Proposed Tech Specs 3.14,4.0,4.1,4.17 & 4.18 Requiring Limiting Condition for Operation for Hydraulic Shock Suppressor Operability & Appropriate Surveillance Requirements to Assure Suppressor Performance & Reliability| ML19312C010 |
| Person / Time |
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| Site: |
Oconee  |
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| Issue date: |
08/15/1975 |
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| From: |
DUKE POWER CO. |
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| To: |
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| Shared Package |
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| ML19312C007 |
List: |
|---|
| References |
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| NUDOCS 7911270784 |
| Download: ML19312C010 (15) |
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Category:TECHNICAL SPECIFICATIONS
MONTHYEARML20155C6391998-10-26026 October 1998
Proposed Tech Specs Tables 4.1-1 & 4.1-2,revised Surveillance Frequency of Refueling Outage to 18 Months
ML20154C1021998-09-30030 September 1998
Proposed Tech Specs Changing UFSAR by Increasing Max Rod Internal Pressure in Spent Fuel Pool from 1,200 to 1,300 Psig
ML20236V9611998-07-27027 July 1998
Corrected TS Bases Pages 3.5-30b of 961211 & 3.16-2 of 970205,correcting Amend Numbers
ML20236T5171998-07-20020 July 1998
Proposed Tech Specs Re one-time Extension to Functional Test Frequencies for Hydraulic & Mechanical Snubbers
ML20236R1901998-07-16016 July 1998
Proposed Tech Specs Extending TS Surveillances TS 4.20.2(a), Table 4.20-1,items 1,2 & 4,TS 4.1-1,Table 4.1-1,items 8,9,10 & 11 & TS 4.5.1.2.1,item b(1)
ML16141A8111994-04-20020 April 1994
Proposed Tech Spec Table 4.1-2, Min Equipment Test Frequency.
ML20044C9581993-05-0404 May 1993
Proposed TS 4.7.1, Control Rod Trip Time Test.
ML18032A3441987-05-29029 May 1987
Proposed Tech Specs,Clarifying Trip Level Setting in Table 3.2.A for Standby Gas Treatment Sys Relative Humidity Heater
ML16134A6761982-08-11011 August 1982
Proposed Tech Spec Revisions Re Reload Design Calculations for Cycle 7
ML16134A6731982-05-0303 May 1982
Proposed Tech Spec Changes Re Core Protection Safety Limits, Protective Sys Max Allowable Setpoints & Rod Position Limits
ML15223A7671982-01-12012 January 1982
Proposed Revisions to Tech Spec Section 3.1.2 Re Heatup Cooldown & Inservice Test Limitations for RCS
ML16148A4421981-11-13013 November 1981
Proposed Tech Spec Revision Re Core Protection Safety Limits Protective Sys Max Allowable Setpoints & Rod Position Limits
ML16148A4351981-10-28028 October 1981
Proposed Revision to Tech Spec Figure 3.5.2-4B2,allowing Cycle 5 to Run at 100% Full Power W/Axial Power Shaping Rods Fully Inserted
ML16148A4281981-08-19019 August 1981
Proposed Revision to Tech Spec Figures 3.5-16a,3.5.19a, 3.5-22 & 3.5-25a Re Extension of Operating Limits
ML15223A7321981-05-29029 May 1981
Proposed Tech Specs 2.1-2,2.1-3,2.1-7,2.3-5,3.2-1,3.2-2, 3.3-1,3.3-2,3.3-3,3.3-4,3.3-5,3.3-6,3.5-9,3.5-10,3.5-15, 3.5-15a & b,3.5-18,3.5-18a,b,c,d & e,3.5-21,3.5-21a & B, 3.5-24,3.5-24a & b,3.8-2 & 3.8-3 Re Core Protection
ML16134A6651980-08-25025 August 1980
Proposed Tech Specs Revision for Cycle 6
ML16148A3351980-07-16016 July 1980
Proposed Revision to Tech Spec 3.3.1.c Allowing Continued Operation of Unit 2 of Full Rated Power While Maint Continues on HPI Pump Until 800718
ML19318B9731980-06-24024 June 1980
Proposed Tech Spec Interpreting Term Operable as Applied to Various Tech Spec Requirements
