ML17329A588
| ML17329A588 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 07/29/1992 |
| From: | Marsh L Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17329A589 | List: |
| References | |
| NUDOCS 9208190275 | |
| Download: ML17329A588 (21) | |
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t UNITEDSTATES NUCLEAR REGULATORY COMMISSION WASHINGTOhl, D. C. 20555 INDIANA MICHIGAN POWER COMPANY DOCKE NO. 50-315 DONALD C.
COOK NUCLEAR PLANT UNIT NO.
1 NENDNENT TO ACI TY OP RAT NG IC NS Amendment No.
166 License No.
DPR-58 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Indiana Michigan Power Company (the licensee) dated March 27, 1992 as supplemented by letters dated April 21, Hay 21, and July 29,
- 1992, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),
and the Commission's rules and regulations set forth in 10 CFR Chapter I;
B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
9208i90275 920729 PDR" ADOCN 05000312 P
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2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No.
DPR-58 is hereby amended to read as follows:
Technical S ecifications The Technical Specifications contained in Appendices A and B,
as revised through Amendment No.
- 166, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION puggy~
Ledyard B. Marsh, Director Project Directorate III-1 Division of Reactor Projects III/IV/V Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
July 29, 1992
ATTACHMENT TO LICENSE AMENDMENT NO.
166 TO FACILITY OPERAT'ING LICENSE NO.
DPR-58 DOCKET NO. 50-315 Revise Appendix A Technical Specifications by removing the pages identified below and inserting the attached pages.
The revised pages are identified by amendment number and contain marginal lines indicating the area of change.
REMOVE vi 3/4 4-7 3/4 4-8 3/4 4-9 3/4 4-10 3/4 4-11 3/4 4-12 3/4 4-13 3/4 4-14 3/4 4-15 3/4 4-16 B 3/4 4-2a B 3/4 4-4 B 3/4 4-5 INSERT vl 3/4 4-7 3/4 4-8 3/4 4-9 3/4 4-10 3/4 4-11 3/4 4-12 3/4 4-13 3/4 4-14 3/4 4-15 3/4 4-16 B 3/4 4-2a B 3/4 4-2b B 3/4 4-4 B 3/4 4-5
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& S L
Hot Standby.
Shutdovn.
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Reactor Coolant Loops..
SECTION g)g.y
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3 4.4 REACTOR COOLANT SYS (Continued) 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION (Continued)
PAGE 3/4 4-2 3/4 4-3 3/4 4-3b 3/4.4.2 3/4.4.3 3/4.4.4 3/4.4.5 3/4.4.6 SAFETY VALVES -
SHUTDOWN SAFETY VALVES - OPERATING.
PRESSURIZER STEAM GENERATORS REACTOR COOLANT SYSTEM ~GE Leakage Detection Systems.
3/4 4-4 3/4 4-5 3/4 4-6 3/4 4-7 3/4 4-15 3/4 4-16 3/4.4.7 3/4.4.8 3/4.4.9 CHEMISTRY..
SPECIFIC ACTIVITY.....
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PRESSURE/TEMPERATURE LIMITS Reactor Coolant System..
Pressurizer....................
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Overpressure Protection Systems...............,.......
3/4.4.10 STRUCTURAL INTEGRITY 3/4.4.11 RELIEF VALVES - OPERATING 3/4.4.12 REACTOR COOLANT VENT SYSTEM Reactor Vessel Head Vents Pressurizer Steam Space Vents................,...,
3/4 4-18 3/4 4-21 3/4 4-25 3/4 4-30 3/4 4-31 3/4 4-33 3/4 4-35 3/4 4-37 3/4 4-39 COOK NUCLEAR PLANT - UNIT 1 VI AMENDMENT NO. M, 444,166
ST LIMITING CONDITION FO OPERATION 3.4.5 Each steam generator shall be OPERABLE.
APPLICABILITY:
MODES 1, 2, 3 and 4.*
ACTION:
With one or more steam generators inoperable, restore the inoperable 0
generator(s) to OPERABLE status prior to increasing T,s above 200 F.
SURVEILLANCE RE UIREMENTS 4.4.5.0 Each steam generator shall be demonstrated OPERABLE by performance of the following augmented inservice inspection program and the requirement of Specification 4.0.5.
4.4.5.1 Steam Generator Sam le Se ~ction and Ins ectio
- Each steam generator shall be 'determined OPERABLE during shutdown by selecting and inspecting at least the miqimum number of steam generators specified in Table 4.4-1.
