ML18065B017

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Proposed Tech Specs,Providing Info Necessary for Completion of Review & Approval Process Re Writting Commitment Be Made That Description of Palisades Operating Requirements Manual Will Be Added to FSAR
ML18065B017
Person / Time
Site: Palisades Entergy icon.png
Issue date: 10/18/1996
From:
CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:
Shared Package
ML18065B016 List:
References
NUDOCS 9610300037
Download: ML18065B017 (10)
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Text

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ATTACHMENT 1 CONSUMERS POWER COMPANY PALISADES PLANT DOCKET 50-255 PENDING TECHNICAL SPECIFICATIONS CHANGE REQUESTS ADDITIONAL INFORMATION Revised Proposed Pages 8 Pages

-- -96io3ooo37 96io1s_________ --1 PDR ADOCK 05000255  ;

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~!SADES PLANT TECHNICAL SPECIFI~IONS TABLE OF CONTENTS SECTION DESCRIPTION PAGE NO 4.0 SURVEILLANCE REQUIREMENTS 4-1 4 .1 OVER PRESSURE PROTECTION SYSTEM TESTS 4-6 4.2 EQUIPMENT AND SAMPLING TESTS 4-7 Table 4.2.1 Minimum Frequencies for Sampling Tests 4-9 Table 4.2.2 Miflimum Frequencies for Equipment Tests 4-11 Table 4.2.3 Ventilation Systems Tests 4-14 4.3 SYSTEMS SURVEILLANCE 4-16 Table 4.3.1 Primary Coolant System Pressure Isolation Valves 4-18 4.4 Deleted 4-19 4.5 CONTAINMENT TESTS 4-19 4.5.1 Integrated Leakage Rate Tests 4-19 4.5.2 Local Leak Detection Tests 4-19 4.5.3 Containment Isolation Valves 4-21 4.5.4 Surveillance for Prestressing System 4-2la 4.5.5 End Anchorage Concrete Surveillance 4-2lc 4.5.6 Dome Delamination Surveillance 4-2lc 4.6 SAFETY INJECTION AND CONTAINMENT SPRAY SYSTEMS TESTS 4-24 4.6.1 Safety Injection System 4-24 4.6.2 Containment Spray System 4-24 4.6.3 Pumps 4-24 4.6.4 Valves 4-24 4.6.5 Containment Air Cooling System 4-25 4.7 EMERGENCY POWER SYSTEM PERIODIC TESTS 4-42 4.7.1 Diesel Generators 4-42 4.7.2 Station Batteries 4-42 4.7.3 Emergency Li~hting 4-43 4.8 MAIN STEAM STOP VALVES 4-44 4.9 AUXILIARY FEEDWATER SYSTEM 4-45 4.10 REACTIVITY ANOMALIES 4-46 4.11 Deleted 4-46 4.12 AUGMENTED ISI PROGRAM FOR HIGH ENERGY LINES 4-60 4.13 Deleted 4-65 4.14 AUGMENTED ISI PROGRAM FOR STEAM GENERATORS 4-66 4.15 PRIMARY SYSTEM FLOW MEASUREMENT 4-70 4.16 ISI PROGRAM FOR SHOCK SUPPRESSORS (Snubbers) 4-71 4.17 INSTRUMENTATION SYSTEMS TESTS 4-75 Table 4.17.1 Surveillance for the RPS 4-76 Table 4-17.2 Surveillance for ESF Functions 4-77 Table 4-17.3 Surveillance for Isolation Functions 4-78 Table 4-17.4 Surveillance for Accident Monitoring 4-79 Table 4-17.5 Surveillance for Alternate Shutdown 4-80 Table 4-17.6 Surveillance for Other Safety Functions 4-81 B4.17 Basis - Instrumentation Systems Surveillance B 4.17-1 4.18 POWER DISTRIBUTION INSTRUMENTATION 4-83 4.18.1 Incore Detectors 4-83 4.18.2 Excore Monitoring System 4-83 4.19 POWER DISTRIBUTION LIMITS 4-84 4.19.1 Linear Heat Rate 4-84 4.19.2 Radial Peaking Factors 4-84 4.20 MODERATOR TEMPERATURE COEFFICIENT (MTC) 4-85 iii Amendment No.

PAL~ES PLANT TECHNICAL SPECIFICATI~

TABLE OF CONTENTS SECTION 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 q-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 REPORTING REQUIREMENTS 6-19 6.6.1 Occupational Radiation Exposure Report 6-19 6.6.2 Radiological Environmental Operating Report 6-19 6.6.3 Radioactive Effluent Release Report 6-19 6.6.4 Monthly Operating Report 6-19 6.6.5 Core Operating Limits Report 6-20 6.6.6 Reserved 6-21 6.6.7 Accident Monitoring Instrument Report 6-21 6.6.8 Containment Structural Integrity Surveillance Report 6-22 6.6.9 Steam Generator Tube Surveillance Report 6-22 6.7 HIGH RADIATION AREA 6-23 iv Amendment No.

