ML20010D134

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Revised Tech Specs 3.12 & 4.13 Re Shock Suppressors (Snubbers)
ML20010D134
Person / Time
Site: Prairie Island  Xcel Energy icon.png
Issue date: 08/14/1981
From:
NORTHERN STATES POWER CO.
To:
Shared Package
ML20010D128 List:
References
NUDOCS 8108240052
Download: ML20010D134 (17)
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Text

m EXHIBIT A Prairie Island Nuclear Generating Plant License Amendment Request dated August 14, 1981 Proposed Changes to the Technical Specifications Appendix A of Operating Licenses DPR-42 and DPR-60 ,

Pursuant to 10 CFR 50.59 and 50.90, the holders of Operating Lienese DPR-42 and DPR-60 hereby propose the following changes to Appendix A, Technical Specifications:

Specifications 3.12 and 4.13, Shock Suppressors (Snubbers)

  • Proposed Changes (a) Revise sections 3.12 and 4.13 of the Technical Specifications to conform to the latest NRC Staff recommendations for snubber operability and surveillance requirements. Refer to Exhibit 3 which contains revised Technical Specification pages.

(b) Revise Table TS.3.12-1 to incorporate safety-related snubbers added as a con-sequence of our IE Bulletin 79-14 analysis and inspection. In addition, two snubbers have been removed because of rerouting of a Unit II Safety Injection line to provide additional shielding.

Reason for Changes These changes are being submitted at the request of the NRC Staff. With the exception of adding requirements for mechanical snubbers, the proposed changes conform to the guidance in the enclosure to a letter dated November 20,1980 from Darrell G Eisenhut, Director, Division of Licensing, USNRC. Mechanical snubbers are not installed or planned for use at the Prairie Island Nuclear Generating Plant.

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( Safety Evaluation The snubber operability and surveillance requirements are revised to incorporate operating experience obtained since comprehensive surveillance of snubbers began several years ago. The resulting changes generally result in increased assurance that snubbers will perform as designed.

These changes will require functional testing of snubbers in high radiation areas and snubbers which are difficult to remove except where preservice functional testing or operating history has demonstrated the snubber will function with a high degree of certainty.

h8240032010814 p ADOCK 05000282 l

PDR

F EXHIBIT B License Amendment Request dated August. 14, 1981 Docket Nos. 50-282 License Nos. DPR-42 50-306 DPR-60 Exhibit B consists of revised pages for the Prairie Island Nuclear Generating Plant Technical Specifica-tions, Appendix A, as listed below showing the proposed changes:

TS-i TS-iii TS.3.12-1 Table TS.3.12-1 (page 1, 2, 3, 4, 5, 6, 7)

TS.4.13-1 TS.4.13-2 TS.4.13-3 TS.4.13-4 (new page)

TS.6.6-2

TS-i REV TECHNICAL SPECIFICATIONS TABLE OF CONTENTS TS SECTION TITLE PAGE 1.0 Definitions TS.1-1 2.0 Safety Limits and Limiting Safety System '

Settings TS.2.1-1 2.1 Safety Limit, Reactor Core TS.2.1-1 2.2 Safety Limit, Reactor Coolant System Pressure TS.2.2-1 2.3 Limiting Safety System Settings, Protective Instrumentation TS.2.3-1 3.0 Limiting conditions for Operation TS.3.1-1 3.1 Reactor Coolant System Ts.3.1-1 3.2 Chemical and Volume Control System TS.3.2-1 3.3 Engineered Safety Features TS.3.3-1 3.4 Steam and Power Conversica System TS.3.4-1 3.5 Instrumentation Seitem TS.3.5-1 3.6 Containment System TS.3.6-1 3.7 Auxiliary Electrical Systems TS.3.7-1 3.8 Refueling and Fuel Handling TS.3.8-1 3.9 Radioactive Ef fluents TS.3.9-1 3.10 Control Rod and Power Distribution Limits TS.3.10-1 3.11 Core Surveillance Instrumentation TS.3.11-1 3.12 Snubbers TS.3.12-1 3.13 Control Room Air Treatment System TS.3.13-1 3.14 Fire Detection and Protection Systems TS.3.14-1 4.0 Surveillance Requirements TS.4.1-1 4.1 Operational Safety Review TS.4.1-1 4.2 Primary System Surveillance TS.4.2-1 4.3 Reactor Coolant System Pressure Isolation Valves TS.4.3-1 4.4 Containment System Tests TS.4.4-1 4.5 Engineered Safety Features TS.4.5-1 4.6 Periodic Testing of Emergency Power System TS.4.6-1 4.7 Main Steam Stop Valves TS.4.7-1 4.8 Steam and Power Conversion Systems TS.4.8-1 4.9 Reactivity Anomalies TS.4.9-1 4.10 Radiation Environmental Monitoring Program TS.4.10-1 4.11 Radioactive Source Leakage Test TS.4.11-1 4.12 Steam Generator Tube Surveillance TS.4.12-1 4.13 Snubbers TS.4.13-1 l