4.4.5.2 Steam Generator Tube Sam le Selection and s ectio
- The steam generator tube minimum sample si.ze, inspection result classification, and the corresponding action required shell be as specif'ed in Table 4.4-2.
The inservice inspection of steam generator tubes shall be performed at the frequencies specified in Specification 4.4.5.3 and the inspected tubes shall be verified acceptable per the acceptance criteria of Specification 4.4.5.4.
The tubes selected for each inservice inspection shall include at least 3% of the total number of tubes in all steam generators; the tubes selected for these inspections shall be selected on a random basis except:
a.
Where experience in similar plants with similar water chemistry
'ndicates critical areas to be inspected, then at least 50% of the tubes inspected shall be from these critical areas.
b.
The first sample of tubes selected for each inservice inspection (subsequent to the preservice inspection) of each steam generator shall include:
1, All tubes that previously had detectable wall penetrations (greater than or equal to 20%) that have not been plugged or repaired by sleeving in the affected area.
~is Specification does not apply in Mode 4 while performing crevice flushing as long as Limiting Conditions for Operation for Specification 3.4.1.3 are maintained.
COOK NUCLEAR PLANT UNIT 1 3/4 4-7 AHENDHEMT No. 4Ak4><<
U S
2.
Tubes in those areas where experience has indicated potential problems'.
A tube inspection (pursuant to Specification 4.4.5.4.a.8) shall be performed on each selected tube. If any selected tube does not permit the passage of the eddy current probe for a tube inspection, this shall be recorded and an adjacent tube shall be selected and subjected to a tube inspection.
c.
The tubes selected as the second and third samples (if required by Table 4.4-2) during each inservice inspection may be subjected to a partial tube inspection provided:
1.
The tubes selected for the samples include the tubes from those areas of the tube sheet array where tubes with imperfections were previously found.
2.
The inspections include those portions of the tubes where imperfections were previously found, d.
Implementation of the steam generator tube/tube support plate interim plugging criteria for one fuel cycle (Cycle 13) requires a
100t bobbin coil inspection for hot leg tube support plate intersections and cold leg intersections down to the lowest cold leg tube support plate with known outer diameter stress corrosion cracking (ODSCC) indications.
The results of each sample inspection shall be classified into one of the following three categories:
Cate~o Ins ect on Results C-1 Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes are defective.
C>>2 One or more tubes, but not more than 1% of the total tubes inspected are defective, or between 5% and 10%
of the total tubes inspected are degraded tubes.
C-3 More than 10% of the total tubes inspected are degraded tubes or more than 1% of the inspected tubes are defective.
COOK NUCLEAR PLANT - UNIT 1 3/4 4-8 AMENDMENT NO.
OQ, ~166
C RCO L
C RE U REMENTS Co Note:
In all inspections, previously degraded tubes must exhibit significant (greater than or equal to 10%) further wall penetrations to be included in the above percentage calculations.
4.4.5.3 Ins ection Fre ue cies
- The above required inservice inspections of steam generator tubes shall be performed at the following frequencies:
a.
The first inservice inspection shall be performed after 6 Effective Full Power Months but within 24 calendar months of initial criticality.
Subsequent inservice inspections shall be performed at intervals of not less than 12 nor more than 24 calendar months after the previous inspection.
If two consecutive inspections following service under AVT conditions, not including the preservice inspection, result in all inspection results falling into the C-l category or iz two consecutive inspections demonstrate that previously observed degradation has not continued and no additional degradation has occurred, the inspection interval may be extended to a maximum of once per 40 months.
b.
If the results of inservice inspection of a steam generator conducted in accordance with Table 4.4-2 at 40 month intervals fall in Category C-3, the inspection frequency shall be increased to at least once per 20 months.
The increase in inspection frequency shall apply until the subsequent inspections satisfy the criteria of Specification 4.4.5,3.a; the interval may then be extended to a maximum of once per 40 months.
Additional, unscheduled inservice inspections shall be performed on each steam generator in accordance with the first sample inspection specified in Table 4.4-2 during the shutdown subsequent to any of the following conditions:
1.
Primary-to-secondary tubes leaks (not including leaks originating from tube-to-tube sheet welds) in excess of the limits of Specification 3.4.6.2.
2.
A seismic occurrence greater than the Operating Basis Earthquake.