ML16148A2691979-11-16016 November 1979
Proposed Changes to Tech Spec Pages 2.1,2.3,3.2 & 3.5. Changes Affect Core Protection Limits,Reactor Protective Sys Max Allowable Setpoints & Vol Requirements for Borated Water Storage Tank
ML19317D2701978-09-25025 September 1978
Proposed Tech Spec 3.5.2 Re Control Rod Group & Power Distribution Limits & Table 4.1-2 Re Min Equipment Test Frequency
ML19317D2621978-09-18018 September 1978
Proposed Revisions to Tech Specs 2.1,2.3,3.2 & 3.5 Re Core Protection Safety Limits & Protective Sys Max Allowable Setpoints
ML19317D2731978-09-0606 September 1978
Revised Tech Spec Page,Figure 2.3-2A,re Protective Sys Max Allowable Setpoints
ML19317D2491978-08-22022 August 1978
Proposed Revision to Tech Specs 4.18 Re Hydraulic Shock Suppressors (Snubbers)
ML19322B9961978-08-21021 August 1978
Proposed Revision to Tech Spec 2.3 Re Cycle 5
ML19312B7931978-07-17017 July 1978
Proposed Tech Spec 3.1.6.4,changing Steam Generator Leak Rate Limit
ML19308D6271978-06-28028 June 1978
Tech Spec Change Request Re Paragraph 2.B(6),stipulating That Byproduct & SNM Associated W/Four Fuel Assemblies Acquired by Fl Power Corp from Duke Power Co Previously Irradiated in Oconee 1 May Be Possessed
ML19317D2301978-06-26026 June 1978
Proposed Tech Specs 2.1,2.3,3.2,3.5 & 4.1 Required to Support Operation of Unit 1 at Full Rated Power During Cycle 5,including Core Protection Safety Limits & Protective Sys Mac Allowable Setpoint
ML19316A6501978-06-22022 June 1978
Proposed Replacement Page for Tech Spec 4.1-2 Re Min Equipment Test Frequency
ML19316A5381978-06-14014 June 1978
Proposed Changes to Tech Specs Re thermal-hydraulics Analysis.Revision to BAW-1486, Unit 3,Cycle 4 Reload Rept.
ML19312B8161978-06-12012 June 1978
Proposed Tech Specs 3.8,4.4 & 4.6 Re Fuel Loading & Refueling,Structural Integrity & Emergency Power Periodic Testing
ML19317D2341978-06-0909 June 1978
Proposed Tech Spec 3.9 Deleting Requirements Not Applicable to Liquid Effluent Monitoring Sys Due to Installation of Offline Monitor
ML19317D2221978-06-0808 June 1978
Proposed Tech Spec 3.1 Allowing Max 1 Gallon Per Minute Leakage Through Steam Generator Tubes Prior to Initiation of Unit Shutdown
ML19317D2121978-06-0202 June 1978
Proposed Tech Spec 4.2 Allowing re-insp of Reactor Coolant Outlet Nozzles at Future Refueling Outage
ML19316A5271978-05-30030 May 1978
Proposed Revisions to Tech Specs 2.3,3.2 & 3.5.2.4 to Support Cycle 4 Operation at Full Power
ML19312B7971978-04-27027 April 1978
Proposed Tech Spec 6.4,incorporating Operating Procedure Requirements Re B&W Small Break ECCS Analysis
ML19312B8091978-04-20020 April 1978
Proposed Tech Spec 3.3 Incorporating New Tech Spec 3.3.8 Requiring Operability of Three HPI Pumps for Each Unit During Power Operation Above 60% Full Power
ML19317D2001978-03-20020 March 1978
Proposed Tech Spec 3.5 Including 6.03% Quadrant Power Tilt Limit & Provision for Notifying NRC If Tilt Exceeds 3.5%
ML19317D2131978-02-21021 February 1978
Proposed Tech Spec 3.1. Incorporating Revision to Pressurization,Heatup & Cooldown Limitations.B&W 780125 Ltr to Util Re Corrections to Errors Discovered in B&W Rept BAW-1436 Encl
ML19317D2211978-02-16016 February 1978
Proposed Tech Spec 2.3 Deleting Loss of One Pump Trip Setpoint,Outdated Info & Setpoints Associated W/Single Loop Operation
ML19340A2641978-02-0808 February 1978
Tech Specs 3.3.2 Through 3.3.5 for ECCS
ML19316A4801978-02-0101 February 1978