, 4.14 Control Room Air Treatment System TS.4.14-1 4.15 Spent Fuel Pool Special Ventilation System TS.4.15-1 4.16 Fire Detection and Protection Systems TS.4.16-1 Y

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TS-iii REV APPENDIX A TECHNICAL SPECIFICATIONS s

LIST OF TABLES TS TABLE TITLE 3.1-1 Unit 1 Reactor Vessel Toughness Data -

3.1-2 Unit 2 Reactor Vessel Toughness Data 3.5-1 Engineered Safety Features Initiation Instrument Limiting Set Points 3.5-2 Instrument Operating Conditions for Reactor Trip 3.5-3 Instrument Operating Conditions for Emergency Cooling System 3.5-4 Instrument Operating Conditions for Isolation Functions 3.5-5 Instrument Operating conditions for ventilation Systems 3.5-6 Instrument Operating Conditions for Auxiliary Electrical System 3.9-1 Radioactive Liquid Waste Sampling and Analysis 3.9 2 Radioactive Gaseous Waste Sampling and Analysis [

3.12-1 Safety Related Snubbers I 3.14-1 Safety Related Fire Detection Instruments 3.15-1 Event Monitoring Instrumentation 4.1-1 Minimum Frequencies for Checks, Calibrations and Test of Instrument Channels 4.1-2A Minimum Frequencies for Equipment Tests 4.1-2B Minimum Frequencies for Sampling Tests 4.2-1 Special Inservice Inspection Requirements 4.4-1 Unit 1 and Unit 2 Penetration Designation for Laakage Tests 4.10-1 Prairie Island Nuclear Generating Plant-Radiation Environmental Monitoring Program Sample Collection and Analysis Environmental Monitoring Program 4.12-1 Steam Generator Tube Inspection s 5.5-1 Anticipated Annual Release of Radioactive Material in Liquid

. Effluents From Prairie Island Nuclear Generating Plant (Per Unit) 5.5-2 Anticipated Annual Release of Radioactive Nuclides in Gaseous Effluent From Prairie Island Nuclear Generating Plant (Per Unit) 6.1-1 Minimum Shift Crew Composition 6.7-1 Special Reports I

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r TS.3.12-1 REV 3.12 SNUBBERS Applicebility Applies to the operability of safety related snubbers.

Obj ect ive To define those conditions of snubber operability necessary to assure -

safe reactor operation.

Specification A. Except as permitted below, all snubbers listed in Table TS.3.12-1 shall be operable above Cold Shutdown. Snubbers may be inoperable in Cold Shutdown and Refueling Shutdown whenever the supported system is not required to be Operable.

B. With one or more snubbers made or found to be inoperable for any reason when Operability is required, within 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />s:

1. Replace or restore the inoperable snubbers to Operable status and perform an engineering evauation per Specifica-tion 4.13.E on the supported component (s), or
2. Declare the supported system inoperable and take the action required by the Technical Specifications for inoperability of that systen..

C. Snubber modifications may be made to safety related systems without prior License Amendment to Table TS.3.12-1 provided that a revision to the Table is included with the next License Amendment Request.

Basis All snubbers are required to be Operable above Cold Shutdown to ensure l that the structural integrity of the reactor coolant system and all other safety related systems is maintained during and following a seismic or other event initiating dynamic loads.

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TABLE TS.3.12-1 (Page 1 of 8)

REV SAFETY RELATED SNUBBERS Snubbers In High Accessible or Especially Radiation Snubber Inaccessib le Difficult Area During No. Location Elevation (A or I) to Remove Shutdown UNIT I .

