ML18067A470
| ML18067A470 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 03/28/1997 |
| From: | CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | |
| Shared Package | |
| ML18067A469 | List: |
| References | |
| NUDOCS 9704070095 | |
| Download: ML18067A470 (23) | |
Text
ATTACHMENT 1 CONSUMERS POWER COMPANY PALISADES PLANT DOCKET 50-255 TECHNICAL SPECIFICATION CHANGE REQUEST ADMINISTRATIVE CONTROLS REVISED PAGES FOR CHANGES STILL UNDER REVIEW Proposed Pages
4 *. 3' SYSTEMS SU~ILLANCE APPLICABILITY Applies to preoperational and inservice structural surveillance of the reactor vessel and other Class 1, Class 2 and Class 3 system components.
OBJECTIVE To insure the integrity of the Class 1, Class 2 and Class 3 piping systems and components.
SPECIFICATIONS a,b,c,d,e,f - Deleted
- g.
A surveillance program to monitor radiation induced changes in the
- mechanical and impact properties of the reactor vessel materials shall be maintained as described in Section 4.5.3 of the FSAR.
- h.
Periodic leakage testing1al.lbl on each check valve listed in Table 4.3.1 shall be accomplished prior to returning to the Power Operation Condition after every time the plant has been placed in the Refueling Shutdown Condition, or the Cold Shutdown Condition for more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> if such testing has not been accomplished within the previous 9 months, and prior to returning the check valves to service after maintenance, repair or replacement work is performed on the valves.
- i.
Whenever integrity of a pressure isolation valve listed in Table 4.3.1 cannot be demonstrated and credit is being taken for compliance with Specification 3.3.3.b, the integrity of the remaining check valve in each high pressure line having a leaking valve shall be determined and recorded daily and the position of the other closed valve located in that pressure line shall be recorded daily.
- j.
Following each use of the LPSI system for shutdown cooling, the reactor shall not be made critical until the LPSI check valves (CK-3103, CK-3118, CK-3133 and CK-3148) have been verified closed.
la>To satisfy ALARA requirements, leakage may be measured indirectly (as from the performance of pressure indicators) if supported by computations showing that the method is capable of demonstrating valve compliance with the leakage criteria.
~Reduced pressure testing is acceptable (see footnote 5, Table 4.3.1).
Minimum test differential pressure shall not be less than 150 psid.
4-16 Amendment No. ~' :/4, -l-d-0, ~' -l-74,
PAL~DES PLANT TECHNICAL SPECIFICAT~
TABLE OF CONTENTS
. SECTWN
- DESCRIPTION PAGE NO 5.0 DESIGN FEATURES 5-1 5.1 SITE 5-1 5.2 CONTAINMENT DESIGN FEATURES 5-1 5.2.1 Containment Structures 5-1 5.2.2 Penetrations 5-2 5.2.3 Containment Structure Cooling Systems 5-2 5.3 NUCLEAR STEAM SUPPLY SYSTEM (NSSS) 5-2 5.3.1 Primary Coolant System 5-2 5.3.2 Reactor Core and Control 5-3
. _5.3.3 Emergency Core Cooling System 5-3 5.4 FUEL STORAGE 5-4 5.4.1 New Fuel Storage 5-4 5.4.2 Spent Fuel Storage 5-4a Figure 5-1 Site Environment TLD Stations 5-5 6.0 ADMINISTRATIVE CONTROLS 6-1 6.1 RESPONSIBILITY 6-1 6.2 ORGANIZATION 6-1 6.2.1 Onsite and Offsite Organizations 6-1 6.2.2 Plant Staff 6-2 6.3 PLANT STAFF QUALIFICATIONS 6-3 6.4 PROCEDURES 6-4 6.5 PROGRAMS AND MANUALS 6-5 6.5.1 Offsite Dose Calculation Manual 6-5 6.5.2 Primary Coolant Sources Outside Containment 6-6 6.5.3 Post Accident Sampling Program 6-6 6.5.4 Radioactive Effluent Controls Program 6-7 6.5.5 Reserved 6-8 6.5.6 Primary Coolant Pump Flywheel Surv. Program 6-8 6.5.7 Inservice Inspection and Testing Program 6-8 6.5.8 Steam Generator Tube Surveillance Program 6-9 6.5.9 Secondary Water Chemistry Program 6-14 6.5.10 Ventilation Filter Testing Program 6-15 6.5.11 Reserved 6-16 6.5.12 Technical Specification Bases Control Program 6-16 6.5.13 Reserved 6-17 6.5.14 Containment Leak Rate Testing Program 6-17 6.5.15 Process Control Program 6-18 6.6 6.6.1 6.6.2 6.6.3 6.6.4 6.6.5 6.6.6 6.6.7 6.6.8 6.6.9 REPORTING REQUIREMENTS Occupational Radiation Exposure Report Radiological Environmental Operating Report Radioactive Effluent Release Report Monthly Operating Report Core Operating Limits Report Reserved Accident Monitoring Instrument Report Containment Structural Integrity Surveillance Steam Generator Tube Surveillance Report 6-19 6-19 6-19 6-19 6-19 6-20 6-21 6-21 Report 6-22 6-22 6-23 6.7 HIGH RADIATION AREA iv Amendment No. ~' -l-74,
6.0 ADMINISTRATIVE CONTROLS I
I I
~
\\
6.4 PROCEDURES 6.4.1 Written procedures shall be established, implemented, and maintained covering the activities referenced below:
- a.
