ML18149A334

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Proposed Tech Specs Re Operation W/Inoperable,Misaligned or Dropped Control Rod
ML18149A334
Person / Time
Site: Surry  Dominion icon.png
Issue date: 10/07/1986
From:
VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:
Shared Package
ML18149A333 List:
References
NUDOCS 8610140401
Download: ML18149A334 (20)
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Text

AITACHMENT 1 PROPOSED TECHNICAL SPECIFICATION CHANGE SURRY POWER STATION UNIT NOS. 1 AND 2 8610140401 861007 PDR ADOCK 05000280 P PDR

e TS 3.12-1 3.12 CONTROL ROD ASSEMBLIES AND POWER DISTRIBUTION LIMITS Applicability Applies to the operation of the control rod assemblies and power distri-bution limits.

Objective To ensure core subcriticality after a reactor trip, a limit on potential reactivity insertions from hypothetical control rod assembly ejection, and an acceptable core power distribution during power operation.

Specification A. Control Bank Insertion Limits

1. Whenever the reactor is critical, except for physics tests and control rod assembly exercises, the shutdown control rods shall be fully withdrawn.
2. Whenever the reactor is critical, except for physics tests and control rod assembly exercises, the full length control rod banks shall be inserted no further than the appropriate limit determined by core burnup shown on TS Figures 3 .12-lA or 3 .12-lB for three-loop operation and TS Figures 3 .12-4A or 3.12-4B for two-loop operation.
3. The limits shown on TS Figures 3.12-lA through 3.12-6 may be revised on the basis of physics calculations and physics data obtained during unit startup and subsequent operation, in accordance with the following:
a. The sequence of withdrawal of the controlling banks, when going from zero to 100% power, is A, B, C, D.
b. An overlap of control banks, consistent with physics cal-

e TS 3.12-8 AT and Overtemperature AT trip settings shall be reduced by the equivalent of 2% power for every 1% quadrant to average power tilt.

c. Inoperable Control Rods
1. A control rod assembly shall be considered inoperable if the assembly cannot be moved by the drive mechanism or the assembly remains misaligned from its group step demand position by more than +/-12 steps. Additionally, a full-length control rod shall be considered inoperable if its rod drop time is greater than 1.8 seconds to dashpot entry.
2. No more than one inoperable control rod assembly shall be permitted when the reactor is critical.
3. If more than one control rod assembly in a given bank is out of service because of a single failure external to the individual rod drive mechanism, i.e. programming circuitry, the provisions of Specifications 3.12.C.l and 3.12.C.2 shall not apply and the reactor may remain critical for a period not to exceed two hours provided immediate attention is directed toward making the necessary repairs. In the event the affected assemblies cannot be returned to service within this specified period the reactor will be brought to hot shutdown conditions.
4. The provisions of Specifications 3.12.C.l and 3.12.C.2 shall not apply during physics tests in which the assemblies are intentionally misaligned.
5. Power operation may continue with one rod inoperable provided that within one hour either:
a. the rod is no longer inoperable as defined in Specification 3,12.C.l, or

b.

  • TS 3.12-9 the rod is declared inoperable and the shutdown margin requirement of Specification 3.12.A.3.c is satisfied. Operation at power may then continue provided that:
1) either:

(a) power shall be reduced to less than 75% of rated power within one tl) hour, and the High Neutron Flux trip setpoint shall be reduced to less than or equal to 85% of rated power within the next -four (4) hours, or (b) the remainder of the rods in the group with the inoperable rod are aligned to within 1~ steps of the inoperable rod within one (1) hour while maintaining the rod sequence and insertion limits of Figure 3.12-1; the thermal power level shall be restricted pursuant to Specifica-tion 3.12.A during.subsequent operation.

2) the shutdown margin requirement of Specifica-tion 3.12.A.3.c is determined to be met within one hour and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> there-after.
3) the hot channel factors are shown to be within the design limits of Specification-3.12.B.1 within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
4) a reevaluation of each accident analysis of Table 3.12-1 is performed within 5 days. This reevaluation shall confirm that the previous analyzed results of these accidents remain valid for the duration of operation under these conditions.

e TS 3.12-10

6. If power has been reduced in accordance with Specifica-tion 3.12.C.5.b, power may be increased above 75% power provided that:

a) an analysis has been performed to determine the hot channel factors and the resulting allowable power level based on the limits of Specification 3.12.B.1, and b) an evaluation of the effects of operating at the increased power level on the accident analyses of Table 3.12-1 has been completed.

