ML20037D661
| ML20037D661 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 08/19/1981 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20037D654 | List: |
| References | |
| NUDOCS 8108280058 | |
| Download: ML20037D661 (6) | |
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I REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continufd) i In addition to the requirements of Specification 4.0.5, each Reactor Coolant Pump flywheel shall be inspected per the recommendations of Regulatory Position C.4.b of Regulatory Guide 1.14, Revision 1, August 1975.
4.4.10.1.2 Augmented Inservice Inspection Program for Main Steam and Main Feedwater Piping - ine unencapsulatec welds greater than 4 incnes in nominal ciameter in the main steam and main feedwater piping runs loceted outside the containment and traversing safety related areas or located in compartments adjoining safety related areas shall be inspected per the following augmented inservice in-spection program using the applicaole rules, acceptance criteria, and repair procedures of the ASME Boiler and Pressure Vessel Code,Section XI,1974 Edition and Addenda through Summer 1976, for Class 2 components.
a.
System integrity and baseline data shall be established by performing a 100% volumetric examinat4cn of each weld prior to exceeding 18 months of operation.
b.
Ea~ch weld shall be examined in accordance with the above ASME Code requirements, except that 100% of the welds shall be examined, cumulatively, during each 10 year 'in-spection interval. The welds to be examined during each l
inspection period shall be selected to provide a repre-l sentative sample of the conditions of the welds.
If these examinations reveal unacceptable str.uctural defects in one or more welds, an additional 1/3 of the welds shall be examined and the inspection schedule for the repaired welds shall revert back to the first 10 year inspection program.
If additional unacceptable defects are detected in the second sampling, the remainder of the welds shall also be inspected.
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CALVERT CLIFFS - UNIT 2 3/4 4-29 0108280058 810819 PDR ADOCK 05000317 P
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hEACTORCOOLANTSYSTEM CORE BARREL MOVEMENT LIMITING CONDITION FOR OPERATION 3.4.11 Core barrel movement shall be limited to less than the Amplitude frocability Distribution (APD) and Spectral Analysis (SA) Alert Levels for the applicable THERMAL POWER level.
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APPLICABILITY: MODE 1.
l ACTION:
a.
With the APD and/or SA exceeding their applicable Alert Levels, POWER OPERATION may proceed provided the following actions are taken:
l.
APD shall be measured and processed at least once per
,, 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, 2.
SA shall be measured at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and shall be processed at least once per 7 days, and 3.
A Special Report, identifying the cause(s) for exceeding the applicable Alert' Level, shall be prepared and suo-mitted to tne Commission pursuant to Specification S.9.2 within 30 days of detection.
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b.
With the APD and/or SA exceeding incir applicaole Action Levels, measure and. process APD and SA data witnin 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to determine if the core barrel motion is exceeding its limits. With the core barrel motion exceecing its limits, reduce the core barrel motion to within its Action Levels within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in HOT STANDBY within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
c.
The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
fALVERT CLIFFS - UNIT 2 3/4 4-30 Amendment No. 39
REACTOR COOLANT SYSTEM SURVEILLANCE REOUIREMENTS 4.4.11 Routine Monitoring Core barrel movement shall be determined to be less than the APD and SA Alert Levels by using the excore neutron detectors to measure APD and SA at the following frequencies:
a.
APD data shall be measured and processed at least once per 7 days.
b.
SA data shall be measured and processed at least once per 31 days.
"I CALVERT CLIFFS - UNIT 2 3/4 4-31 Amendment No. 39
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REACTOR COOLANT SYSTEM 3/4.12 LETDOWN LINE EXCESS FLOW LIMITING CONDITION FOR OPERATION 3.4.12 The bypass valve for the excess flow check valve in the letdown line shall be closed.
APPLICABILITY: MODES 1, 2, 3 and 4.
ACTION:
With the above bypass valve cpen, restore the valve to it closed posi-tion within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE REQUIREMENTS 4.4.12 The bypass valve for the excess flew check valve in the letdown line shall ce determined closed within a hours prior to entering MODE 4 from MODE 5.