AFSH-22 A&B Main and Avv- 773'-4-1/4" A AFSH-36 iliary Steata 745'-7-1/4" A AFSH-39 699'-10-1/4" A AFSH-48 699'-6-1/4"" A MSDH-25 A&B 736'-6-7/16" e MSDH-26 A&B 756'-7-1/4" A MSDH-29 756'-7-1/4" A MSDH-30 736-6-7/16" A

tSH-48 A&B , 739'-1-11/16 A l

MSH-62 A&B 735'-6" A MSH-68 A&B 755'-8" A UNIT II AFSH-2 Main and Auxiliary 749'-4" A AFSH-19 Steam 745'-7-1/4" A AFSH-20 745'-7-1/4" A AFSH-24 745'-6" A AFSH-29 A&B 721'-1-9/16" A AFSH-33 707'-5" A AFSH-39 696'-6-1/4" A AFSH-40 696'-6-1/4" A AFSH-44 750'-7-1/2" A AFSH-46 750'-7" A MSDH-17 739'-0" A MSDH-18 759'-0" A MSDH-19 739'-0" A MSDh-20 759'-0" A

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TABLE TS.3.12-1 (Page 2 of 8)

REV SAFETY RELATED SNUBBERS Snubbers In High Accessible or Especially Radiation Snubber Inaccessible Difficult Areas During No Location Elevation (A or I) to Remove Shutdown ,

UNIT II MSH-23 A&B Main and Auxiliary 739'-1-3/16" A l MSH-54 A&B Steam 756'-0-1/16" I MSH-81 A;B 735'-9" A MSH-82 A&J 755'-8" A MSH-83 761'-13/16"" I UNIT I RHRRH-5 Safety Injection 723'-4-1/4" 1 RHRRH-41 698'-11" I RHRRH-58 67 0'-0" A RHRRH-60 670'-0" A RP CH-160 718'-1/2" I RSIH-92 714'-11" I RSIH-93 714'-11" I RSIH-95 711'-2" I RSIH-96 711'-2" I RSIH-98 701'-2" I RSIH-163 717'-9" I RSIH-167 717'-9" I RSIH-413 A&B 722'-8" A RSIH-414 716'-10" I RSIH-442 717'-9-1/2" I RSIR-469 707'-6-1/2" I

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RSIH-469 707'-6-1/2" I kSIH-476 707'-1-3/4" I SIRH-9 737'-0" I SIRH-ll 718'-6"" I SIRH-17 730'-0" I SIRH-18 730'-0" , I i SIRH-22 711'-4" I SIRH-23 A&B 711'-4" I SIRH-26 705'-0" I l

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TABLE TS.3.12-1 (Page 3 of 8)

REV SAFETY RELATED SNUBBERS Snubbers In High Accessible or Especially Rasiation Snubber Inaccessible Difficult Areas During No. Location Elevation (A or I) to Remove Shutdown UNIT II RHRH-13 Safety Injection 673'-9" A RHRH-14 674'-0" A RHRH-52 670'-6" A RHRH-54 670'-6" A RHRRH-19 700'-11" I RHRRH-23 711'-2" I RHRRH-28 707'-4" I RSIH-265 699'-9" I RSIH-268 713'-9-3/16" I RSIH-343 719'-8-11/16" I RSIH-349 703'-l1" I

. RSIH-350 703'-11" I RSIH-353 A&B 701'-9" I SIM-53 710'-3" A SIRH-4A 711'-6-1/8" I SIRH-4B 711'-3" I SIRH-7 716'-3-1/16" I SIRH-18 722'-6" I l

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TABLE TS.3.12-1 (Page 4 of 8)

REV SAFETY RELATED SHOCK SUPPRESSORS (SNUBBERS)

Snubbers In High Accessible or Especially Radiation Snubber Inaccessible Difficult Areas During No. Location Elevation (A or I) to Remove Shutdown UNIT I -

RCRH-5 A&B Reactor Coolant 732'-6" I RCRH-12 A&B 720'-7" I RCRH-26 762'-8" I RCRH-27 A&B 761'-7" I RCRH-34 764'-7" I RCRH-45 765'-1" I RERH-46 765'-1" I RCRH-47 745'-10" I RHRRH-15 705'-6" I RHRRH-27 705'-6" I RHRRH-29 A&B 705'-6" I UNIT II RCRH-5 Reactor Coolant 731'-6" I RCRH-8 717'-6" I RCRH-9 712'-0" I RCRH-14 705'-9" I RCRH-25 732'-2" I RCRH-26 757'-7" I RCRH-31 764'-1" I RCRH-45 724'-6" I RCRH-46 758'-3" I RCRH-47 760'-3" I RCRH-48 765'-1" I RCRH-49 765'-1" I RRCH-279 A&B 724'-9" I RRCH-282 723'-2" I P2CH-284 A&B 725'-8" I RHRRH-2 699'-0". I RHRRH-4 705'-11" I RHRRH-9 705'-11" I RHRRH-15 699'-0" I