The applicable procedures recommended in Appendix "A" of Regulatory Guide 1.33, Revision 2, Appendix A, February 1978.
- b.
Refue 1 i ng operations.
- c.
Surveillance and test activities of safety-related equipment.
- d.
Site Fire Protection Program implementation.
- e.
All programs specified in Specification 6.5.
- f.
Site Security Plan implementation.
- g.
Site Emergency Plan implementation.
6-4 Amendment No. -l-6, 31-, @, ~'.J.W.,.f-2-7., !M, -l-74,
6.0.
e ADMINISTRATIVE CONTROLS PROGRAMS AND MANUALS The following programs shall be established, implemented, and maintained:
6.5.1 Offsite Dose Calculation Manual (ODCM)
- a.
The ODCM shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm and trip setpoints, and in the conduct of the radiological environmental monitoring program; and
- b.
The ODCM shall also contain (1) the radioactive effluent controls and radiological environmental monitoring activities and (2) descriptions of the information that should be included in the Radiological Environmental Operating Report, and Radioactive Effluent Release Report required by Specification 6.6.2. and Specification 6.6.3.
- c.
Changes to ODCM:
- 1.
Shall be documented and records of reviews performed shall be retained. This documentation shall contain:
- a.
Sufficient information to support the change together with the appropriate analyses or evaluations justifying the changes, and
- b.
A determination that the change will maintain the level of radioactive effluent control required by 10 CFR 20.1302, 40 CFR 190, 10 CFR 50.36a, and 10 CFR 50, Appendix I, and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations.
- 2.
Shall become effective after approval by the plant superintendent.
- 3.
Shall be submitted to the NRC in the form of a complete, legible copy of the entire ODCM as a part of or concurrent with the Radioactive Effluent Release Report for the period of the report in which any change to the ODCM was made.
Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (e.g., month/year) the change was implemented.
6-5 Amendment No. -&£, -l-§4, -l-74,
6.0 ADMINISTRAT~ CONTROLS Radioactive Effluent Controls Program A program shall be provided conforming with 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to members of the public from radioactive effluents as low as reasonably achievable.
The program (1) shall be contained in the Offsite Dose Calculation Manual (ODCM), (2) shall be implemented by operating procedures, and (3) shall include remedial actions to be taken whenever the program limits are exceeded.
The program shall include the following elements:
- a.
Limitations on the operability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the
- ODCM,
- b.
Limitations on the concentrations of radioactive material released in liquid effluents to unrestricted areas conforming to 10 times the value in 10 CFR 20, Appendix B, Table 2, Column 2.
- c.
Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 1302 and with the methodology and parameters in the ODCM,
- d.
Limitation on the annual and quarterly doses or dose commitment to a member of the public from radioactive materials in liquid effluents released from each unit to unrestricted areas conforming to 10 CFR 50, Appendix I,
- e.
Limitations on the dose rate resulting from radioactive material released in gaseous effluents to areas beyond the site boundary conforming to the doses associated with 10 times the value listed in 10 CFR 20, Appendix B, Table 2, Column 1.
- f.
Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the site boundary conforming to 10 CFR 50, Appendix I,
- g.
Limitations on the annual and quarterly doses to a member of the public from Iodine-131, Iodine-133, tritium and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released from each unit to areas beyond the site boundary conforming to 10 CFR 50, Appendix I,
- h.
Limitations on the annual doses or dose commitment to any member of the public due to releases of radioactivity and to radiation from uranium fuel cycle sources conforming to 40 CFR 190.