CALVERT CLIF'FS-UNIT 2 3/4 4-32 Amendment No. 5
_ REACTOR CCOLANT SYSTEM BASES The actual shift in RT of the vessel material will be established periodically durir.g operatiggTby removing and evaluating, in accordance with ASTM E185 73, reactor vessel material irradiation surveillance specimens installed near the inside wall of the reactor vessel in the core arer..
Since the neutron spectra at the irradiation samples and vessel inside radius are essentially identical, the measured transition shift for a sample can be applied with confidence to the adjacent section of the reactor vessel. The heatup and cooldown curves must be recalcu-lated when the ART,,
ferent from the ca7b determined from the surveillance capsule is dif-ulated ART for '.he equivalent capsele radiation NOT exposure.
The pressure-temperature limit lines shown on Figure 3.4-2 for reactor criticality and for inservice leak and hydrostatic testing have been provided to assure compliance with the minimum totaperature require-ments of Appendix G to 10 CFR 50.
The maximum RT for all reactor coolant % stem pressure-retaining materials, with the Nception of the reactor pressure vessel, has been g
determined to be 50 F.
The Lowest Service Temperature limit line shown on Figure 3.4-2 is based upon this RT (Sumner Addenda of 1972) of Section III of kk since Article NS-23324 A Vessel Ccde requires the Lcwest Service Temperature to be RT Selow this temperature, the shb. + 100*F for piping, pumps and valves.
em pressure must be limited to a maximum of 20% of the system's hydrostatic test pressure of 3125 psia.
The number of reactor vessel irradiation surveillance specimens and the frequencies for removing and testing these specimens are provided in i
Table 4.4-5 to assure compliance with the requirements of Appendix H l
to 10 CFR Part 50.
l The limitations imposed en the pressurizer heatup and cooldown rates ard spray water temperature differential are provided to assure that the pressurizer is operated within the design criteria assumed for the fati-l gue analysis performed in accordance with the ASME Code requirements.
The OPERASILITY of two PORVs or an RCS vent opening of greater than 1.3 square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs are 1 275*F.
Either PORV has adequate relieving capability to protect the RCS from overpressurization when the transient is limited to either (1) the start of an idle RCP with the secondary water temperature of the st.eam 4
measured by a surface contact instrument) generator 1 6 F (34*F when above the coolant temperature in the reactor vessel or (2) the start of a HPSI pump and its injection into a water solid RCS.
CALVERT CLIFFS-UNIT 2 B 3/4 4-11 Amendment No. 16 l
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REACTOR COOLANT SYSTEM BASES 3/4.4.10 STRUCTURAL INTEGRITY The inspection programs for the ASME Code Class 1, 2 and 3 components ensure that the structural integrity of these components will be maintained at an acceptable level throughout the life of the plant.
To the extent applicable, the inspection program for these components is in compliance with Section XI of the ASME Boiler and Pressure Vessel Code.
3/4.4.11 CORE BARREL MOVEMENT This specification is provided to ensure early detection of excessive core barrel movement if it should occur.
Core barrel movement will be detected by using four excore neutron detectors to obtoin Amplitude Probability Distribu-tion (APD) and Spectral Analysis (SA).
Baseline ccre barrel movement Alert Levels and Action Levels will be cc,firmeJd_duri.ng_each re;ctor. startup_
test program'following a core releao.
Data frcm these detectors is to be reduced in two forms.
Root mean square I
(RMS) values are computed from the APD of the signal amplitude.
These RMS magnitudes include variations due botn to various neutronic effects and internals motion.
Consequently, these signals alone can only provide a gross l
meas,ure of core barrel motion.
A more accurate assessment of core barrel l
motion is cotained from the Auto and Cross Power Spectral Oensities (PSD, XPSD),
l phase (c) and coherence (COH) of these signals.
These data result from the SA
'of the excore detector signals.
A modification to the required monitoring program may be justified by an
' analysis of the data obtained and by an examination of the affected parts during the plant shutdown at tne end of any fuel cycle.
l 3 / 4. a.12 LETCOWN LINE EXCESS FLOW This scecification is provided to ensure that the bypass valve for the I
excess flow check valve in the letdown line will be maintained closed during l
Plant operation.
This bypass valve is required to be closed to ensure that the effects of a pipe rupture downstream of this valve will not exceed the accident analyses assumptions.
l CALVERTCLI/FS-UNIT 2 B 3/4 4-12 Amendment No. 5,39
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