Proposed Changes to Tech Spec 3.7 Re Limiting Conditions for Operation & Surveillance Requirements for 125 Volt Distribution Sys
ML19317D2271978-01-23023 January 1978
Proposed Tech Spec 3.5 Incorporating Control Rod Position & Axial Imbalance Limits to Period After 100 Plus or Minus 10 Effective Full Power Days
ML19312B7891978-01-0303 January 1978
Proposed Tech Spec 2.3-9 & 10 Re Computer Software Used to Process Incore Detector Signal.Includes Modified power-imbalance Trip Setpoints to Account for Bias in Positive Imbalance Measured by Incore Detector Sys
ML19316A5081977-12-0202 December 1977
Amend to Tech Specs 3.9 Re Radwaste Discharge
ML19312B8001977-12-0202 December 1977
Proposed Tech Spec 6.6.2.1 Providing Requirement for Prompt Written Notification of Certain ROs by Telephone,Mailgram or Facsimile Transmission
ML19312B7771977-11-0909 November 1977
Proposed Tech Spec 3.5.2 Permitting Operation of Unit 1 During Cycle 4 in Unrodded Mode
ML19312B8061977-10-31031 October 1977
Proposed Tech Specs 6.6-1,-2,-3 & -4 Deleting Redundant Info Currently Being Reported in Annual Operating Rept
ML19312B8101977-10-26026 October 1977
Proposed Tech Specs 3.5-9 & 3.5-24a,deleting Existing Reactor Core Quadrant Power Tilt & Control Rod Position Limits & Instituting More Conservative Limits
ML19312B8151977-10-0707 October 1977
Proposed Changes to Tech Specs 3.7 & 4.6 Re Auxiliary Electrical Systems & Emergency Power Periodic Testing
ML19329A4011977-10-0606 October 1977
Revised Tech Specs,Table 4.1-3 Re Min Sampling Frequency
1998-09-30
[Table view]
Category:TECHNICAL SPECIFICATIONS & TEST REPORTS
MONTHYEARML20211A9881999-08-18018 August 1999
Rev 5 to DPC Nuclear Security Training & Qualification Plan
ML20204B4141999-03-27027 March 1999
Revised Oconee Nuclear Station Selected Licensee Commitments, List of Effective Pages
ML20155C6391998-10-26026 October 1998
Proposed Tech Specs Tables 4.1-1 & 4.1-2,revised Surveillance Frequency of Refueling Outage to 18 Months
ML20154L7191998-10-0505 October 1998
Rev 8 to DPC Nuclear Security & Contingency Plan, for Oconee,Mcguire & Catawba Nuclear Stations
ML20154C1021998-09-30030 September 1998
Proposed Tech Specs Changing UFSAR by Increasing Max Rod Internal Pressure in Spent Fuel Pool from 1,200 to 1,300 Psig
ML20236V9611998-07-27027 July 1998
Corrected TS Bases Pages 3.5-30b of 961211 & 3.16-2 of 970205,correcting Amend Numbers
ML20236T5171998-07-20020 July 1998
Proposed Tech Specs Re one-time Extension to Functional Test Frequencies for Hydraulic & Mechanical Snubbers
ML20236R1901998-07-16016 July 1998
Proposed Tech Specs Extending TS Surveillances TS 4.20.2(a), Table 4.20-1,items 1,2 & 4,TS 4.1-1,Table 4.1-1,items 8,9,10 & 11 & TS 4.5.1.2.1,item b(1)
ML20217F2841998-04-20020 April 1998
Rev 7 to DPC Nuclear Security & Contingency Plan, for Oconee,Mcguire & Catawba Nuclear Stations
ML20141F1521997-06-25025 June 1997
Rev 4 to Nuclear Security Training & Qualification Plan
ML20134K5401996-10-31031 October 1996
Rev 6 to Chemistry Manual 5.1, Emergency Response Guidelines
ML20117J0181996-08-15015 August 1996
Revised Chapter 16 of Oconee Selected Licensee Commitments Manual
ML20095H1291995-12-0505 December 1995
Rev to ONS Selected Licensee Commitments (SLC) Manual, Revising SLC 16.6.1, Containment Leakage Tests to Reflect Current Plant Configuration & Update Testing Info
ML20091P2461995-08-21021 August 1995
Rev to ONS Selected Licensee Commitments Manual
ML16141A8111994-04-20020 April 1994
Proposed Tech Spec Table 4.1-2, Min Equipment Test Frequency.
