ML20211K779
| ML20211K779 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 06/24/1986 |
| From: | DUKE POWER CO. |
| To: | |
| Shared Package | |
| ML19292F492 | List: |
| References | |
| TAC-61775, TAC-61776, NUDOCS 8606300257 | |
| Download: ML20211K779 (9) | |
Text
m REA_CTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)
All n'nplugged tubes that previously had detectable wall 1) o penetrations (greater than 20%),
2)
Tubes in those areas where experience has indicated potential I
problems
- and 3)
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.
The tubes selected as the second and third samples (if required by c.
Table 4.4-2) during each inservice inspection may be subjected to a partial tube inspection provided:
1)
The tubes selected for these samples include the tubes from '
those areas of the tube sheet array where tubes with imperfections were previou!ily found, and 2)
The inspections include those portions of the tubes where imperfections were previously found.
The results of each sample inspection shall be classified into one of the following three categories:
Category Inspection 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.
Note:
In all inspections, previously degraded tubes must exhibit significant (greater than 10%) further wall penetrations to be included in the above percentage calculations.
- Including adjacent " intact" tubes.
8606300257 860624 PDR ADOCK 05000369 P
PDR McGUIRE - UNITS 1 and 2 3/4.4-12
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SURVEILLANCE REQUIREMENTS (Continued) 4.4.5.4 Acceptance Criteria a.
As used in this specification:
1)
Imperfection means an exception to the dimensions, finish or contour of a tube from that required by fcbrication drawings or specifications.
Eddy-current testing indications below 20% of the nominal tube wall thickness, if detectable, may be considered as imperfections; 2)
Degradation means a service-inducsd cracking, wastage, wear or general corrosion occurring on either inside or outside of a tube; 3)
Degraded Tube means a tube containing imperfections greater than or equal to 20% of the nominal wall thickiiess caused by degradation; 4)
% Dearadation means the pet'centage of the tube wall thickness affected or removed by degradation; 5)
Defect means an imperfection of such severity that it exceeds the plugging limit.
A tube containing a defect is defective; 6)
Pluacing Limit means the imperfection depth at er beyond which the tube shall be removed from service and is equal to 40% of the nominal tube wall thickness; (NOTE 1) l 7)
Unserviceable describes the condition of a tube if it leaks or contains a defect large enough to affect its structural integ-rity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as specified in 4.4.5.3c, above; 8)
Tube Inspection means an inspection of the steam generator tube from the point of entry (hot leg side) completely around the U-bend to the top support of the cold leg; and 9)
Adjacent Tube is the tube one row higher and in the same column.
10)
Intact Tube is a tube that is not a degraded tube.
NOTE 1:
This definition does not apply to the area of tubes in the tubesheet region more than 2 inches below the top of the tubesheet except when the tube has a degraded tube in the adjacent position (one row higher in number, same column).
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McGUIRE - UNITS 1 and 2 3/4 4-14
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REACTOR COOLANT SYSTEM BASES 3/4.4.4 RELIEF VALVES The power-operated relief valves (PORVs) and steam bubble function to relieve RCS pressure during all design transients up to and including the design step load decrease with steam dump. Operation of the PORVs minimizes the undesirable opening of the spring-loaded pressurizer code safety valves.
Each PORV has a remotely operated block valve to provide a positive shutoff capability should a relief valve become inoperable.
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 main-tained. 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.
t 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 chemistry limits found to result in negligible corrosion of the steam generator tubes.
If the secondary coolant chemistry is not maintained within these limits, localized corrosion may likely result in stress corrosion cracking. The extent of cracking during t
plant operation would be limited by the limitation of steam generator tube leakage between the Reactor Coolant System and the Secondary Coolant System (reactor-to-secondary leakage = 500 gallons per day per steam generator).
Cracks having a reactor-to-secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the loads imposed dur.ing normal operation and by postulated accidents.
Operating plants have demonstrated that reactor-to-secondary leakage of 500 gallons per day per steam generator can readily be detected by radiation monitors of steam generator blowdown.
Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged.
Wastage-type defects are unlikely with proper chemistry treatment of the secondary coolant. However, even if a defect should develop in service, it will be found during scheduled inservice steam generator tube examinations, j
Plugging will be required for 6 tubes with imperfectionsgexceeding the l
plugging limit of 40% of the tube nominal wall thickness.
Steam generator tube inspections of operating plants have demonstrated thel capability to reliably detect degradation that has penetrated 20% of the original tube wall 4
thickness.
- in the b spain rT0 tor)
McGUIRE - UNITS 1 and 2 8 3/4 4-3 i
For tubes with imperfections found greater than two inches below the top of the tubesheet, plugging is not required except when the adjacent tube is a degraded tub'e as defined by Specificaton 4.4.5.4.
The acceptance criteria of Specification 4.4.5.4 (degraded tube) applies to the full length of the tube.