3.
A loss-of-coolant accident requiring actuation of the engineered safeguards.
4.
A main steam line or feedwater line break.
d.
'ubes left in service as a result of application of the tube support plate interim plugging criteria shall be inspected by bobbin coil probe during all future refueling outages.
COOK NUCLEAR PLANT - UNIT 1 3/4 4-9 AMENDMENT NO. 08,166
C OR CO SYS UR ILLANC t
u d 4 '.5.4 cce tance Criteria a.
As used in this Specification:
contour of a tube or sleeve from that required by fabrication drawings or specifications.
Eddy-current testing indications below 20t of the nominal wall thickness, if detectable, may be considered as imperfections.
2.
general corrosion occurring on either inside or outside of a tube or sleeve.
De raded Tube or Sleeve means an imperfection greater than or equal to 20'4 of the nominal wall thickness caused by degradation.
4, Percent De radatio means the amount of the wall thickness affected or removed by degradation.
5.
Defect means an imperfection of such severity that it exceeds the repair limit.
6.
Re air Plu in Limit means the i;..perfection depth at or beyond which the tube or sleeved tube shall be repaired or removed from service.
Any tube which, upon inspection, exhibits tube wall degradation of 40 percent or more of the nominal tube wall thickness shall be plugged or repaired prior to returning the steam generator to service.
Any sleeve which, upon inspection, exhibits wall degradation of 29 percent or more of the nominal wall thickness shall be plugged prior to returning the steam generator to service, In addition, any sleeve exhibiting any measurable wall loss in sleeve expansion transition or weld zones shall be plugged.
This definition does not apply for tubes experiencing outer diameter stress corrosion cracking confirmed by bobbin probe inspection to be within the thickness of the tube support plates.
See 4.4.5.4.a.l0 for the plugging limit for use within the thickness of the tube support plate.
7.
Unserviceable describes the condition of a tube or sleev'e if it leaks or contains a defect large enough to affect its structural integrity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as specified in 4.4.5.3,c, above.
8.
or sleeve from the point of entry (hot leg side) completely COOK NUCLEAR PLANT - UNIT 1 3/4 4-10 hMENDMENT NO. 44, ~,gag
around the U-bend to the top support of the cold leg.
For a tube in which the tube support plate elevation interim plugging limit has been applied, the inspection will include all the hot leg intersections and all cold leg intersections down to, at least, the level of the last crack indication.
9.
Qee~vi g a tube is permitted only in areas where the sleeve spans the tubesheet area and whose lower joint is at the primary fluid tubesheet face.
10'e Su o t ate te u
te is used for disposition of a steam generator tube for continued service that is experiencing outer diameter initiated stress corrosion cracking confined within the thickness of the tube support plates.
For application of the tube support plate interim plugging limit, the tube's disposition for continued service will be based upon standard bobbin probe signal amplitude.
The plant-specific guidelines used for all inspections shall be amended as appropriate to accommodate the additional information needed to evaluate tube support plate signals with respect to the above voltage/depth parameters.
Pending incorporation of the voltage verification requirement in ASME standard verifications, an ASME standard calibrated against the laboratory standard will be utilized in the Donald C.
Cook Nuclear Plant Unit 1 steam generator inspections for consistent voltage normalization.
1.
A tube can remain in service if the signal amplitude of a crack indication is less than or equal to 1.0 volt, regardless of the depth of tube wall penetration, if, as a result, the projected end-of-cycle distribution of crack indications is verified to result in primary-to-secondary leakage less than 1 gpm in the faulted loop during a postulated steam line break event.
The methodology for calculating expected leak rates from the projected crack distribution must be consistent with VCAP-13187, Rev. 0.
2.
A tube should be plugged or repaired if the signal amplitude of the crack indication is greater than 1.0 volt except as noted in 4.4.5.4.a.10.3 below.
3.
A tube can remain in service with a bobbin coil signal amplitude greater than 1.0 volt but less than or equal to 4.0 volts if a rotating pancake probe inspection does not detect degradation.
Indications of degradation with a bobbin coil signal amplitude greater than 4.0 volts will be plugged or repaired.
COOK NUCLEAR PLANT - UNIT 1 3/4 4-11 AMENDMENT NO. OS, k%4,166
b.
The steam generator shall be determined OPERASLZ after completing the corresponding actions (plugging or sleeving all tubes exceeding the repair limit and all tubes containing through-eall cracks) required by Table 4.4-2.