6-7 Amendment No. -l-§4, +74,
ATTACHMENT 2 CONSUMERS POWER COMPANY PALISADES PLANT DOCKET 50-255 TECHNICAL SPECIFICATION CHANGE REQUEST ADMINISTRATIVE CONTROLS REVISED PAGES FOR CHANGES STILL UNDER REVIEW Existing Pages Marked to Show Proposed Changes
e e
PALISADES PLANT TECHNICAL SPECIFICATIONS TABLE OF CONTENTS
- . S'ECTION DESCRIPTION PAGE NO 5.0 DESIGN FEATURES 5-1 5.1 SITE 5-1 5.2 CONTAINMENT DESIGN FEATURES 5-1 5.2.1 Containment Structures 5-1 5.2.2 Penetrations 5-2 5.2.3 Containment Structure Cooling Systems 5-2 5.3 NUCLEAR STEAM SUPPLY SYSTEM (NSSS) 5-2 5.3.1 Primary Coolant System 5-2 5.3.2 Reactor Core and Control 5-3 5.3.3 Emergency Core Cooling System 5-3 5.4 FUEL STORAGE 5-4 5.4.1 New Fuel Storage 5-4 5.4.2 Spent Fuel Storage 5-4a Figure 5-1 Site Environment TLD Stations 5-5 6.0 ADMINISTRATIVE CONTROLS 6-1 6.1 RESPONSIBILITY 6-1 6.2 ORGANIZATION 6-1 6.2.1 Onsite and Offsite Organizations 6-1 6.2.2 Plant Staff 6-2 6.3 PLANT STAFF QUALIFICATIONS 6-3 6.4 PROCEDURES 6-4 6.5 PROGRAMS AND MANUALS 6-5 6.5.1 Offsite Dose Calculation Manual 6-5 6.5.2 Primary Coolant Sources Outside Containment 6-6 6.5.3 Post Accident Sampling Program 6-6 6.5.4 Radioactive Effluent Controls Program 6-7 6.5.5 Reserved 6-8 6.5.6 Primary Coolant Pump Flywheel Surv. Program 6-8 6.5.7 Inservice Inspection and Testing Program 6-8 6.5.8 Steam Generator Tube Surveillance Program 6-9 6.5.9 Secondary Water Chemistry Program 6-14 6.5.10 Ventilation Filter Testing Program 6-15 6.5.11 Reserved 6-16 6.5.12 Technical Specification Bases Control Program 6-16 6.5.13 Reserved 6-17 6.5.14 Containment Leak Rate Testing Program 6-17 6.5.15 Process Control Program 6-18 6.6 6.6.1 6.6.2 6.6.3 6.6.4 6.6.5 6.6.6 6.6.7 6.6.8 6.6.9 REPORTING REQUIREMENTS Occupational Radiation Exposure Report Radiological Environmental Operating Report Radioactive Effluent Release Report Monthly Operating Report Core Operating Limits Report Reserved Accident Monitoring Instrument Report Containment Structural Integrity Surveillance Steam Generator Tube Surveillance Report 6-19 6-19 6-19 6-19 6-19 6-20 6-21 6-21 Report 6-22 6-22 6-23 6 24 6.7 6.8 HIGH RADIATION AREA REVIEW AND AUDIT iv Amendment No. -l-6-2-, 174 October 31, 1996
.4.. 3 e
SYSTEMS SURVEILLANCE APPLICABILITY Applies to preoperational and inservice structural surveillance of the reactor vessel and other Class 1, Class 2 and Class 3 system components.
OBJECTIVE To insure the integrity of the Class 1, Class 2 and Class 3 piping systems and components.
SPECIFICATIONS a, b, c,d, e!ii!i - Deleted
- f.
A 100% volijmetric examiAatioA of the regeAerative heat exchaAger f)rimary side shell to t1:1be sheet welds aAd f)rimary head, shall be f)erformed at least oAce each § years.
- g.
A surveillance program to monitor radiation induced changes in the mechanical and impact properties of the reactor vessel materials shall be maintained as described in Section 4.5.3 of the FSAR.
- h.
Periodic leakage testing1al.lbl on each check valve listed in Table 4.3.1 shall be accomplished prior to returning to the Power Operation Condition after every time the plant has been placed in the Refueling Shutdown Condition, or the Cold Shutdown Condition for more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> if such testing has not been accomplished within the previous 9 months, and prior to returning the check valves to service after maintenance, repair or replacement work is performed on the valves.
- i.
Whenever integrity of a pressure isolation valve listed in Table 4.3.1 cannot be demonstrated and credit is being taken for compliance with Specification 3.3.3.b, the integrity of the remaining check valve in each high pressure line having a leaking valve shall be determined and recorded daily and the position of the other closed valve located in that pressure line shall be recorded daily.
- j.