The definition of an intact tube is conservatively taken to be a tube that is not a degraded tube (no indication is greater than 20% through wall). Should future evaluations justify a value that is less conservative, the definition of intact tube may be changed. Similarly, the distance into the tubesheet from the top face of the tubesheet, the P* value, is conservatively chosen to be two inches. This includes the factors discussed in the Westinghouse Safety Evaluation report (which resulted in a P* value of 1.282 inches) with i
an additional uncertainty for eddy current testing. Further evaluations may justify a reduction in the P* value of two inches, at which time that may be changed.
Future inspections shall include at least 3% of the adjacent " intact" tubes.
If any adjacent tube is found to be degraded in any part of the tube, then the tube below it shall be plugged.
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1 McGuire - Units 1 and 2 B 3/4 4-3a
REACTOR COOLANT SYSTEM BASES-STEAM GENERATORS (Continued)
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 10 CFR Sections 50.72 and 50.73 prior to resumption of
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plant operation.
Such cases will be considered by the Commission on a case-by-case basis and may result in a requirement for analysis, laboratory exami-nations, tests, additional addy-current inspection, and revision of the Technical Specifications, if necessary.
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 l
provided to monitor and detect leakage from the reactor coolant pressure boundary.
These Detection Systems are consistent with the recommendations of Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection 4
l Systems," May 1973.
3/4.4.6.2 OPERATIONAL 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 8
to a threshold value of less than 1 gpm.
This threshold value is sufficiently low to ensure early detection of additional leakage.
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 I
the allowed limit.
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.
4 The CONTROLLED LEAKAGE limitation restricts operation when the total flow supplied to the reactor coolant pump seals exceeds 40 gpm with the modulating valve in the supply line fully open at a nominal RCS pressure of 2235 psig.
l 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 10 CFR Prrt 100 dose guideline values in the event of either a steam generator tube rupture or I
steam line break.
The 1 gpa 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 integrity is maintained in the event of a main steam line rupture or under LOCA conditions.
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Amendment No.52((Unit 1 McGUIRE - UNITS 1 and 2 B 3/4 4-4 Amendment No.33 Unit 2)
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Justification and Safety Analyses The proposed' change to the Technical Specification is to prevent the unnecessary plugging of steam generator tubes due to degradation in the tubesheet region as a result of posaible Primary Water Stress Corrosion Cracking (PWSCC). This has been identified as a potential problem in Westinghouse plants with mode D steam generators, thus McGuire is not immune to this problem.
An extensive safety analysis is contained in Attachment 3, which assures that i
the structural integrity of the steam generator is maintained.
The unnecessary plugging of steam generator tubes is an undesirable situation. The plugging of a large number of tubes would result in a reduced flow rate in the reactor coolant system, which may affect safety analysis assumptions. An ALARA consideration is also involved as a radiation dose of cpproximately 310 mRen is expected for each tube plugged.
The proposed changes take into account the necessity of assuring the ctructural integrity of adjacent tubes as these tubes, in accordance with the Westinghouse safety analysis, may be called upon to prevent pullout of the offected tube. The definition of adjacent tube is being added te the specification to provide a clear definition of adjacent tube that no question will exist.
I Duke has also chosen to propose a P* value of two inches which is greater than the Westinghouse Safety Evaluation value of 1.282 inches. Duke has chosen this value to allow for additional eddy current testing uncertainty cnd to allow for expedited review.
The bases for this specification are also modified to reflect these proposed changes to further explain the revised criteria and provide guidance for when adjacent tubes may be degraded and the basis for the criteria of a degraded j
rule.
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ATTACHMENT 3
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DAP-85-108 Westinghouse Water Reactor Ba 355 Electric Corporation Divisions Pittsburg1 Pennsylvania 15230 0355 April 25,1985 NS-OPLS-OPL-85-181 MPS #35924 Mr. K. S. Canady, Manager Nuclear Engineering Services Duke Power Company P. O. Box 33189 Charlotte, North Carolina 28242 Attention:
R. Gill McGUIRE NUCLEAR STATION UNITS NUMBER 1 AND 2 TUBESHEET PLUGGING CRITERIA SAFETY EVALUATION
Dear Mr. Canady:
Per a request of Bob Gill of Duke to Gary Whiteman of Westinghouse on April.15, 1985, find attached proprietary and non-proprietary versions of infonnation titled "McGuire Units 1 and 2 Tubesheet Region Plugging Criterion".
In addition to the above, the following enclosures are also provided for your use.
1.
Information which should be included in your NRC transmittal letter.
2.
Westinghouse letter " Application for Withholding Proprietary Infonnation from Public Disclosure (CAW-85-035).
3.
One copy of Affidavit CAW-81-79.
Please transmit the original of item (2) to the NRC in your submittal.
If there are any questions on the above or attached, please contact Roy Owoc on 412 - 374-4037.
Very truly ours, Yw L. L. Williams, Manager NSID Projects Mid South Area R. H. Owoc/pj cc: H. Purcell, IL K. S. Canady, 6L, 6A T. F. Wyke, IL, lA R. C. Futrell, IL R. O. Sharpe, IL, lA