TABLE TS.3.12-1 (Page 5 of 8)

REV SAFETY RELATEL SNUBBERS Snubbers In righ Accessible or Especially Radiation Snubber Inaccessible Difficult Areas During No. Loc at ion Elevation (A or I) to Remove Shutdown UNIT I CW-359 Cooling Water 705'-8" A CW-380 706'-11" A CW-385 709'-0" A CW-394 731'-0" A CWH-395 746'-6" A CWH-405 707'-10" A CW-429 722'-11" A CWH-432 722'-11" A CW-433 735'-11" A CWH-434 735'-11" A CWH-436 737'-11" A CWRH-80 730'-0" I CWRH-81 729'-0" I CWRH-82 730'-0" I UNIT .I-CWH-34 Cooling Water 709'-3" A CWH-35 746'-8" A CW-39 710'-6" A CWH-40 710'-6" A CWH-44 730'-11" A CWH-45 709'-0" A CWH-49 723'-0" A CWH-50 723'-10" A CWH-52 736'-0" A CWH-54 738'-0" A

E TABLE TS.3.12-1 (Page 6 of 8)

REV S,AFETY RELATED SNUBBERS Snubbers In High Accessible or Especially Radiation Snubber Inaccessible Difficult Areas During No. Location Elevation (A or I) to Remove Shutdown UNIT I AFWH-72 Feedwater 752'-0" I ,

AFWR-82 728'-11" A AFWH-84 728'-11" A UNIT II AFWH-72 A&B Feedwater 706'-3/4" A FWH-72 A&B 751'-0" I UNIT I

  • 25.12620.003 - 4 760'-9-1/2" I X
  • 25.12620.003 - 5 760'-9-1/2" I X
  • 25.12620.003 - 6 760'-9-1/2" I X
  • 25.12620.003 - 7 760'-9-1/2" I X
  • 25.12620.003 - 8 760'-9-1/2" I X
  • 25.12620-003 - 10 760'-9-1/2" I X
  • 25.12620.003 - 15 760'-9-1/2" I X UNIT II
  • 25.12620.003 - 1 760'-9-1/2" I X
  • 25.12620.003 - 2 760'-9-1/2" I X
  • 25.12620.003 - 9 760'-9-1/2" I X
  • 25.12620.003 - 11 760'-9-1/2" I X
  • 25.12620.003 - 12 760'-9-1/2" I X
  • 25.12620.003 - 13 760'-9-1/2' I X
  • 25.12620.003 - 14 760'-9-1/2" I X
  • 25.12620.003 - 16 760'-9-1/2" I X UNIT I CVCH-182 Chemical & Vol 707'-6" A RCRH-16 A&B Control 705'-2" I RCRH-19 705'-2" I RCRH-21 705'-7" I RCRH-23 A&B 715'-11" I RCVCH-907 A&B 717'-11" I RCVCH-1293 712'-0" I RPCH-22 703'-1" I RPCH-23 703'-1" I RPCH-121 707'-9" I RPCH-139 704'-4" I RPCH-140 707'-7" I RPCH-146 714'-7" I RPCH-147 714'-10" I WDRH-24 707'-9" I Notes
  • Preservice documented testing of 900K Anker Holth snubbers has qualified their operability for all design conditions. Functional testing specified in 4.13 C is not required.