D. Core Quadrant Power Balance:

1. If the reactor is operating above 75% of rated power with one excore nuclear channel out of service, the core quadrant power balance shall be determined:
a. Once per day, and
b. After a change in power level greater than 10% or more than 30 inches of control rod motion.
2. The core quadrant power balance shall be determined by one of the following methods:
a. Movable detectors (at least two per quadrant)
b. Core exit thermocouples (at least four per quadrant)

E. Rod Position Indicator Channels

1. The rod position indication system shall be operable and capable of determining the control rod positions within +/-12 steps.
2. If a rod position indicator channel is out of service, then:
a. For operation above 50% of rated power, the position of the RCC shall be checked indirectly using core instrumentation (excore detectors and/or incore thermo-couples and/or movable incore detectors) at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and immediately after any motion of the non-indicating rod exceeding 24 steps, or

e TS 3.12-11

b. Reduce Power to less than 50% of rated power within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. During operations below 50% of rated power, no special monitoring is required.
3. If more than one rod position (RPI) indicator channel per group or two RPI channels per bank are inoperable, then the requirements of Specification 3.0.1 will be followed.

Basis The reactivity control concept assumed for operation is that reactivity changes accompanying changes in reactor power are compensated by control rod assembly motion. Reactivity changes associated with xenon, samarium, fuel depletion, and large changes in reactor coolant temperature (operating temperature to cold shutdown) are compensated for by changes in the soluble boron concentration. During power operation, the shutdown groups are fully withdrawn and control of power is by the control groups. A reactor trip occurring during power operation will place the reactor into the hot shutdown condition. The control rod assembly insertion limits provide for acheiving hot shutdown by reactor trip at any time, assuming the highest worth control rod assembly remains fully withdrawn, with sufficient margins to meet the assumptions used in the accident analysis. In addition, they provide a limit

e TS 3.12-16 be compensated for by tighter axial control. Four percent is the appropriate allowance for measurement uncertainty for F~H obtained from a full core map (3 38 thimbles, including a minimum of 2 detectors per core quadrant, monitored) taken with the movable incore detector flux mapping system. Measurement of the hot channel factors are required as part of startup physics tests, during each effective full power month of operation, and whenever abnormal power distribution conditions require a reduction of core power to a level based on measured hot channel factors. The incore map taken following core loading provides confirmation of the basic nuclear design bases including proper fuel loading patterns. The periodic incore mapping provides additional assurance that the nuclear design bases remain inviolate and identify operational anomalies which would, otherwise, affect these bases.

For normal operation, it ha~ been determined that, provided certain conditions are observed, the enthalpy rise hot channel factor F~H limit will be met.

These conditions are as follows:

1. Control rods in a single bank move together with no individual rod insertion differing by more than 15 inches from the bank demand position. An indicated misalignment limit of 13 steps precludes a rod misalignment no greater than 15 inches with consideration of maximum instrumentation error.
2. Control rod banks are sequenced with overlapping banks as shown in TS Figures 3.12-lA, 3.12-lB.
3. The full length control bank insertion limits are not violated.
4. Axial power distribution control procedures, which are given in terms of flux difference control and control bank insertion limits are observed. Flux difference refers to the difference

e TS TABLE 3.12-1 TABLE 3.12-1 ACCIDENT ANALYSES REQUIRING REEVALUATION IN THE EVENT OF AN INOPERABLE ROD Rod Cluster Control Assembly Insertion Characteristics Rod Cluster Control Assembly Misalignment*

Large and Small Break Loss of Coolant Accidents Single Reactor Coolant Pump Locked Rotor Major Secondary Pipe Rupture Rupture of a Control Rod Drive Mechanism Housing (Rod Cluster Control Assembly Ejection)

e TS FIGURE 3.12-2 DELETE

e TS FIGURE 3.12-3 DELETE

e e TS FIGURE 3.12-5 DELETE

e e TS FIGURE 3.12-6 DELETE

ATTACHMENT 2 SAFETY EVALUATION SURRY POWER STATION UNIT NOS. 1 AND 2

SAFETY EVALUATION Sections 3.12C and 3.12F of the Surry Technical Specifications establish limitations for operating the plant with an inoperable or bottomed Rod Cluster Control Assembly (RCCA). These limitations include special requirements for the evaluation of hot channel factors for both normal operating and accident (viz., rod ejection) conditions, as well as the imposition of a special set of control bank insertion _limits to be used under these conditions *. The intent of the alternate limits is to provide a measure of assurance that the requirements for shutdown margin, normal operation peaking factors and ejected rod worths and peaking factors will continue to be met during the period of operation with the inoperable or bottomed rod.

