ML18113A877
| ML18113A877 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 02/26/1979 |
| From: | Stallings C VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7902280286 | |
| Download: ML18113A877 (12) | |
Text
I VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 February 26, 1979 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation Attn:
Mr. A. Schwencer, Chief Operating Reactors Branch No.
Division of Operating Reactors U.S. Nuclear Regulatory Commission Washington, DC 20555
Dear Mr. Denton:
Serial No. 078 LQA/ESG:esh Docket Nos.
50-280 50-281 License Nos. DPR-32 DPR-37 AMENDMENT TO OPERATING LICENSES SURRY POWER STATION UN IT NOS.* 1 AND 2 PROPOSED TECHNICAL SPECIFICATION CHANGE NO. 76 Pursuant to 10 CFR 50.90, the Virginia Electric and Power Company hereby requests an amendment, in the form of changes to the Technical Speci-fications, to Operating Licenses DPR-32 and DPR-37 for the Surry Power Station*~
Unit Nos. 1 and 2.
The proposed changes are enclosed and have been designated as Change No. 76.
Presently, there are five part length control rod assemblies installed in each unit.
These assemblies were originally designed to provide the operator with the capability to position extra poison selectively in the core to control axial power distribution and to suppress xenon oscillations.
However, the NRC has not permitted the use of the part length rods.
The present Technical Speci-fications require that the part length rods be fully withdrawn from the core, except during physics tests.
Operating experience at Surry has shown that the part length rods are not required.
Consequently, we now plan to remove the part length rods from Surry Units 1 and 2 while locking the lead screws in the fully withdrawn position with anti-rotation devices.
Thimble plugs, or burnable poison rods (in future cycles),
will be inserted into the locations formerly occupied by the part length rods.
Based on an evaluation of this procedure, it has been determined that no thermal, hydraulic or mechanical problems wi 11 result. The FSAR does not take credit for the shutdown capability of the part length rods in any safety analysis.
Several important benefits will result from removal of the part length control rods.
As much as a full day reduction in outage time for each refueling is expected from the elimination of the necessity of unlatching and later relatch-ing the part length rods and lead screws.
In addition, once the rods have been removed, the seals at the top of the lead screw housings need never be opened \\
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VIRGINIA ELECTRIC AND POWER COMPANY TO Mr. Harold R. Denton 2
during refueling.
This virtually eliminates any possibility of leakage of these seals during plant startup following an outage, essentially reducing to zero the risk of significantly extending.the outage to cool down, depressurize, and repair a leak at the.seals.
Further, the elimination of the unlatching and relatching procedure will significantly reduce personnel radiation exposure during each refueling outage.
Removal of the part length rods is scheduled for the steam generator replacement outage on each unit.
We request, therefore, that this amendment be approved prior to the completion of the current outage on Unit No. 2.
Between such time as the amendment is approved, and the part 1 ength rods are removed from Unit No. 1 during its steam generator outage, we will continue the present prac-tice of maintaining the part length rods fully withdrawn during operation.
This proposed change has been reviewed and approved by the Station Nuclear Safety and Operating Committee and the System Nuclear Safety and Opera-ting Committee.
It has been determined that this request does not involve an unreviewed safety question.
We have evaluated this request in accordance with the criteria in 10 CFR 170.22.
Since the Technical Specifications now require the part length rods to be fully withdrawn during operation, and the FSAR does not take credit for the part length rods, the Staff should be able to conclude that removal of these rods does not involve a significant hazards consideration.
Accordingly, this request has been determined to be Class 111 for Unit 2.
The duplicate revis*ion for Unit 1 has been designated Class I.
A check in the amount of $4,400.00 is enclosed in payment of the amendment fees.
Enclosures:
Very truly yours,
. C. M. Stallings Vice President-Power Supply and Production Operations
- 1.
Proposed Technical'Specification Change No. 76
- 2.
Check in amount of $4,400.00 cc:
Mr. James P. O'Reilly, Director Office of Inspection and Enforcement Region I I
COMMONWEALTH OF VIRGINIA CITY OF RICHMOND
)
) s. s.
)
Before me, a Notary Public, in and for the City and Common-wealth aforesaid, today personally appeared C. M. Stallings, who being duly sworn, made oath and said (1) that he is Vice President-Power Supply and Production Operations, of the Virginia Electric and Power Company, (2) that he is duly authorized to execute and file the fore-going Amendment in behalf of that Company, and (3) that the statements in the Amendment are true to the best of his knowledge and belief.