C ~
Steam generator tube repairs may be mad>> in accordance vith the methods described in either WCAP-12623 or CEN-313-P.
4.4. 5. 5
$L~~~
as Following each inservice inspection of steam generator tubes, if there are any tubes requiring plugging or sleevtng, the number of tubes plugged or sleeved in each steam generator shall be reported to the Commission vithin 15 days.
b.
The complete results of the steam generator tube inservice inspection shall be included in the Annual Operating Report for the period in vhich this inspection was completed.
This report shall include:
1.
Number and extent of tubes inspected.
2.
Location and percent of vali-thickness penetration for each indication of an imperfection.
3.
Identification of tubes plugged or sleeved.
c
~
Results of steam generator tube inspections vhich fall into Category C-3 and require prompt notification of the Commission shall be reported pursuant to Specification 6.9.1 prior to resumption of plant operation.
The vritten follow of this report shall provide a description of investigations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.
d.
The results of inspections performed under 4.4.5.2 for all tubes in vhich the tube support plate interim plugging criteria ""s been applied shall be reported to the Commission vithin 15
.sys
'ollcnring the inspection.
The report shall include:
1.
Listing of applicable tubes.
e.
2.
Location (applicable intersections per tube) and extent of degradation (voltage).
The results of steam line break leakage analysis performed under T/S 4.4.5.4.a.10 vill be reported to the Commission prior to restart for Cycle 13.
COOK NUCLEAR PIAHT - VNZT 1 3/4 4-12 AMENDMENT NO'. )IHIP,166
EACTOR COO hNT YSTEHS TABLE 4.4-]
MINIMUM WUHBER OF STEAM GENERATORS TO BE rwspECTEn nURIwc INSERvicE iwst ECTrow Preservice Ins ction No. of Steam Generators per Unit First Inservice Ins ction Second
& Subse ent Inservice Inspections Yes Four One able Notat one 1.
The inservice inspection may be limited to one steam generator on a rotating schedule encompassing 3
Ni of the tubes (where N is the number of steam generators in the plant) if the results of the first or previous inspections indicate that all steam generators are performing in a like manner.
Note that under some circumstances, the operating conditions in one or more steam generators may be found to be more severe than those in other steam generators.
Under such circumstances the sample sequence shall be modified to inspect the most severe conditions.
2.
The third and fourth steam generators not inspected during the first inssrvice inspection shall hs inspected during the second and third inspections, respectively.
The fourth and subsequent inspections shall follow the instructions described in 1 above.
COOK NUCLEAR PLhNT - UNIT 1 3/4 4-13 AMENDMENT NO.gtII~fg),166
TkSLE 4.4-2 STEkH GATOR TOIL rHSPECT Sample Size k
minimum of S Tubes per S.C.
Result kction Required C-1 H/k kction Required H/k Action Required H/h C-2 C-3 Plug or slee~a defective tubes and inspect additional 2S tubes in this S.C.
Inspect all tubes in this S.G., plug or sleeve defective
- tubes, and inspect 2S tubes in each other S.C.
C-1 C-2 C-3 All other S.G.s are C-l Hone Plug or sleeve defecCive Cubes and inspect additional 4S tubes in this S.C.
Perform action for C-3 result of first sample Nc".ae C-1 C-2 C-3 N/h N/A None Plug or sleeve defective tubes Perform action for C-3 result of first sample N/A V/A Prompt noti-fication to HRC pursuant Co pacification
.9.1 Some S.C.s C-2 but no additional S.C. are C-3 Additional S.G.
i@
C-3 Perform action for C-2 result of second sample Inspect all tubes in each S.C.
and plug or sleeve defective tubes.
Prompt notific-ation to NRC pursuant to specification 6.9.1 N/A N/A e num er o steaa generators t e un t, an n
ere
~ num er oz unamamr HO.EN,LPR.>66 steaa generators inspected during 'an inspection "OCR NUCLEAR HJH? - UIIT 1 3/4 4-14
J
0 0
3.4.6.1 The folloving Reactor Coolant System leakage detection systems shall be OPERABLE:
a.
One of the containment atmosphere particulate radioactivity monitoring channels (ERS-1301 or ERS-1401),
b.
The containment sump level and flov monitoring system, and c.
Either the containment humidity monitor or one of the containment at=osphere gaseous radioactivity monitoring channels (ERS-1305 or ERS-1405).