  • 4.3 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 - Deleted

f. A 100% volumetric examination of the regenerative heat exchanger primary side shell to tube-sheet welds and primary head, shall be performed at least once each 5 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 testing 1"1*'b' 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 prtor 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 I reactor shall not be made critical until the LPSI check valves (CK-3103, CK-3118, CK-3133 and CK-3148) have been verified closed.

1*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 . .§6., :/4, ~' ~'

4.5 CONTAINMENT TE~

.. 4.5.2 Local Leak Detection Tests (continued}

d. Test Frequency (l} Individual penetrations and containment isolation valves shall be leak rate tested at a frequency of at least every six months prior to the first postoperational integrated leak rate test and at a frequency of at least every refueling thereafter, not exceeding a two-year interval, except as specified in (a} and (b} below:

(a} The containment equipment hatch and the fuel transfer tube shall be tested at each refueling shutdown or after each time used, if that be sooner.

(b} A full air lock penetration test shall be performed at six-month intervals. During the period between the six-month tests when containment integrity is required, a reduced pressure test for the door seals or a full air lock penetration test shall be performed within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after either each air lock door opening or the first of a series of openings.

(2} Each three months the isolation valves must be stroked to the position required to fulfill their safety function unless it is established that such operation is not practical during plant operation. The latter valves shall be full-stroked during each cold shutdown.

I 4.5.3 Containment Isolation .Valves

a. The isolation valves shall be demonstrated operable by performance of a cycling test and verification of isolati-0n time for auto isolation valves prior to returning the valve to service after maintenance, repair or replacement work is performed on the valve or tts associated actuator, control or power circuit.
b. Each isolation valve shall be demonstrated operable by verifying that on each containment isolation right channel or left channel test signal, applicable isolation valves actuate to their required position during cold shutdown or at least once per refueling cycle.
c. The isolation time of each power operated or automatic valve shall be determined to be within its limit as specified in Table 3.6.1 when tested in accordance with Section XI of the ASME Boiler and Pressure Vessel Code.

4-21 Amendment No. ~, ~,

4.5 CONTAINMEN-STS

..* 1 4.5.4 Surveillance for Prestressinq System

a. Tendon inspection shall be accomplished at five-year intervals for the life of the plant. The scheduled inspection dates for all subsequent inspecti.ons may be varied by not more than plus or minus one year from the base schedule.
b. The surveillance tendons shall be randomly but representatively selected from each of the following groups:
1. A minimum of 4 dome tendons including one from each dome tendon group.
2. A minimum of 4 vertical tendons.
3. A minimum of 5 hoop tendons~

For each inspection, the tendons shall be selected on a random basis except that those tendons whose routing has been modified to clear penetrations shall be excluded from the sample.

c. During each tendon inspection, the following field testing shall be performed:
1. Lift-off readings shall be taken for each of the surveillance tendons. The tests shall include the following actions:

(a) One tendon, randomly selected from each group of tendons during each inspection, shall be subjected to essentially complete detensi.oning to identify broken or damaged wires.

(b) The simultaneous measurement of elongation and jacking force during retensioning shall be made at a minimum of three approximately equally spaced levels of force between the seating force and zero.

2. While the tendon is tn the detensioned state, each wire in the tendon will be checked for continuity.
3. Three wires, one from each of a vertical, a hoop and a dome tendon wfll be removed and identified for inspection. At each successive surveillance, the wires will be selected from different tendons. Each of the inspection wires removed will be visually inspected for evidence of corrosion or other deleterious effects and samples taken for laboratory testing.
4. The sheathing filler sha11 be inspected visually for color and coverage and samples shall be obtained for laboratory testing.
5. Tendon anchorage hardware such as bearing plates, stressing washers, shims and buttonheads shall be visually inspected for evidence of corrosi-0n or other deleterious effects.

4-21a Amendment No. -14, .a.l-, 66, +!, -l-99,

4.5 CONTAINMENT TE~

4.5.4 Surveillance for Prestressinq System (continued}

d. Following the field testing of 4.5.4c, the following laboratory testing shall be done:

I. Three tensile test specimens shall be cut from each of the three inspection wires removed (one from each end and one from the middle). One additional specimen shall be cut from the wire determined by field visual inspection to have the greatest amount of corrosion. Each of the wire samples shall

.be tested for ultimate strength, yield strength, and elongation.

2~ The sheathing fill~r samples shall be taken from each end of each tendon examined. Verti ca1 tendon samp 1es sha 11 be taken from the lower end. Samples shall be thoroughly mixed and analyzed for reserve alkalinity, water content, and con~entration of water soluble chlorides, nitrates, and sulfides. Analyses shall be performed in accordance with the procedures and within the acceptance limits specified in ASME Code Section X1, Table IWL-2525-1.