Given under my hand and notarial seal this~ day of Ee/a:va r: v
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My Commission expires i2a>d v,a5y *.:z~. )"7¢, J (SEAL}
PROPOSED TECHNICAL SPECIFICATION CHANGE NO. 76
I.
culations and TS 3.12-2 ph~s data obtained.during Unit ~tup and subsequent operation, will be permitted.
- c.
The shutdown margin with all6wance for a stuck control rod assembly shall be greater than or equal to 1.77% reactivity under ~11 steady-state operation conditions,.except for physics tests, from ~era to full power, including effects of axial power distribution.
The shut-down margin as used here is defined as the amount by which the reactor 0
core would be subcritical at hot shutdown conditions (T
).547 F) avg-if all control rod assemblies were tripped, assuming that the highest worth control rod assembly remained fully withdrawn, and assuming no changes in xenon, or boron.
- 4.
Whenever the reactor is subcritical, except for physics tests, the critical rod position, i.e., the rod position at which criticality would be achieved if the control rod assemblies were withdrawn in normal sequence with no other reactivity changes, shall not be lower than the insertion limit for zero power.
- 5.
Deleted
- 6.
Insertion limits do not apply during physics tests or during periodic excerise of individual rods.
However, the shutdown margin indicated above must be maintained except for the low power physics test to measure control rod worth and shutdown margin.
For this test the reactor may be critical with all but one full length control rod, expected to have the highest worth, insertedo
- TS 3.12-9 of 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 3.12.C.l and 3.12.C.2 shall not apply during physics.tests\\in which the assemblies are intentionally misaligned.
- 5.
If an inoperable full-length rod is located below the 200 step level and is capable of being tripped, or if the full-length rod is located below the 30 step level whether or not it is capable of being tripped, then the insertion limits in TS Figure 3.12-2 apply.
- 6.
If an inoperable full-length rod cannot be located~ or if the inoperable full-length rod is located above the 30 step level*
and cannot be tripped, then the insertion limits in TS Figure 3.12-3 apply.
- 7.
Deleted
- 8.
If a full-length rod becomes inoperable and reactor operation is continued the potential ejected rod worth and associated transient power distribution peaking factors shall be determined by analysis within 30 days.
The analysis shall include due allowance for non-uniform fuel depletion in the neighborhood of the inoperable rod.
If the analysis results in a more limiting hypothetical transient than the cases reported in the safety analysis, the unit power level shall be reduced to an
- TS 3.12-10 analytically determined part power level which is consistent with the safety analysis.
D.
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.
E.
F.
- 1.
Once per day, and
- 2.
After a change in power level greater than 10% or more than 30 inches of control rod motion.
The core quadrant power balance shall be determined by one of the following methods:
- 1.
Movable detectors (at least two per quadrant)
- 2.
Core exit thermocouples (at least four per quadrant)
Inoperable Rod Position Indicator Channels
- 1.
If a rod position indicator channel is out of service then:
- a.
For operation between 50% and 100% of rated power, the position of the RCC shall be checked indirectly by core instrumentation (excore detector and/or thermocouples and/or movable incore detectors) every shift or subsequent to motion, of the non-indicating rod, exceeding 24 steps, whichever occurs first.
- b.
During operation below 50% of rated power no special moni-toring is required.
- 2.
Not more than one rod position indicator (RPI) channel per group nor two RPI channels per bank shall be permitted to be inoperable at any time.
Misaligned or Dropped Control Rod
- 1.
If the Rod Position Indicator Channel is functional and the associated full length control rod is more than
Basis
- TS 3.12-11 15 inches out of alignment with its bank and cannot be realigned, then unless the hot channel factors are shown to be within design limits as specified in ~ection 3.12.B.1 within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, power shall be reduced so as not to exceed 75% of permitted power.
- 2.
To increase power above 75% of rated power with a full length control rod more than 15 inches out of alignment with its bank an analysis shall first be made to determine the hot channel factors and the resulting allowable power level based on Section 3.12.B.
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 concen-tration.
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 achieving 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 on the maximum inserted rod worth in the unlikely event of a hypothetical assembly ejection, and provide for acceptable nuclear peaking factors.
The limit may be deter~
mined on the basis of unit startup and operating data to provide a more realistic limit which will allow for more flexibility in unit operation and I
still assure TS 3.12-12
- 1* e.
comp 1ance with the shutdown requirement.
he maximum shut-down margin requirement occurs at end of core life and is ~ased on the value used in the analysis of the hypothetical steam break accident.
