ML042790218
| ML042790218 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 10/04/2004 |
| From: | Tam P NRC/NRR/DLPM |
| To: | |
| References | |
| TAC MC2405 | |
| Download: ML042790218 (4) | |
Text
4.2 REACTIVITY CONTROL Applicability: Applies to the surveillance requirements for reactivity control.
Obiective: To verify the capability for controlling reactivity.
Specification:
A. SDM shall be verified:
- 1. Prior to each CORE ALTERATION, and
- 2. Once within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> following the first criticality following any CORE ALTERATION.
B. The control rod drive housing support system shall be inspected after reassembly.
C. The maximum scram insertion time of the control rods shall be demonstrated through measurement and, during single control rod scram time tests, the control rod drive pumps shall be isolated from the accumulators:
- 1. For all control rods prior to thermal power exceeding 40% power with reactor coolant pressure greater than 800 psig, following core alterations or after a reactor shutdown that is greater than 120 days.
- 2. For specifically affected individual control rods following maintenance on or modification to the control rod or control rod drive system which could affect the scram insertion time of those specific control rods in accordance with either "a" or "b" as follows:
a.1 Specifically affected individual control rods shall be scram time tested with the reactor depressurized and the scram insertion time from the fully withdrawn position to 90% insertion shall not exceed 2.2 seconds, and a.2 Specifically affected individual control rods shall be scram time tested at greater than 800 psig reactor coolant pressure prior to exceeding 40% power.
- b. Specifically affected individual control rods shall be scram time tested at greater than 800 psig reactor coolant pressure.
- 3. On a frequency of less than or equal to once per 180 days of cumulative power operation, for at least 20 control rods, on a rotating basis, with reactor coolant pressure greater than 800 psig.
D. Each partially or fully withdrawn control rod shall be exercised c; least once each week. This test shall be performed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in the event power operation is continuing with two or more inoperable control rods or in the event power operation is continuing with one filly or partially withdrawn rod which cannot be moved and for which control rod drive mechanism damage has not been ruled out. The surveillance need not be completed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> if the number of inoperable rods has been reduced to less than two and if it has been demonstrated that control rod drive mechanism collet housing failure is not the cause of an immovable control rod.
OYSTER CREEK 4.2-1 Amendment No: 178, 198, 249
BASIS:
Adequate SDM must be demonstrated to ensure that the reactor can be made subcritical from any initial operating condition. Adequate SDM is demonstrated by testing before or during the first startup after fuel movement, control rod replacement, or shuffling within the reactor pressure vessel. Control rod replacement refers to the decoupling and removal of a control rod from a core location, and subsequent replacement with a new control rod or a control rod from another core location. Since core reactivity will vary during the cycle as a function of fuel depletion and poison burnup, the beginning of cycle (BOC) test must also account for changes in core reactivity during the cycle. Therefore, to obtain the SDM, the initial measured value must be increased by an adder, "R", which is the difference between the calculated value of maximum core reactivity during the operating cycle and the calculated BOC core reactivity. If the value of R is negative (that is, BOC is the most reactive point in the cycle), no correction to the BOC measured value is required.
The SDM may be demonstrated during an in sequence control rod withdrawal, in which the highest worth control rod is analytically determined, or during local criticals. Local critical tests require the withdrawal of out of sequence control rods.
The frequency of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after reaching criticality is allowed to provide a reasonable amount of time to perform the required calculations and have appropriate verification.
During REFUEL MODE, adequate SDM is required to ensure that the reactor does not reach criticality during core alterations. An evaluation of each in vessel fuel movement during fuel loading (including shuffling fuel within the core) shall be performed to ensure adequate SDM is maintained during refueling. This evaluation can be a bounding analyses that demonstrate adequate SDM for the most reactive configurations during the refueling may be performed to demonstrate acceptability of the entire fuel movement sequence. For the SDM demonstrations that rely solely on calculation, additional margin must be added to the SDM limit of 0.38% delta k/k to account for uncertainties in the calculation.
The control rod drive housing support system (2)is not subject to deterioration during operation. However, reassembly must be assured following a partial or complete removal.
The scram insertion times for all control rods(3) will be determined at the time of each refueling outage. The scram times generated at each refueling outage when compared to scram times previously recorded gives a measurement of the functional effects of deterioration for each control rod drive. Scram time testing with the reactor depressurized is adequate to ensure that the control rod will perform its intended scram function during startup of the plant until scram time testing above 800 psig reactor coolant pressure is performed prior to exceeding 40% power.
OYSTER CREEK 4.2-3 Amendment No.: 75 , 124 , 17-8, 249
4.10 ECCS RELATED CORE LIMITS Applicability: Applies to the periodic measurement during power operation of core parameters related to ECCS performance.
Obiective: To assure that the limits of Section 3.10 are not being violated.
Specification:
A. Average Planar LHGR.
The APLHGR for each type of fuel as a function of average planar exposure shall be checked daily during reactor operation at greater than or equal to 25% rated thermal power.
B. Local LHGR.
The LHGR as a function of core height shall be checked daily during reactor operation at greater than or equal to 25% rated thermal power.
C. Minimum Critical Power Ratio (MCPR).
- 1. MCPR shall be checked daily during reactor operation or greater than or equal to 25% rated thermal power.
- 2. The MCPR operating limit shall be determined within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of completing scram time testing as required in Specification 4.2.C.
Bases:
The LHGR shall be checked daily to determine whether fuel burnup or control rod movement has caused changes in power distribution. Since changes due to burnup are slow, and only a few control rods are moved daily, a daily check of power distribution is adequate.
The minimum critical power ratio (MCPR) is unlikely to change significantly during steady state power operation so that 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is an acceptable fre-quency for surveillance. In the event of a single pump trip, 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> surveillance interval remains acceptable because the accompanying power reduction is much larger than the change in MAPLHGR limits for four loop operation at the corresponding lower steady state power level as compared to five loop operation. The 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> frequency is also acceptable for the APRM status check since neutron monitoring system failures are infrequent and a downscale failure of either an APRM or LPRM initiates a control rod withdrawal block, thus precluding the possibility of a control rod withdrawal error.
Because the transient analysis takes credit for conservatism in the scram speed performance, it must be demonstrated that the specific scram speed distribution is consistent with that used in the transient analysis. Surveillance 4.1 0.C.2 determines the actual scram speed distribution which is compared to the assumed distribution. The OYSTER CREEK 4.10-1 Amendment No.: 75, 249
MCPR operating limit is then determined based either on the applicable limit associated with scram times of Specification 3.2.B.3 or actual scram times. The MCPR operating limit must be determined once within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after each set of scram time tests required by Surveillance 4.2.C because the effective scram speed distribution may change during the cycle. The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> completion time is acceptable due to the relatively minor changes in scram speed expected during the operating cycle.
At core power levels less than or equal to 25% rated thermal power the reactor will be operating at or above the minimum recirculation pump speed.
For all designated control rod patterns which may be employed at this point, operating plant experience and thermal hydraulic analysis indicate that the resulting APLHGR, LHGR and MCPR values all have considerable margin to the limits of Specification 3.10. Consequently, monitoring of these quantities below 25% of the rated thermal power is not required.
OYSTER CREEK 4.10-2 Amendment No. 249