ML15224A3521993-10-26026 October 1993
Safety Assurance Directive 6.1, Oconee Nuclear Site Safety Assurance Emergency Response Organization.
ML20044C9581993-05-0404 May 1993
Proposed TS 4.7.1, Control Rod Trip Time Test.
ML20045F0721993-04-0707 April 1993
Rev 9 to Corporate Process Control Program Manual.
ML20097D9451992-05-28028 May 1992
Rev 11 to Training & Qualification Plan
ML20096C2561992-04-30030 April 1992
Public Version of Revised Crisis Mgt Implementing Procedures (Cmip),Including Rev 46 to CMIP-1,Rev 39 to CMIP-4,Rev 45 to CMIP-5,Rev 49 to CMIP-6,Rev 48 to CMIP-7,Rev 42 to CMIP-9 & Rev 3 to CMIP-15
ML20096D5451992-04-0707 April 1992
Public Version of Revised Crisis Mgt Implementing Procedures (Cmip),Including Rev 45 to CMIP-1,Rev 27 to CMIP-13 & Notification of Deletion of CMIP-8.Procedure CMIP-8 Reserved for Future Use
ML20092M5891992-02-0606 February 1992
Public Version of Revised Crisis Mgt Implementing Procedures,Including Rev 43 to CMIP-1,Rev 43 to CMIP-5,Rev 33 to CMIP-8,Rev 25 to CMIP-13,Rev 6 to CMIP-18,Rev 4 to CMIP-22 & Rev 38 to CMIP-4
ML20094H2391992-01-0101 January 1992
Rev 33 to McGuire Nuclear Station Odcm
ML20094H2511992-01-0101 January 1992
Rev 34 to Catawba Nuclear Station Odcm
ML20087F3931991-12-11011 December 1991
Public Version of Rev 13 to Crisis Mgt Implementing Procedure CMIP-11, Emergency Classification - Mcguire. W/ 920121 Release Memo
ML20086K8831991-11-18018 November 1991
Public Version of Revised Crisis Mgt Implementing Procedures (Cmips),Including Rev 42 to CMIP-1,Rev 29 to CMIP-2,Rev 42 to CMIP-5,Rev 12 to CMIP-11 & Rev 37 to CMIP-21
ML20086G7711991-10-16016 October 1991
Public Version of Revs to Crisis Mgt Implementing Procedures (Cmip),Including Rev 41 to CMIP-1,rev 37 to CMIP-4,rev 41 to CMIP-5,rev 46 to CMIP-6 & CMIP-7,rev 32 to CMIP-8,rev 40 to CMIP-9,delete CMIP-12 & Rev 24 to CMIP-13
ML20082C7441991-06-11011 June 1991
Public Version of Rev 13 to Crisis Mgt Implementing Procedure CMIP-12, Classification of Emergency for Oconee Nuclear Station
ML20076A7241991-06-10010 June 1991
Public Version of Crisis Mgt Implementing Procedures, Including Rev 39 to CMIP-1,rev 28a to CMIP-2,rev 44 to CMIP-7,rev 39 to CMIP-9 & Rev 10 to CMIP-11
ML20081J9841991-06-10010 June 1991
Rev 3 to EDA-1, Procedure for Estimating Food Chain Doses Under Post-Accident Conditions
ML20081K0101991-06-0606 June 1991
Rev 8 to EDA-3, Offsite Dose Projections for McGuire Nuclear Station
ML20072S0521991-03-15015 March 1991
Public Version of Revised Crisis Mgt Implementing Procedures,Including Rev 9 to CMIP-11, Classification of Emergency for McGuire Nuclear Station & Rev 11 to CMIP-12, Classification of Emergency for Oconee Nuclear Station
ML20066G3621991-02-0101 February 1991
Public Version of Crisis Mgt Implementing Procedures, Including Rev 28 to CMIP-2,Rev 34 to CMIP-4,Rev 38 to CMIP-5,Rev 43 to CMIP-6,Rev 42 to CMIP-7,Rev 29 to CMIP-8, Rev 37 to CMIP-9 & Rev 34 to CMIP-21
ML20082P7611991-01-0101 January 1991
Rev 30 to Odcm,Catawba Nuclear Station
ML20072S9621991-01-0101 January 1991