The rod insertion limits are based on end of core life conditions.-
The shut-down margin for the entire cycle length is established at 1.77% reactivity.
All other accident analyses with the exception of the chemical and volume control system malfunction analysis are based on 1% reactivity shutdown margin.
Relative positions of control rod banks are determined by a specified control rod bank overlap.
This overlap is based on the consideration of axial power shape control.
The specified control rod insertion limits have been revised to limit the potential ejected rod worth in order to account for the effects of fuel dens i fica t ion..
The various control rod assemblies (shutdown banks, control banks A, B, C and D) ar-e-ea:ch to be moved as a bank, that is, with all assemblies.. in the bank within one step (5/8 inch) of~ the: bank pos*ition *., Position indication is provided by two methods:* -a,digital count of ac:t;:uating--puls~s:
which shows the demand position o.f,_the banks and a linear position indicator, Linear Variable Differential Transformer, which indicates th~
actual assembly position.
The position indication accura~y of the Linear Differential Transformer is approximately ~5% of span
(+7.5 inches) under steady state conditions.
The relative accuracy of the linear position indicator is such that, with the most adverse errors, an alarm is actuated if any two assemblies within a bank deviate by more than 14 inches.
In the event that the linear position indicator is not in service, the effects of
TS 3.12-13 malpositioned control rod assemblies are observable from nuclear and process information displayed in the :Main Control Room and by core thermocouples and in-core movable detectors.
Below 50% power, no special monitoring is required for malpositioned control rod assemblies with inoperable rod position indicators because, even with an unnoticed cqmplete assembly misalignment (full length control rod assembly 12 feet out of alignment with its bank) opera-tion at 50% steady state power does not result in*exceeding core limits.
The specified control rod assembly drop time is consistent with safety analyses that have been performed *
.An inoperable control rod assembly imposes additional demands on the operators.
The permissible number of inoperable control rod assemblies is limited to one in order to limit the magnitude of the operating burden, but such a failure would not prevent dropping of the operable control rod assemblies upon reactor trip.
Two criteria have been chosen as a design basis for fuel performance related to fission gas release, pellet temperature and cladding mechanical properties.
First, the peak value of fuel centerline. temperature must not exceed 4700°F.
Second, the minimum DNBR in the core must not be less than 1.30 in normal operation or in short term transients.*
In addition to the above, the peak linear power density, the nuclear enthalpy rise hot channel factor, and the hot assembly enthalpy rise factor must not exceed their limiting values which result from the large break loss of coolant accident analysis based on the ECCS acceptance criteria limit of 2200°F on peak clad temperature.
This is required to meet the initial conditions assumed for the loss of coolant accident.
To aid in specifying the limits on power distribution the following hot channel factors are defi-µed.
TS 3.12-16 For normal operation, it has been determined that, provided certain condi-tions are observed, the enthalpy rise hot channel factor, ~H, limit will be met; these conditions are as follows:
- 1.
Control rods in a single bank move together with no individual rod insertion differeing 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, and 3.12-2.
- 3.
The full length control bank insertion limits are not violated.
DELETED
- 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 between the top and bottom halves of two-section excore neutron detectors.
The flux difference is a measure of the axial offset which is defined as the difference in normalized power between the top and bottom halves of the core.
N The permitted relaxation in F~H with decreasing power level allows radial power shape changes with rod insertion to the insertion limits. It has been determined that provided the above conditions 1 through 4 are observed, this hot channel factor limit is met.
TS 5.3-2
- 3.
Reload fuel will be similar in design to the initial core.
The enrich-ment of reload fuel will not*exceed 3.60 weight percent of U-235.
- 4.
Burnable poison rods are incorporated in the initial core.
There are 816 poison rods in the form of 12 rod clusters, which are located in vacant control rod assembly guide thimbles.
The burnable poison rods consist of pyrex clad with stainless steel.
- 5.
There are 48 full-length control rod assemblies in the reactor core.
The full-length control rod assemblies contain a*l44-inch length of silver-indium-cadmium alloy clad with stainless steel.
- 6.
Surry Unit 1, Cycle 4, Surry Unit 2, Cycle 3, and subsequent cores will meet the following criteria at all times during the operation lifetime.
- a.
Hot channel factor limits as specified in Section 3.12 shall be met.