A disadvantage to the approach of providing alternate insertion limits for misaligned/inoperable rod situations is that the effects of rod misalignment are highly location and loading pattern dependent. It is therefore difficult to arrive at a set of revised limits which provide adequate *protection of the core loading pattern without unduly penalizing the operational flexibility of the plant.

More recent versions of Technical Specifications (such as the Westinghouse Standardized Technical Specifications, Revision 4) reflect a recognition of this difficulty by eliminating the alternate insertion limits and replacing*

them with a more detailed requirement for the timely evaluation of shutdown margin, peaking factors and the potential impact of inoperable rod conditions on the various accident analyses documented in the UFSAR.

.e Virginia Electric and Power Company is proposing an update to Sections 3.12C and 3.12F of the Surry* Technical Specifications which would eliminate the alternate rod insertion limits for both three-loop and two-loop operation (two-loop operation is prohibited by Section 3. 3 of the Technical Specifications). The limits have been replaced by requirements for evaluation of an inoperable rod which is similar to the requirements of the Standardized Technical Specifications except that a) provision is made for continued operation with a rod which is immobile (i.e., non-trippable) and b) provision is made to increase power above 75% of thermal power once the appropriate safety evaluations have been completed. These two provisions are presently permitted by the existing Technical Specifications, and therefore do not represent any relaxation in operating requirements. Also, some alteration in the sequencing of operator actions has been done to ensure that the operator directs immediate attention to performing those actions which should be completed first (e.g., the requirement to reduce power or realign the inoperable rod's group within one hour has been placed ahead of the requirement to reevaluate the accidents within 5 days).

The proposed Technical Specification change is consistent with actions previously taken during June 14, 1984 when Surry Unit No. 1 experienced a stuck rod cluster control assembly. Although not required by the existing Technical Specifications, various accident evaluations were performed to ensure that the Chapter 14 accident analysis was bounding. These accident evaluations are consistent with those in the proposed Te~hnical Specification change. NRC review of the 1984 event is documented in Inspection Report IEIR 84-21 dated August 17, 1984.

e e As part of the evaluation of the proposed Technical Specification change, a detailed review of the accident analyses in Chapter 14 of the Surry UFSAR has been performed. . It was determined that operation of the plant under the proposed revisions will ensure that none of the accident analyses are invalidated by the presence of an inoperable or bottomed rod.

10 CFR 50.59 Safety Review The proposed changes will not result in an unreviewed safety question as defined in 10 CFR 50.59. Specifically,

1. the revised specifications will not increase the probability of occurrence of any malfunction or accident addressed in the UFSAR or subsequent updates. No equipment modifications or design changes are involved, but only a revision to certain operational constraints.

The consequences of the Chapter 14 accidents will continue to be bounded by their associated analyses by virtue of the requirement to maintain the power peaking factors, shutdown margin and other significant safety parameters within appropriate design limits.

2. no new accident types or equipment malfunction scenarios will be introduced as a result of -operating in accordance with the revised specifications.
3. since the existing safety analyses wili remain bounding, there is no reduction in any safety margin.

10 CFR 50.92 Significant.Hazards Determination Pursuant to 10 CFR 50.92, the proposed change does not involve a significant hazards consideration because operation of Surry Unit Nos. 1 and 2 would not:

1. involve a significant increase in the probabi~ity or consequences of an accident. .previously evaluated. This change only. involves a revision to certain operational constraints. The accident probabilities will not change. The consequences of the Chapter 14 accidents will continue to be bounded by their associated analyses by virtue of the requirement to maintain the significant safety parameters within design limits.
2. create the possibility of a new or different kind of accident from.

any accident previously identified. It has been determined that a new or different kind of accident will not be possible due to this change. No equipment modifications or design changes are involved.

3. involve a*significant reduction in a margin of safety. The proposed revisions provide an even greater degree of assurance than the current specification that all of the acceptance criteria for the transient analyses presented in the UFSAR are met and the appropriate safety margins are maintained.

The Commission has provided examples of changes considered unlikely to involve significant hazards considerations. Example ii, which was published in the Federal Register, Volume 48, No. 6 7, April 6, 1983, page 14864 partially

  • states "a change that constitutes an additional limitation, restriction or control not presently included in the Technical Specifications: for example, a more stringent surveillance requirement". The proposed change is similar to this example. Therefqre, pursuant to 10 CFR 50. 92 based on_ the above consideration it has been determined that this change does not involve a significant hazards consideration.

e A'ITACHMENT 3 VOUCHER CHECK SURRY POWER STATION UNIT NOS. 1 AND 2

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D1sb. Vendor Bk Number Dare Number* Amount To The Order UNITED STATES NUCLEAR $150-00 Of REGULATORY COMMISSION IIIASHINGTON J>C 20555

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