Public Version of Rev 12 to Crisis Mgt Implementing Procedure CMIP-12, Classification of Emergency for Oconee Nuclear Station
ML20082P7711991-01-0101 January 1991
Rev 31 to Odcm,Mcguire Nuclear Station
ML20072P9621990-11-0808 November 1990
Rev 9 to Crisis Mgt Implementing Procedure CMIP-12, Classification of Emergency for Oconee Nuclear Station
ML20028H2211990-10-31031 October 1990
Public Versions of Revised Crisis Mgt Implementing Procedures,Including Rev 37 to CMIP-1,Rev 27a to CMIP-2,Rev 33 to CMIP-4,Rev 37 to CMIP-5 & Rev 42 to CMIP-6
ML20059F4301990-08-22022 August 1990
Public Version of Rev 27 to Crisis Mgt Implementing Procedure CMIP-2, News Group Plan
ML20063Q2721990-08-14014 August 1990
Public Version of Revised Crisis Mgt Implementing Procedures,Including Rev 36 to CMIP-1,Rev 32 to CMIP-4,Rev 36 to CMIP-5,Rev 41 to CMIP-6,Rev 40 to CMIP-7,Rev 27 to CMIP-8,Rev 35 to CMIP-9 & Rev 7 to CMIP-11
ML20043F4841990-05-23023 May 1990
Public Version of Crisis Mgt Implementing Procedures, Consisting of Rev 20 to CMIP-13 & Deletion of CMIP-15 & CMIP-17
ML20043B2651990-05-0909 May 1990
Revised Crisis Mgt Implementing Procedures,Including Rev 35 to CMIP-1,Rev 26 to CMIP-2,Rev 31 to CMIP-4,Rev 35 to CMIP-5,Rev 40 to CMIP-6,Rev 39 to CMIP-7,Rev 26 to CMIP-8, Rev 33 to CMIP-9,Rev 2 to CMIP-14 & Rev 10 to CMIP-16
ML20043F4621990-04-20020 April 1990
Rev 5 to Oconee-specific Process Control manual.W/900606 Ltr
ML20006C0571990-01-18018 January 1990
Public Version of Rev 33 to Crisis Mgt Implementing Procedure CMIP-1, Recovery Manager & Immediate Staff & Rev 24 to CMIP-2, News Group Plan.
ML16152A8951990-01-0202 January 1990
Rev 33 to Public Version of Crisis Mgt Plan for Nuclear Stations.
ML15264A1571990-01-0202 January 1990
Revised Crisis Mgt Implementing Procedures,Including Rev 32 to CMIP-1,Rev 29 to CMIP-4,Rev 33 to CMIP-5,Rev 38 to CMIP-6,Rev 37 to CMIP-7,Rev 32 to CMIP-9,Rev 1 to CMIP-14 & Rev 30 to CMIP-21
ML20012A3801990-01-0101 January 1990
Rev 28 to, Offsite Dose Calculation Manual,Oconee,Mcguire & Catawba Nuclear Stations.
ML20012A3791990-01-0101 January 1990
Rev 27 to, Offsite Dose Calculation Manual,Oconee Nuclear Station.
ML20011D2441989-12-0101 December 1989
Crisis Mgt Implementing Procedures.
ML20012A3731989-11-15015 November 1989
Rev 4 to, Process Control Program Oconee Nuclear Station.
1999-08-18
[Table view] |
Text
__ - _ _ _
n o _
3.14 HYDRAULIC SHOCK SUPPRESSORS Applicability Applies to all modes of operation except cold shutdown and refueling shut-down.
Objective To assure piping integrity in the event of a severe transient or seismic disturbance.
Specification 3.14.1 Except as permitted by 3.14.2 and 3.14.3, the reactor shall not be heated above 200 F unless all hydraulic shock suppressors listed in Table 4.18-1 are operable.
3.14.2 If a hydraulic shock suppressor is determined to be inoperable, continued operation is permitted for a period not to exceed 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, unless the suppressor is sooner made operable.