Following each use of the LPSI system for shutdown cooling, the reactor shall not be made critical until the LPSI check valves
{CK-3103, CK-3118, CK-3133 and CK-3148) have been verified closed.
1a1To satisfy ALARA requirements, leakage may be measured indirectly
{as from the performance of pressure indicators) if supported by computations showing that the method is capable of demonstrating valve compliance with the leakage criteria.
1b1Reduced pressure testing is acceptable {see footnote 5, Table 4.3.1).
Minimum test differential pressure shall not be less than 150 psid.
4-16 Amendment No..§.a., ~' H-0, -142-, +74,
6.0
. ADMINISTRAT~ CONTROLS
. ; 6.* 4.:
PROCEDURES 6.4.1 Written procedures shall be established, implemented, and maintained covering the activities referenced below:
6.4.2 6.4.3
- a.
The applicable procedures recommended in Appendix 11A 11 of Regulatory Guide 1.33, Revision 2, Appendix A, February 1978.
- b.
Refueling operations.
- c.
Surveillance and test activities of safety-related equipment.
- d.
Site Fire Protection Program implementation.
- e.
All programs specified in Specification 6.5.
- f.
Site Security Plan implementation.
- g.
Site Emergency Plan implementation.
Procedijres aAd chaAges shall be approved prior to implemeAtatioA by the appropriate* seAior departmeAt maAager predesigAated by the PlaAt GeAeral MaAager Sijbject to the revie'dS per SpecificatioAs 6.8.1.6 aFld 6.8.3.
Temporary chaAges to procedijres of SpecificatioFI 6.4.l above may be made provided:
- a.
The iAteAt of the origiAal procedijre is Rot altered.
- b.
The chaAge is approved by two members (or desigAated alterF1ates) of the PRC, at least oAe of whom holds a SeF1ior Reactor Operator LiceAse.
- c.
The chaAge is docijmeF1ted, Sijbseqijefltly reviewed by PlaAt Safety aF1d LiceAsiAg withiFI 30 days of issijaAce aAd approved by the appropriate* seAior departmeAt maAager predesigAated by the PlaAt GeAeral MaAager.
The determtF1atioF1 of the appropriate seF1ior departmeAt maAager is based OA the activities addressed by the specific procedijre aF1d
'i'iil l be predes i gF1ated i A ~lri ti Fig by the Pl aF1t GeAeral MaAager.
6-4 Amendment No. +/-&, ~' e-9, +s, -I-Ga-, 2-7, 54, 74,
e
.6.. 0 ADMINISTRATIVE CONTROLS PROGRAMS AND MANUALS The following programs shall be established, implemented, and maintained:
6.5.1 Offsite Dose Calculation Manual (ODCM)
- a.
The ODCM shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm and trip setpoints, and in the conduct of the radiological environmental monitoring program; and
- b.
The ODCM shall also contain (1) the radioactive effluent controls and radiological environmental monitoring activities and (2) descriptions of the information that should be included in the Radiological Environmental Operating Report, and Radioactive Effluent Release Report required by Specification 6.6.2. and Specification 6.6.3.
- c.
Changes to ODCM:
- 1.
Shall be documented and records of reviews performed shall be retained. This documentation shall contain:
- a.
Sufficient information to support the change together with the appropriate analyses or evaluations justifying the changes, and
- b.
A determination that the change will maintain the level of radi Q.~.c:.tJY.~ effluent control required by 10 CFR 20.106 !ttll&S.:0.:2, 40 CFR 190, 10 CFR 50.36a, and 10 CFR 50, Appendi'x=*=*=*=r;*=*=*=*=*and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations.
- 2.
Shall become effective after approval by the plant superintendent.
- 3.
Shall be submitted to the NRC in the form of a complete, legible copy of the entire ODCM as a part of or concurrent with the Radioactive Effluent Release Report for the period of the report in which any change to the ODCM was made.
Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (e.g., month/year) the change was implemented.
6-5 Amendment No. 8-£, ~, -l-74,
6.0 ADMINISTRATIVE ITROLS
.6~5.4* ** Radioactive Effluent Controls Program A program shall be provided conforming with 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to members of the publtc from radioactive effluents as low as reasonably achievable.
The program (1) shall be contained in the Offsite Dose Calculation Manual (ODCM), (2) shall be implemented by operating procedures, and (3) shall include remedial actions to be taken whenever the program limits are exceeded.
The program shall include the following elements:
- a.
Limitations on the operability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the
- ODCM,
- b.
Limitations on the concentrations of radioactive material released J:i~:~::i:i~~i~::li:i:il~e~}~ ~~, u~~~!~~~~t~~ ~~~~! ~n:i~r~~~~m~
0 2j~jjglll!)jffilji
- c.