P~ocedures shall be established to minimize voids and to assure that the volume of sheathing filler removed has been replaced up.on comp 1et i.on of the inspection and amounts documented.

e. Acceptance ~riteria shall be as follows:

I. The average of all measured tendon forces for each type of tendon shall be equal to or greater than the minimum required prestress level, of 584 kips per tendon for dome tendons and, 615 kips per tendon for ho.op and vert ica.l tendons. The measured force i~ each individual tendon shall not be less than 95% of the predict~d force, or (a) the measured force in not more than one tendon is between 90% and 95% of the predicted force, and

(~) The measured forces in two tendons located adjacent to the tendon in (a) above are not less than 95% of the predicted forces, and (c) the meas.ured forces in all the remaining sample tendons

~re not less than 95% of the predicted force.

If measured force in any tendon is less than 90% of its predicted force, the tendon shall be completely detensioned and a determination shall be made as to the cause of such an occurrence and corrective actlon shall be taken. In addition, all such tendons shall have their forces measured as additional tendons in the next scheduled inspection period.

The Commi ss i.on sha.11 be notified in accordance with Paragraph 4.5.4f.

4.. 21b Amendment No. -14, -!&, -l-99,

CONTAINMEN~STS e i

4.5

' 4:5.4 Surveillance for Prestressinq System (continued)

2. Inspection wires shall indicate no significant loss of section by corrosion or pitting.
3. Tensile test specimens cut from inspection wires shall be tested for ult tmate strength. Failure at less than 11. 78 kips of any one of the test samples requires the Commission be notified in accordance with specification 4.5.4f.
4. Tendon anchorage hardware shall be free of significant corrosion, pitting, cracks or other deleterious effects.
f. If any element of the prestressing system fails to meet the acceptance criteria of 4.5.4e., the reporting provisions of 10 CFR 50.73 shall apply.

4.5.5 End Anchorage, Concrete Surve.i 11 ance

a. A VT-1 visual examination shall be performed on the end anchorage concrete surface at the surveillance tendon anchor points for signs of cracking., popouts, spa 11 i.ng, or corrosion. Concrete cracks having widths greater than 0.010 shall be evaluated and documented.
b. The end anchorage concrete surveillance inspection interval sha 11 be the same as tendon surveillance interval.
c. Acceptance criteria I. Crack widths sha 11 be measured by using' optical comparators or wire feeler gauge. Movements shall be measured by using demountable mechanical extensometers.
2. Concrete anchorage areas are acceptable if no concrete cracks are wider than 0.010 inches and no signs of new or progressive deterioration since the previous inspection are found.
3. Concrete surface conditions exceeding those stated in 4.5.5c.2 above shall be evaluated for the effect on tendon and containment structural integrity. The results of evaluation shall be included in the final surveillance report.

4.5.6 Dome Del ami nation Surveillance If, as a result of a prestresstng system inspection under Section 4.5.4, correct-i ve retens ion fog of five percent (8) or more of the total number of dome tendons is necessary to restore their liftoff forces to within the limits of Specification 4.5.4, a dome delamination inspection shall be performed with.in 90 days following such corrective re.tensioning. The results of this inspecti.on shall be reported to the NRC.

4-2lc Amendment No. -14, 6, +/-99,

6.0 ADMINISTRA~E CONTROLS 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 PUBLIC from radioactive effluents as low as reasonably achievable. The program (1) shall be contained in the 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:

1) 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,
2) Limitations on the concentrations of radioactive material released in liquid effluents to UNRESTRICTED AREAS conforming to 10 CFR 20, Appendix B, Table II, Column 2.
3) Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.106 and with the methodology and parameters in the ODCM,
4) 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 Appendix I to 10 CFR 50,
5) 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 CFR 20, Appendix B, Table II, Column l~
6) 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 Appendix I to 10 CFR 50,
7) 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 Appendix I to 10 CFR 50,
8) 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. ~'

6.b 6.5.5 ADMINISTRATIVE Reserved C~ROLS 6.5.6 Primary Coolant Pymp Flywheel Surveillance Program Surveillance of the primary coolant pump flywheels shall consist of a 100% volumetric inspection of the upper flywheels each refueling*.

6.5.7 Inservice Inspection and Testing Program This program provides controls for inservice inspection and testing of ASME Code Class I, 2, and 3 components including applicable supports.

The program shall include the following:

a. Testing frequencies specified in Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda (B&PV Code) as follows:

B&PV Code terminology Required interval for inservice testing for performing inservice activities testing activities Weekly s 7 days Monthly s 31 days Quarterly or every 3 months s 92 days Semiannually or every 6 months s 184 days Every 9 months s 276 days Yearly or annually s 366 days Biennially or every 2 years s 731 days

b. The provisions of Surveillance Requirement 4.0.2 are applicable to the above required intervals for performing inservice testing activities;
c. The provisions of Surveillance Requirement 4.0.3 are applicable to inservice testing activities; and
d. Nothing in the B&PV Code shall be construed to supersede the requirements of any Technical Specification.
  • The volumetric examination of the upper primary coolant pump flywheels is not required during the refueling outage at the end of cycle 12.

6-8 Amendment No.

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