3.14.3 If the requirements of 3.14.1 and 3.14.2 cannot be met, the reactor shall be in a cold shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
Bases Suppressors are designed to prevent unrestrained pipe motion under dynamic ,
loads as might occur during an earthquake or severe transient, while allowing l normal thermal motion during startup and shutdown. The consequence of an inoperable suppressor is an increase in the probability of structural damage to piping as a result of a seismic or other event initiating dynamic loads.
It is therefore required that all hydraulic suppressors required to protect the primary coolant system or any other safety system or component be operable during reactor operation.
Since the suppressor protection is required only during relatively low probability events, a period of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is allowed for repairs or replace-ments. In case a shutdown is required, the allowance of 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> to reach a cold shutdown condition will permit an orderly shutdown consistent with
~
standard operating procedures. Since plant startup should not commence with knowingly defective safety-related equipment, Specification 3.14.1 prohibits startup with inoperable suppressors.
l l
f I
3 '-
yens'ro7Pp
~ r, 4 SURVEILLANCE RE0HIREMENTS 4.0 SURVEILLANCE STANDARDS Applicability Applies to surveillance requirements which relate to tests, :alibrations and inspections necessary to assure that the quality of structures, systems and components is maintained and that operation is within the safety limits and limiting conditions for operation.
Obj ective To specify minimum acceptable surveillance requirements.
Specification 4.0.1 Surveillance of structures, systems, components and parameters shall be as specified in the various subsections to this Technical Specification section, Section 4.0, except as per-mitted by Technical Specifications 4.0.2 and 4.0.3 below.
4.0.2 Minimum surveillance frequencies, unless specified otherwise, may be adjusted as follows to facilitate test scheduling:
Maximum Allowable Specified Frequency Interval Between Surveillances Five times per week 2 days Two times per week S days Weekly 10 days Bi-Weekly 20 days Monthly 45 days Bi-Monthly 90 days Quarterly 135 days Semiannually 270 days Annually 18 months l 18 Months 24 months l 4.0.3 ' If conditions exist such that surveillance of an item is not necessary to assure that operation is within the safety limits and limiting conditions for operation, surveillance need not be performed if such conditions continue for a length of time greater than the specified surveillance interval. Surveillance waived as a result of this specification shall be performed prior to returning to conditions for which the surveillance is necessary to assure that operation is within safety limits and limiting conditions for operation.
4.0-1
Y%
Table 4.1-2 MINIMUM EQUIPMENT TEST FREQUENCY Item Test Frequency
- 1. Control Rod Movement ( } Movement of Each Rod Bi-Weekly
- 2. Pressurizer Safety Valves Setpoint 50% Annually
- 3. Main Steam Safety Valves Setpoint 25% Annually
- 4. Refueling System Interlocks Functional Prior to Refueling
- 5. Main Steam Stop Valves (l) Movement of Each Stop Monthly Valve
- 6. Reactor Coolant System ( } Evaluate Dail, Leakage
- 7. Condenser Cooling Water Functional Annually System Gravity Flow Test
- 8. High Pressure Service Water Functional Monthly Pumps and Power Supplies
- 9. Spent Fuel Cooling System Functional Prior to Refueling
( ) Applicable only when the reactor is critical.
( ) Applicable only when the reactor coolant is above 200 F and at a steady-state temperature and pressure.
4.1-9
A
.,' .a 4.17 (RESERVED) i e
t 4.17-1
,n 4.18 HYDRAULIC SHOCK SUPPRESSORS Applicability Applies to hydraulic shock suppressors used to prctect the Reactor Coolant System or other safety-related systems.
Obj ective To verify that required hydraulic shock suppressors are operable.
Specification 4.18.1 All hydraulic shock suppressors listed in Table 4.18-1 whose seal material has been demonstrated by operating experience, lab testing or analysis to be compatible with the operating environment shall be visually inspected to verify operability as follows:
Number of Suppressors Found Next Required Inoperable During Last Inspection Inspection Interval 0 18 months 1 Annually 2 Semiannually 3,4 Triannually :
5,6,7 Bi-Monthly :
>8 Monthly Note: The required inspection interval shall not be lengthened more than one step per inspection.
4.18.2 All hydraulic shock suppressors listed in Table 4.18-1 whose seal materials have act been demonstrated to be compatible with the operating environment shall be visually inspected for operability monthly.
4.18.3 Every 18 months at least two representative suppressors from a relatively severe environment shall be completely disassembled and examined for damage and abnormal seal degradation.