APPLICABILITY:
MODES l, 2, 3 and 4.
~CTION:
Vith only tvo of the above required leakage detection systems
- OPERABLE, operation may continue
=or up to 30 days provided grab samples of the containment atmosphere are obtained and analyzed at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> vhen the required gaseous and/or particulate radioactivity monitoring channels are inoperable; othervise, be in at least HOT STANDBY vithin the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN vithin thc fol'ovtng 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURV IL NC U
M.
S 4.4.6.1 The leakage detection systems shall be demonstrated OPERABLE by:
Containment atmosphere particulate and gaseous (if being used) moni.toring system-performance of CHANNEL CHECK, CHANNEL CALZBRATZON and CHANNEL FUNCTIONAL TEST at the frequencics specified in Table 4.3-3, b.
Containmcnt sump level and flov monitoring system-performance of CHANNEL CALZBRATZON at least once per ld months, c ~
Containment humidity monitor (if being used) - performance of CHANNEL CALZBRATZON at least once per 18 months.
COOK NUCLEAR PLANT - UNZT 1 3/4 4-15 AKHDlKNTNO. )$$,166
OR OPERA 3.4.6.2 Reactor Coolant System leakage shall be limited to:
a.
No PRESSURE BOUNDARY LEAKAGE, b.
1 GPM UNIDENTIFIED LEAKAGE, c.
600 gallons per day total primary-to-secondary leakage through all steam generators and 150 gallons per day through any one steam generator for Fuel Cycle 13, d.
10 GPM IDENTIFIED LEDGE from the Reactor Coolant System, e.
Seal line resistance greater than or equal to 2.27E-l ft/gpm2 and, f.
1 GPM leakage from any reactor coolant system pressure isolation valve specified in Table 3.4-0.
APPLICABILITY:
MODES 1, 2,
3 and 4.~
ACTION:
a.
With any PRESSLRE BOt:NDARY LEAKAGE, be 'n at least HOT STANDBY within 6 hou s and in COLD SHUTDOWN w'=hin the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
b.
With any Reactor Coolant System leakag greater than any one of the above limits, excluding PRESSURE BOUNL'>3Y LEAKAGE, reduce the leakage rate ro within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
Ce With any reactor coolant system pressure isolation valve(s) leakage greater than the above limit, except when:
1.
The leakage is less tha". or equal to 5.0 gpm, and 2.
The most recent measured leakage does not exceed the previous measured leakage* by an amount that reduces the To satisfy ~QUA requirements, measured 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, Specification 3.4.6.2.e is applicable with average pressure within 20 psi of the nominal full pressure value.
COOK NUCLEAR PLANT - UNIT 1 3/4 4-16 AMENDMENT NO. ~,],66 Order dated April 20, 1981
The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS vi11 be maintained.
The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision l.
Inservt.ce 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.
Znservice inspection of steam generator tubing also provides a means of'haracterizing 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 second-ary coolant vill be maintained'ithin those chemistry limits found to result in negligible corrosion of the steam generator tubes.
Zf the secondary coolant chemistry is not maintained vithin these parameter limits, localized corrosion may likely result in stress corrosion cracking.
The extent of cracking during plant operation vould be limited by the limitation of steam generator tube leakage betveen the primary coolant system and the secondary coolant system.
The allovable primary-to-secondary leak rate is 150 gallons per day per steam generator for one fuel cycle (Cycle 13).
Axial or circumferentially oriented cracks having a primary-to-secondary leakage less than this limit during operation vill have an adequate margin of safety to vithstand the loads imposed during normal operation and by postulated accidents.
Leakage in excess of this limit will require plant shutdom and an inspection, during vhich the leaking tubes vill be located and plugged or repaired.
A steam generator vhile undergoing crevice flushing in Mode 4 is available for decay heat removal and is operable/operating upon reinstatement of auxiliary or main feed flov control and steam control.
wastage-type defects are unlikely vith the all volatile treatment (AVT) of secondary coolant.
- Hovever, even if a defect of similar type should develop in service, it vill be found during scheduled inservice steam generator tube examinations.
Plugging or sleevtng vi11 be required for all tubes vith imperfections exceeding the repair limit vhich is defined in Specification 4.4.5.4.a.
Steam generator tube inspections of opera
'.ng plants have demonstrated the capability to reliably detect degradatxon that has penetrated 20% of the original, tube vali thickness.
