ML20107D225
| ML20107D225 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 02/11/1985 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20107D228 | List: |
| References | |
| NUDOCS 8502220277 | |
| Download: ML20107D225 (2) | |
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4.
SURVEILLANCE STANDARDS t
t-During Reactor Operational Conditions for which a Limiting Condition for
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Operation does not require a system / component to be operable, the associated
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surveillance requirements do not have to be performed. Prior to declaring a system / component operable, the associated surveillance requirement must be current. The above appitcability requirements assure the operability of systems / components for all Reactor Operating Conditions when required by the Limiting Conditions for Operation.
4.1 CPERATIONAL SAFETY IEVIEW Acclicability Applies to items directly related to safety limits and limiting conditions for opeintion.
Otiective To specify the, minimum frequency and type of surveillance to be applied to i
unit equipment,and conditions.
Soecification 4.'1.1 The minimum frequency and type of surveillance required for reactor protection system and engineered safety feature protection system instrumentation when the reactor is critical shall De as stated in Taole 4.1-1.
4.1.2 Equipment and sampling test shall be performed as detailed in Tables 4.1-2,and 4.1-3.
4.1.3 Each post accident monitoring intrumentation channel shall be -
i demonstrated OPERABLE by the performance of the check, test and calibration at _the frequencies shown in Table 4.1-4.
-e Bases
~ Check Failtres such as blown instrument fuses, defective indicators, or faulted amplifiers which result in " upscale" or "downscale" indication can be easily recognized by sisple observation of the functioning of an instrument or system. Furthermore, such failures are, in many cases, revealed by alarm or annuciator action. Comparison of output and/or state of independent channels measuring the same variable stpplements this type of built-in surveillance.
Based on experietta in operation-of both conventional and nuclear systems, when the unit is in o p. tion, the minimum checking frequency stated is deemed-adequate for reactor system instrumentation.
Calibration Calibration shall be performed to asstre the present;ation and acquisition of accurate information. The nuclear flux (power range) channels amplifiers shall be checked and calibrated if necessary, every shift against a heat Dalance standard. The frequency of heat palance checks will assure that the difference between the out-of-core instrumentation and the heat' Dalance remains less than 45.
4-1 Amendment No. #$. pp,100 (Corrected February 11, 1985) 1 8502220277 850211 PDR ADOCK OS000299 l
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4.2.5 1he licewee shall submit a report or application for license e and-ment to the NRC within 90 days after any time 'that Crystal River Unit No'. 3 fatis to maintain a cumulative reactor utilization factor of at least 65%.
The report shall provide justification for continued operation of 1MI-1 with the reactor vessel surveillance program conducted at Crystal River Unit No. 3, or the application for license amendment shall propose an alternate program for conduct of the TMI-1 reactor vessel surveillance program.
For the purpose of this technical specification, the definition of commercial operation is that given in Regulatory Guide 1.16. Revision 4.
The definition of cumulative reactor utilization factor is:
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Cumulative reactor utilization factor - (Cisn'ulative megawatt hours (thermal) since attainment of comercial operation at 100% power x (100)) divided by (licensed power (PWt) x (Cumulative hours since attainment of connercial operation at 100% power)).
I 4.2.6 In addition to the mports required by Specification 4.2.4, a report j
shall be sthmitted to the NRC prior to September 1,1982, which sunmarizes the first five years of operating experience with the TMI-1 t
integrated surveillance program performed at a host reactor. If, at the _ time of submission of this report, it is desired to continue the surveillance program at a host reactor, such continuation shall be justified on the basis of the attained operating experience.
4.2.7 A surveillance program for the pressure isolation valves between the primary coolant system and the low pressure injection system shall be as follows:
1.
Periodic leakage testing (a) at test differential pressure greater l
than 150 psid shall be accomplished for the valves listed in Table i-3.1.6.1 for the following conditions:
i (a) prior to achieving hot shutdown after returning the valve to service following maintenance repair or mplacement work, and 8
(b) prior to achieving hot shutdown following a cold shutdown of greater than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> duration unless testing has been per-
. formed within the pmvious 9 months.
2.
Whenever integrity of a pressure isolation valve listed in Table 3.1.6.1 cannot be demonstrated, the integrity of the other remainine valve in each high pressure line having a leaking valve shall be determined and recorded daily. In addition, the position of one i[
other valve located in the high pressure piping shall be recorded l
daily.
i Bases j'q a.-
Specifications 4.2.1 and 2 ensure that inservice inspection of ASME Code Class 1, 2 and 3 components and inservice testing of ASME Code Class 1, 8-2 and 3 pumps and valves will be performed in accordance with a periodi-r
- cally updated version of Section II of the ASME Boiler and Pressure vessel Code and Addenda as required by 10 CFR 50.55a(s). Relief from any of the above requirements has been provided in writing by the NRC and is not a part of these technical specificatipos..
WTo satisfy ALARA requirements, leakage may be measured indirectly (as from the per-fonnance of pressure iridicators)'if accomplished in accordance with approved pro-cedures and supported by computations showing that the method is capable of demon-strating valve compliance with the leakage criteria.
Amenhent No. X, WM. Onder 4H: 4AMWe7 4-12 71 (Corrected February 11, 1985) r
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