ML20002C801
| ML20002C801 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 07/27/1978 |
| From: | CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | |
| Shared Package | |
| ML20002C799 | List: |
| References | |
| NUDOCS 8101120032 | |
| Download: ML20002C801 (5) | |
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92 90 PRIMARY SYSTD1 SURVEILLANCE 91 APPLICABILITY Applies to preoperational and inservice structural surveillance of the reactor vessel and other Class 1, Class 2 and Class 3 system components.
-92 OBJECTIVE To insure the integrity of the Class 1, Class 2 and Class 3 piping systems and components.
93 SPECIFICATIONS a.-
The structural integrity of ASME Classes 1, 2 and 3 components, as determined by 10 CFR 50, Section 50.55a, shall be verified and maintained at an acceptable level in accordance with Section XI of the ASME B&PV Code with applicable addenda as required by 10 CFR 50, Section 50 5 5a(g), except where specific relief has been granted by the NRC.
I b.
Inservice testing of ASME Classes 1, 2 and 3 pumps and valves, as deter =ined by 10 CFR 50, Section 50 55a, shall be perfor=ed in accordance with Section XI of the ASME B&PV Code vith applicable addenda as required by 10 CFR 50, Section 50.55a(g), except where specific relief has been granted by the NRC, and where provisions i
of Sections 11.4.1.h, k.l.5 and 11.h.3.h'take precedence.
c.
Sufficient records of each inspection shall ce kept to allow compari-son and evaluation of future tests.
(See also Sections 6.9.4 and 6.10.2.g.)
d.
_The inservice inspection program shall be reevaluated as required by 10 CFR 50, Section 50.55a(g)(5) to consider incorporation of new in-spection techniques that have been proven practical, and the conclusions of the evaluation shall be used as appropriate to update the inspection program.
e.
A surveillance program to monitor radiation induced changes in the mechanical and bspact properties of the reactor vessel materials shall be maintained as described in Section h.1.l(h) of these Technical Specifications.
9.h BASIS T
The inspection program i=plementsSection XI of the ASME Boiler.and Pressure Vessel Code to the maximum extent practical. It is recognized that plant design and construction vere completed approximately seven years prior to the development'of Section XI and it is, thergfore, not possible to comply fully with the code.
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E 352. (Contd)
(b) Water addition to the containment sphere must be manually stopped before the accu =ulated water level reaches an elevation of 596 feet.
(c)
(Deleted) 3.6 CONTAIfB!ENT REQUIPSENTS Containment sphere integrity shall be maintained during power operation, refueling operation, shutdown and cold shutdown conditions except as specified by a system of procedures and controls to be established for occasions when containment must be breached during cold shutdown.
3.7 CONTAINMENT SPHERE LEAKAGE TESTING For the purpose of this specification, leakage rate is defined as the percent of the contained atmosphere (weight basis) which escapes per day (2h' hours) under the defined pressure conditions through any leaks in the containment boundary and all isolation valves and their asso-ciated piping.
The maximum allowable integrated leakage rate shall not exceed 0.5% per day of the containment atmosphere (weight basis) at the design pressure of 27 psig. The procedure for containment sphere leakage testing shall be:
(a) At least once every 6 months, the personnel lock, the equipment lock and the sphere supply-and-exhaust ventilation valves shall be pressurised, with air to 20 psig, to test their leak tight-ness. The sum of leakage rates from these valves and locks shall be less than 0.25%/ day'of the containment atmosphere (weight basis) at 20 psig.
(b) At least once each reactor refueling but in no case intervals greater than two years, the following valves shall be tested for operability from both the manual and automatic modes of operation and, at the same time, shall be tested for leak tightness by means of a pressure test utilizing air or the normal working fluid at a pressure not less than 20 psig:
Main Steam Isolation (M0-7050)
' Main Steam Drain (MO-7065)
Clean-Up Syste= Resin Sluice (CV-h091, CV h092, CV h093)
Reactor and Fuel Pit Drain Isolation (CV-h027, CV-kl17)
Reactor Enclosure Clean Sump Isolation (CV h031, CV h102)
Reactor Enclosure Dirty Su=p Isolation (CV-h025, CV-kl03)
- 0perability, automatic controls, and instrumentation tests required only if valve is opened for use during operation.
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'All'significant leaks (drcps/second) revealed by these tests shall require repair of valve seals and retests.
' Automatic controls and instrumentation associated with these
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Evalves shall be tested at approximately quarterly intervals; these tests may be conducted with a simulated. signal or in.
such other manner as to obviate plant shutdown.
(c) At least once each reactor shutdown for refueling but in no case intervals greater than two years, the following shall be visually examined for evidence of corrosion, cracking or deterioration:
All Electrical and Accessible Piping Penetration Nipple Welds.