TABLE TS.3.12-1 (Pagt 7 of 8)

REV SAFETY RELATED SNUBBERS Snubbers In High Accessible or Especially Radiation Snubber Inaccessible Difficult Areas During No. Location Elevation (A or I) ,fo Remove Shutdown UNIT II RCVCH-1396 Chemical & Vol 702'-10" 1 RCVCH-1505 Control 708'-6" I RCVCH-1513 710'-1" I 719'-1"

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RCVCH-1524 I RCVCH-1574 721'-0" I RCVCH-1668 705'-5" I RCVCH-1373 722'-11" I RCVCH-1389 706'-1" I RRCH-253 704'-4" , I RRCH-255 704'-8" I RRCH-261 707'-2" I RRCH-288 707'-2" I RRCH-291 704'-6" I RRCH-292 704'-7" I CVCH-166 708'-0" A UNIT I CCH-304 Comp Cooling 717'-7" A CCH-373 712'-4" A CCH-376 A&B 700'-5" A CCH-377 703'-0" A CCH-378 708'-4" A CCH-380 670'-8" A CCH-381 A&B 671'-4" A CCH-397 699'-3" A CCH-398 A&B 671'-4" A UNIT II CCH-161 Comp Cooling 717'-7" A CCH-166 719'-11" A CCH-167 720'-0" A CCH-172 720'-0" A CCH-173 708'-5" A CCH-176 705'-3" A CCH-179 A&B 671'-4" A CCH-180 670'-8" A i CCH-181 708'-4" A i CCH-182 704'-2" A CCH-185 A&B 671'-4" A CCH-186 670'-10" A UNIT I RCSH-81 Containment Spray 76"'-9" I RCSH-82 760'-8" I RSCH-83 A&B 732'-1" I UNIT II CSH-75 A&B Containment Spray 731'-19" I CSH-76 752' 7" I CSH-79 751 ' -9 I CSH-82 A&B 731'-11' I CSH-83 767'-2" I CSH-84 767'-2" I CSH-210 698'-0" I CSH-215 698'-0" A CSH-224 710'-6" A

TS.4.13-1 REV 4.13 SNUBBERS Applicability Applies to periodic testing and surveillance requirec. ants of safety related hydraulic snubbers.

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Obj ect ive To verify the integrity and operability of hydraulic snubbers.

Specification The following surveillance requirements apply to all snubbers listed in Table TS.3.12-1. These requirements augment the inspections required by Section XI of the ASME Code.

A. Visual Inspection of snubbers shall be conducted in accordance with the following schedule:

No. of Snubbers Found Next Required In::?rable per Inspection Period Inspection Period 0 18 conths + 25%

1 12 months II 25%

2 6 months 17 25%

3,4 124 days f; 25%

5,6,7 62 days j; 25%

8 or more 31 days j; 25%

The required inspection incerval shall not be lengthened more than one step at a time.

Snubbers may be categorized in two groups, " accessible" or

" inaccessible" based on their accessibility for inspection during reactor operation. These two group: may be inspected independently according to the above schedule.

B. Visual inspections shall verify (1) that there are no visible indications of damage or impaired operability, (2) attachments to the supporting structure are secure, and (3) in those locations where snubber movement can be manually induced without disconnect-ing the snubber, that the snubber has freedom of movement and is not frozen up. Snubbers which appear inoperable as a result of visual inspection may be determined Operable for the purpose of establishing the next visual inspection interval by:

TS.4.13-2 REV

a. Clearly establishing the cause of the rejection for that particular snubber and for others that may be generically susceptible; and
b. Functionally testing the affected snubber in the as-found condition and finding it operable per Specification 4.13.D.

i However, when the fluid port of a hydraulic snubber is found to be uncovered, the snubber shall be considered inoperable for purposes -

of establishing the next visual inspection interval. All hydraulic snubbers connected to an inoperable common hydraulic fluid reservoir shall be considered as inoperable snubbers.

C. Except as specified below, functional testing of snubbers listed in Table TS.3.12-1 shall be conducted at least once per 18 months during cold shutdown. Ten percent of the total of each type snubber shall be functionally tested either in place or in a bench tet. For each snubber that does not meet the functional test acceptance criteria in Specification 4.13.D below, an additional ten percent of that type of snubber shall be functionally tested until no more failures are found or all snubbers of that type have been tested.

The representative sample selected for functional testing shall include the various configurations, operating environments, and the range of size and capacity of the snubbers. Twenty five percent of the sample shall include snubbers from the following three categories,

a. The first snubber away from a reactor vessel nozzle
b. Snubbers within five feet of heavy equipment (valve, pump, turbine, motor, etc)
c. Snubbers within ten feet of the discharge of a safety /

relief valve Snubbers identified in Table TS.3.12-1 as "High Radiation Area" or

" Difficult to Remove" are exempt from functional testing provided a justifiable basis for exemption is presented for Commission review; snubber life testing is performed to qualify snubber operability for all design conditions; or snubbers of the same type, configuration, and similar service have been tested for a ten year period and no failures have occurred. In such exempt cases, a qualitative test report shall be on file to substitute for the required functional testing.