Bases l
l All saf ety-related hydraulic suppressors are visually inspected for overall l integrity and operability. The inspection will include verification of proper j orientation, adequate hydraulic fluid level and proper attachment of suppressor to piping and structures.
l The inspection frequency is based upon maintaining a constant level of l suppressor protection. Thus the required inspection interval varies inversely with the observed suppressor failures. The number of inoperable suppressors found during a required inspection determines the time interval 4.18-1
?
- a d for the next required inspection. Inspections performed before that interval has elapsed may be used as a new reference point to determine the next in-spection. However, the results of such early inspections performed before the original required time interval has elapsed may not be used to lengthen the required inspection interval. Any inspection whose results require a shorter inspection interval will override the previous schedule.
Experience at operating facilities has shown that the required surveillance program should assure an acceptable level of suppressor performance provided that the seal materials are compatible with the operating environment.
Suppressors containing seal material which has not been demonstrated by operating experience, lab tests or analysis to be compatible with the operating environment should be inspected more frequently (every month) until material I compatibility is confirmed or an appropraite changeout is completed.
Examination of defective suppressors at reactor facilities and material tests performed at several laboratories (Reference 1) has shown that millable gum polyurethane deteriorates rapidly under the temperature and moisture conditions present in many suppressor locations. Although molded polyurethane exhibits greater resistance to these conditions, it also may be unsuitable for appli-cation in the higher temperature environments. Data are not currently available to precisely define an upper temperature limit for the molded polyurethane. Lab tests and in-plant experience indicate that seal materials are available, primarily ethylene propylene compounds, which should give satisfactory performance under the most severe conditions expected in reactor installa tions.
To complement the visual external inspections, disassembly and internal examination for component damage and abnormal seal degradation should be performed. The examination of two units, each refueling cycle, selected from relatively severe environments should adequately serve this purpose.
Any observed wear, breakdown or deterioration will provide a basis for additional inspections.
REFERENCE (1) Report, H. R. Erickson, Bergen-Patterson, to K. R. Goller, NRC, October 7, 1974.
Subject:
Hydraulic Shock Sway Arrestors 4.18-2
m .-
t Table 4.18-1 Safety-Related Hydraulic Shock Suppressors Oconee 1 Location (Engineering System Number) Sketch / Hanger Number Main Steam Line (01A) 1-124 1-125 1-127 1-128 1-129 l-130 1-132 1-134 1-135 1-147 1-149 l-151 1-152 H llA H 12A H 10B H llB Main Steam Bypass To Condenser (01 A-1) 1-941 1-944 1-945 Main Steam Supply to Auxiliary Equipment ( 01A-3) 1-3135 Main Steam Supply to Emergency Feedwater Pump Turbine 1-1305 (OlA-4) 1-1310 l 1-1315 Main Feedwater Line (03) H 7B H 10A l I
Emergency Feedwater Line (03A) 1-1289 1-1292 1-1293 1-1294 i
1-1295 i 1-1296
( 1-1297
- l-1298 j 1-1299 l 4.18-3
~
,m Table 4.18-1 (Continued)