Monitoring, sampling, and analysis of radioactive li.qyJd and gaseous effluents in accordance with 10 CFR 20.-!% :l:S.!QZ. and with the methodology and parameters in the ODCM,
- d.
Limitation on the annual and quarterly doses or dose commitment to a member of the public from radioactive materials in liquid effluents released from each unit to unrestricted areas conforming to 10 CFR 50, Appendix I,
- e.
Limitations on the dose rate resulting from radioactive material rel eased in gaseous effluents to areas b~Y.<?..1'!.<L.t.b.~.... S..i..t.~.... ~<?.Y.1.1.ct.<l.r.Y......
- ~ii:i~r~~~g 2~~ l~~e~~~~s B ~s~~~ ~ !t~ i~ t~o:f:~AA~:1:"~~lli8n!:t¥!l~:::g:!:::::::~rn::~:E::il:
- f.
Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the site boundary conforming to 10 CFR 50, Appendix I,
- g.
Limitations on the annual and quarterly doses to a member of the public from Iodine-131, Iodine-133, tritium and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released from each unit to areas beyond the site boundary conforming to 10 CFR 50, Appendix I,
- h.
Limitations on the annual doses or dose commitment to any member of the public due to releases of radioactivity and to radiation from uranium fuel cycle sources conforming to 40 CFR 190.
6-7 Amendment No. -l-54,.f-74,
J
1e L -
I e*
1e
-~* '
l,
ATTACHMENT 3 CONSUMERS POWER COMPANY PALISADES PLANT DOCKET 50-255 TECHNICAL SPECIFICATION CHANGE REQUEST ADMINISTRATIVE CONTROLS REVISED PAGES FOR CHANGES STILL UNDER REVIEW Table of CPC-2A Locations for Relocated Material
Table of CPC-2A Locations for Relocated Material Technical Specifications CPC-2A Description 6.4.2 Policy6.1 Procedure Approval 6.4.3 Appendix A, item 2h Temporary Changes 6.8.1 Appendix 8 Requirements for PRC 6.8.1.1 81 PRC Functions 6.8.1.2 82 PRC Composition 6.8.1.2 83 PRC Alternates 6.8.1.4 84 PRC Meeting Frequency 6.8.1.5 85 PRC Quorum 6.8.1.6 86 PRC Review Responsibilities 6.8.1.6a 86.a & 86.f Procedures, Programs, & Etc 6.8.1.6b - e 86.b-e respectively Review Items 6.8.1.6f 86.g Reportable Events 6.8.1.6g 86.h Significant Items 6.8.1.6h 86.1 Review Group Reports 6.8.1.6i 86.j Operating Experience 6.8.1.6 final paragraph 86 final paragraph Routing Review 6.8.1.7 87 PRC Authority 6.8.1.7a 87.a Recommend to Manager 6.8.1.7b 87.b Determine USQs 6,8.1.7c 87.c Notify VP of Disagreements 6.8.1.7 second paragraph 87 second paragraph Approval of Routed Items 6.8.1.7 third paragraph 87 third paragraph Routed Items to Meeting 6.8.1.8 88 PRC Meeting Records 6.8.2 AppendixC Independent Review 1
1..
Table of CPC-2A Locations for Relocated Material Technical Specifications CPC-2A Description 6.8.2.1 C1 ISRG Function 6.8.2.1a-k C1.a - k respectively ISRG Review Items 6.8.2.2 C2 ISRG Composition 6.8.2.3 C3 Use of Consultants 6.8.2.4 C4 ISRG Responsibilities 6.8.2.4.1 C4.1 ISRG Review 6.8.2.4.1 a - j C4.1 a - j respectively ISRG Review Items 6.8.2.4.2 01 Audits 6.8.2.4.2a - I 01.a - I respectively Required Audits 6.8.2.4.2 final paragraph 01 final paragraph Managers to Receive Audits 6.8.2.4.3 C4.1 final paragraph NPAD Reviewer Qualification 6.8.2.5 cs ISRG Authority 6.8.2.6 C6 ISRG Records 6.8.2.6a C6a Results of Reviews to VP 6.8.2.6b C7.b Annual Report to Managers 6.8.3 89 Design Review Group 6.8.3,1 89 Review Group Function 6.8.3.2 89.1 Review Group Qualifications 6.8.3.3 89.2 Reviews for PRC 6.8.3.3a-d 89.2a - d respectively Review Items 6.8.3.4 89.3 Significant Items to PRC 6.8.3.5 89.4 Review Group Reports 2