In addition to the regular sample and specified re-sampling, snubbers which failed the previous functional test shall be 4

retested during the next test period. If a spare snubber has been installed in place of a failed snubber, then both the failed i snubber, if it is repaired and installed in another position, and the spare snubber shall be retested.

TS.4.13-3 REV i

If any snubber selected for functional testing either fails to lockup or fails to move (i.e. frozen in place) the cause shall be evaluated and all snubbers subject to the same defect shall be functionally tested. This testing is in addition to the regular sample and specified re-samples.

D. Hydraulic snubber functional tests shall verify that:

a. Activation (restraining action) is achieved within the specified range of velocity or acceleration in both tension and compression.
b. Snubber bleed, or release rate, where required, is within the specified range in compression or tension. For snubbers specifically required to not dicplace under continuous load, the ability of the snubber to withstand load without displace-ment shall be verified.

E. An engineering evaluation shall be performed for all components supported by inoperable snubbers. The purpose of this engineering evaluation shall be to determine if the components were adversely affected by the inoperable snubber (s) to ensure that the components remain capable of meeting the designed service.

F. The installation and maintenance records for each snubber listed in Table TS.3.12-1 shall be reviewed at least once every 18 months to verify that the indicated service life will not be exceeded prior to the next scheduled snubber service life review. If the indicated service life will be exceeded, the snubber service life shall be reevaluated or the snubber shall be replaced or recondi-tioned to extend its service life beyond the date of the next scheduled service life review. This reevaluation, replacement, or reconditioning shall be indicated in the records.

Basis The visual inspection frequency is based upon maintaining a constant level of snubber protection to systems. Therefore, the required inspection interval varies inversely with the observed snubber failures and is determined

! by the number of inoperable snubbers found during an inspection. Inspections performed before that interval has elapsed may be used as a new reference point to determine the next inspection. However, the results of such early inspections performed before the original required time interval has elapsed (nominal time less 25%) may not be used to lengthen the required inspection interval. Any inspection whose results require a shorter inspection interval will override the previous schedule.

TS.4.13-4 REV When the cause of the rejection of a snubber is clearly established and remedied for that snubber and for any other snubbers that may be generically susceptible, and verified by inservice functional testing, that snubber may be exempted from being counted as inoperable. Generically susceptible snubbers are those which are of a specific make or model and have the same design features directly related to rejection of the snubber by visual inspection, or are similarly located or exposed to the same environmental conditions such as temperature, radiation, and vibration. -

When a snubber is found inoperable, an engineering evaluation is performed, in addition to the determination of the snubber mode of failure, in order to determine if any safety-related component or system has been adversly af fected by the inoperability of the snubber. The engineering evaluation shall determine whether or not the snubber mode of failure has imparted a significant effect or degradation on the supported component or system.

To provide assurance of snubber functional reliability, a representre.ive sample of 10% of the installed snubbers will be functionally tested during plant shutdowns at 18 month intervals. Observed failures of these sample snubbers shall require functional testing of additional units.

The service life of a snubber is evaluated via manufacturer input and information through consideration of the snubber service conditions and associated installation and maintenance records (newly installed snubber,

- seal replaced, spring replaced, in high radiation area, in high temperature area, etc. . .). The requirement to monitor the snubber service life is included to ensure that the snubbers periodically undergo a performance evaluation in view of their age and operating conditions. These records will provide statistical bases for future consideration of snubber service life.

All safety-related snubbers installed or planned for use at Prairie Island are hydraulic snubbers. No mechanical snubbers are used.

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TS.'.6-2 REV

'6. Plant radiation and contamination surveys.

7. Changes made to the plant as it is described in the Final Safety Analysis Report, reflected in updated, corrected and as-built drawings.
8. Cycling beyond normal limits for those components that have been designed to operate safely for a limited number of cycles beyond such limits.
9. Reactor coolant system in-service inspections.
10. Minutes of meetings of the Safety Audit Committee.
11. Records of Environmental Qualification which are covered under the provisions of paragraph 6.8.
12. Records of the service lives of all safety-related snubbers, including the date at which the service life commences and associated installation and maintenance records.

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