Location (Engineering System Number) Sketch / Hanger Number Emergency Feedwater Line (03A) (Continued) 1-5600 1-5601 1-5602 1-5603 1-5604 1-5605 1-5606 H 7B Reactor Coolant System (50) 1-4100 1-4102 1-4104 1-4105 1-4107 1-4109 1-4111 1-4112 1-4113 1-4115 1-4116 1-4117 H1 H3
, H4
5 2- 7
- . 8 H9 H 10 H 11 H 12 H 1A H 2A H 3A High Pressure Injection Sysr.em (51) H 17A )
H lE l 1
i Low Pressure Injection System (53) H 5 (2,NS-EW)
H 40C H 41C o
Reactor Building Spray System (54) 1-2139 1-2149 H 9A E 9B 4-18-4
~
n Table 4.18-1 (Continued) l Location (Engineering System Number) Sketch / Hanger Number 1
Pressurizer Relief Valve Discharge (57) H5 H6 H9 H 10 H 11 H 14 l H 15 H 17 H 18 j H 22 i H 26 H 27 I
t 1
l 4
l I
]
4 l
S 1
l l
i i
1 4.18-5
. l 1
- - _ . . - - - . , . , _ . . , _ _ , , . _ _ . ._ . . - . _ __ l
Table 4.18-1 Safety-Related Hydraulic Shock Suppressors Oconee 2 Location (Engineering System Number) Sketch / Hanger Number Main Steam Line (OlA) 2-127 2-128 2-129 2-130 2-134 2-135 2-147 2-149 2-151 2-152 H 2A H 8A H 2B H 8B Main Steam Bypass to Condenser (OlA-1) 2-941 2-944 2-945 Main Steam Supply to Auxiliary Equipment (OlA-3) 2-3135 -
Main Steam Supply to Emergency Feedwater Pump Turbine 2-1309 (OlA-4) 2-1322 2-1323 2-1324 2-1326 2-1327 2-1329 2-1333 Main Feedwater Line (03) H 6A & H 7A H 6B Emergency Feedwater Line (03A) 2-1289 2-5656 2-5663 2-5685 2-5691 H 1A H 3A 4.18-6 l
i
.m Table 4.18-1 (Continued)
Locacion (Engineering System Number) Sketch / Hanger Number i
Emergency Feedwater Line (03A) (Cont'd) H 5A H 7A H 1B Reactor Coolant System (50) 2-4100 2-4105 2-4107 2-4109 2-4111 2-4112
- 2-4113 2-4114 2-4115 2-4117 l 2-4119 2-4120 )
H1 l H3 l H4 l H5 H7 H8 H9 H 10 H 11 H 12 H 1A H 2A H 3A High Pressure Injection System (51) 2-4482 H 2A H 1E Low Pressure Injection (53) 2-2086 2-2089 2-4206 H3 H 1E Reactor Building Spray System (54) 2-2139 2-2149 2-2172 2-2174 H 9A H 9B 4.18-7
a - m A- a:- a - + - - - - e s- - u m -
Table 4.18-1 (Continued) 1 Location (Engineering System Number) Sketch / Hanger Number Spent Fuel Cooling (56) H9 H 10 Pressurizer Relief Valve Dir.:harge (57) H7 H9 H 15 H 16 H 17 1
H 20 I H 21 H 23 l
H 25 3 H 26 i
l I
f 1
1
. 4.18-8 i
~'
A Table 4.18-1 Safety-Related Hydraulic Shock Suppressors Oconee 3 Location (Engineering System Number) Sketch / Hanger Number Main Steam Line (01A) 3-124 3-125 3-126 3-128 3-129 3-130 3-131 3-132 3-133 3-135 3-147 3-149 I H 2A H 8A H 2B H 8B Main Steam Bypass to Condenser (01A-1) 3-956 3-957 3-959 3-960 Main Steam Supply to Auxiliary Equipment (01A-3) 3-3109 Main Steam Supply to Emergency Feedwater Pump Turbine 3-1311 (OLA-4) 3-1312 l 3-1314 l 3-1316 3-1317 3-1318 3-1319 3-1320 Main Feedwater Line (03) H 6A & H 7A l H 6h Emergency Feedwater Line (03A) 3-1274 3-1379 3-1280 3-5606 3-5624 3-5628 4.18-9 H 1A
,.m s Table 4.18-1 (Continued)
Location (Engineering System Number) Sketch / Hanger Number Reactor Coolant System (50) 3-4100 ;
3-4105 3-4107 3-4109 3-4111 3-4112 3-4113 3-4114 3-4115 3-4117 3-4119 3-4120 H1 H3 H4 H5 H7 H8 H9 H 10 H 11 H 12 H 1A H 2A H 3A High Pressure Injection System (51) 3-2214 H 2A H 1E Low Pressure Injection System (53) 3-4271 3-4273 3-4280 3-4281 3-4282 3-4287 3-4288 H3 H IC Reactor Building Spray System (54) 3-2140 3-2165 3-2174 H 9A i
H 9B l
l 4.18-10 l
l l
J Table 4.18-1 (Continued)
Location (Engineering System Number) Sketch / Hanger Number Spent Fuel Cooling System (56) 3-5700 3-5703 3-5707 3-5709 3-5712 3-5716 3-5718 H9
- H 10 Pressurizer Relief Valve Discharge (57) H7 H9 H 15 H 16 H 17 H 2O H 21 H 23 H 25 H 26 4.18-11
, --- - - , , , ~ .
.__