Tubes experiencing outer diameter stress corrosion cracking within the thickness of the tube support plates are plugged or repaired by the criteria of 4.4.5.4.a.10.
COOK NUCLEAR PLANT -
VNXT 1 8 3/4 4-2a AMENT NO. kN 454,166
4 4 G
(Continued) whenever the results of any steam generator tubing inservice fnspectfon fall into Category C-3, these results rill be promptly reported to the Commission pursuant to Specificatfon 6.9.1 prior to resumption of plant operatfon.
Such cases sfll be considered by the Commfssfon on a case-by-case basfs and may result in a requirement for analysis, laboratory examinations,
- tests, additfonal eddy-current inspectfon, and revfsion of the Technfcal Specfficatfons, ff necessary.
COOK NUCLEhR PMT - UNIT 1 B 3/4 4-2b mmmm'O.gee
Maintaining an operating leakage limit of 150 gpd per steam generator (600 gpd total) for Puel Cycle 13 will minimise the potential for a large leakage event during steam line break under LOCh, conditions.
Based on the NDE uncertainties, bobbin coil voltage distribution and crack growth rate from the previous inspection, the expected leak rate following a steam line rupture is limited to below 1 gpm.
This is less than the 120 gpm used to calculate the offsite doses within 10 percent of 10 CPR 100 guidelines.
Leakage in the intact loops is limited to 150 gpd. If the projected end of cycle distribution of crack indications results in primary-to-secondary leakage greater than 1 gpm in the faulted loop during a postulated steam line break event, additional tubes must be removed from service in order to reduce the postulated primary-to-secondary steam line break leakage to below 1 gpm.
PRESSURE BOUNDhRY LFAKhGE of any magnitude is unacceptable since it may be indicative of an impending gross failure of the pressure boundary.
Should PRESSURE BOUNDhRY LEhKhGE occur through a component which can be isolated from the balance of the Reactor Coolant System, plant operation may continue provided the leaking component is promptly isolated from the Reactor Coolant System since isolation removes the source of potential failure.
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 LOCh.
Leakage from the RCS Pressure Isolation Valves is IDENTIPIED L1BKhGE and will be considered as a
portion of the allowed limit.
3 4.4,7 CHEMIS The limitations on Reactor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is minimised and reduces the potential for Reactor Coolant System leakage or failure due to stress corrosion.
Haintaining 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 State Limits, 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.
COOK NUCLEhR PLhHT - UNIT 1 B 3/4 4-4 uamuMENT NO. m,166
The surveillance requirements pzovide adequate assurance that concentrations in excess of the limits vill be detected in sufficient time to take corrective action.
4.4.8 SPECIFIC ACTIV 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 boundazy villnot exceed an appropriately small fraction of Part 100 limits folloving a steam generator tube rupture accident in con]unction vith an assumed steady state primary-to-secondary steam generator leakage rate of 1.0 GPH.
The values for the limits on specific activity represent interim limits based upon a parametric evaluation by the NRC of typical site locations.
These values are conservative in. that specific site parameters of the Cook Nuclear Plant
- site, such as site boundary location and meteorological conditions, vere not considered in this evaluation.
The NRC is finalising site specific criteria which vill be used as the basis for the reevaluation of the specific activity limits of this site.
This reevaluation may result in higher limits.
Offsite doses folloving a main steam line break aze limited to 10 percent of the 10 CFR 100 guideline.
The restriction is based on a Cook Nuclear Plant site-specific radiological evaluation that assumes a post-accident primary-to-secondary leak rate of 120 gpm in the faulted loop and a primary coolant specific activity concentration corresponding to 1\\ fuel defects (approximately 4.6 microCuries/gram dose equivalent Z-131), rather than a specific activity of 1.0 microCuries dose equivalent I-131.
Reducing T,s to less than 500 F prevents the release of activity should a
steam generator tube rupture since the saturation pressure of the primary coolant is belov the lift pressure of the atmospheric steam relief valves.
The surveillance requirements provide adequate assurance that excessive specific activity levels in the primary coolant vill be detected in sufficient time to take corrective action.
Infozmation obtained on iodine spiking vill be used to assess the parameters associated vtth spiking phenomena.
h reduction in fzequency of isotopic analyses follcnring pover changes may be permissible if justified by the data obtained.
COOK NUCLEAR PLEX - UNIT 1 B 3/4 4-5