirements of 10 CFR 50.92 and has j determined that the proposed changes would not pose a significant hazards consideration. Specifically, operation of the North Anna Power Station in accordance with the proposed Technical Specification changes will not:

Involve a significant increase in the probability or consequences of an accident previously evaluated.

Eliminating the conservative primary-to-secondary leakage limits associated with the replaced steam generators and the operability requirements for the leakage monitoring instrumentation does not change the operation of the plant. The steam generators will be operated, inspected, and maintained in the same manner. No new accident initiators are established as a result of Page 1 of 2

the proposed changes. Therefore, the probability of occurrence is not increased for any accident previously evaluated.

Removing the conservative primary-to-secondary leakage limits associated with the replaced steam generators and the operability requirements for the leakage monitoring instrumentation does not change the operation of the plant. Although the conservative leakage limits are being deleted, the remaining leakage limits will maintain the dose rate, in the event of a tube rupture, within the analyzed limits. Therefore, there is no increase in the consequences of any accident previously analyzed, Create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed changes do not affect the operation of the plant. The steam generators will be operated, inspected, and maintained in the same manner.

There are no modifications to the plant or steam generators as a result of the change. No new accident or event initiators are created by the removal of the conservative primary-to-secondary leakage limits associated with the replaced steam generators and the operability requirements for the leakage monitoring instrumentation. Therefore, the proposed changes do not create I the possibility of any accident or malfunction of a different type. I involve a significant reduction in the margin of safety as defined in the bases on any Technical Specifications.

The proposed changes have no effect on any safety analyses assumptions.

The remaining limits maintain primary-to-secondary leakage within the

accident analysis assumptions. The proposed changes only eliminate overly conservative primary-to-secondary leakage requirements and the operability

and surveillance requirements for the leakage monitoring system associated with the replaced steam generators. Therefore, the proposed changes do not result in a significant reduction in the margin of safety.

Therefore, there is no unreviewed safety question generated by this administrative change to the Technical Specifications.

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Page 2 of 7

ML20206G836

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