-All Accessible Piping Welds to Nipples All Expansion Joints and Welds on Expansion Joints
. Potting Compound in All Electrical Penetrations Insulation at piping penetration welds shall be removed to 4
permit visual examination.
The probable 'cause of any significant corrosion, cracking or deterioration revealed by such visual examination shall be t
. determined, and evaluated in terms of likelihood of recurrence and probable effect upon other containment sphere penetration
. components. An individual component leak detection test shall be performed with air at 10 psig on the faulty component prior to its repair or modification. The faulty component, and other components'if necessary, shall be repaired or modified, and an individual component leak detection test performed with air at 10 psig upon each repaired.or modified component. All components so' repaired or modified shall be visually reexamined at appro-priate intervals, but not less frequently than once every six months, until the adequacy of annual visual inspection is
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reestablished to the operator's satisfaction.
i After. cutting into the sphere or its components, or any dis-assembly of components that would affect sphere integrity, an individual component leakage rate or an integrated leakage rate test, whichever is deemed more appropriate by the operator, shall be performed, with air at a pressure not less than 10 psic.
It shall be pennissible to employ a leak detection test in lieu.
of the above for insuring containment integrity following dis-assembly of the emergency condenser or the gasketed, bolted -
'closurs of the' coaxial cable electrical penetrations.
The individual component leakage rate determined from the above tests when: combined with the previously measured integrated i
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18 4.1. 2 - ( Contd )
periods of power operation. The following are absolute limits
- which, if exceeded, shall necessitate reactor shutdown. Corrective action vill necessarily be taken at more. stringent limits to mini-mize the possibility of these absolute limits ever being reached.
Conductivity (Micromno/cm)
Maximum 5
Maximum Transient
- 10 pH (Lower and Upper Limits) h.0 and 10.0 Chloride Ion-(Ppm) 1.0 Eqailibrium Halogen Radioactivity (pc/ml) 35 Boron (Ppm) 100
-(c) Leakage Limits 1.
If the primary coolant system leakage exceeds 1 gpm and the source of leakage is not identified, the reactor shall be
. placed in the hot shutdown condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, and cooldown to'a cold shutdown condition shall be initiated within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
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If leakage.from the primary ' coolant system exceeds 10 gpm, the reactor shall be placed in the hot shutdown condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, and cooldown to a cold shutdown condition shall be initiated within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
3 (Deleted) 4.1.3 Primary System Shielding b
Reactor shielding is ordinary concrete with a density of approximately 3
150 lb/ft. Thickness varies in plan and elevation to suit structural-requirements. The shielding thickness directly opposite the core shall.
be approximately 9 feet, 6 inches. The control rod drive room, which is directly beneath the reactor, has ordinary concrete walls which shall be approximately h feet thick.
A removable shield plug of a thickness k feet, 6-1/2 inches, consisting of k feet, 4 inches of concrete and 2-1/2 inches of lead, shall close the opening above the top of the reactor.
" Conductivity is expected to increase temporarily after start-ups from cold shutdown. The' maximum transient value here stated is the maximum permissible and applies only to the period subsequent to a cold shutdown between criti-cality and'2h hours after reaching 20% rated power.
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136 (NOTE: This is the new format of the Specifications to be issued in the future. Therefore, the numbering system may conflict with existing sections. Both are still applicable.)
Limiting Conditions for Operation Surveillance Requirement 3.15 REACTOR DEPRESSURIZATION SYSTEM h.1.5 REACTOR DFPRESSURIZATION SYSTD1 Applicability:
Applicability:
Applies to the operating status of the Reactor Applies to periodic testing requirements for Depressurization System (RDS).
the RDS.
Objective:
Objective:
To assure the operability of the BDS and when To verify operability of the RDS.
Working in conjunction with the emergency core cooling system to allow cooling of the reactor Specification:
fuel in the even of a Loss of Coolant Accident.
A.
The isolation valves shall be test-operated Specification:
at least once every three months.
A.
The RDS shall be operable at all power B.
The depressurizing valves sh'all be test-levels and when the reactor is critical operated during each cold shutdown; how-with the head on or when in hot shutdown ever, in the case of frequent cold shutdowns.
conditions.
these valves need not be exercised more often than once every three months per Section B.
The limiting conditions for operr. tion of IWV-3410 Summer 1973 Addenda of the ASME the instrumentation and actuating circuitry B&PV Code Section XI.
which initiates and controls the RDS are given in table 3.5 2.h.
C.
The instrumentation shall be functionally tested, calibrated and checked as indicated in Table h.5.2.h.
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