ML19253A173
| ML19253A173 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 07/25/1979 |
| From: | Schwencer A Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19253A174 | List: |
| References | |
| NUDOCS 7908200019 | |
| Download: ML19253A173 (13) | |
Text
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i NUCLEAR REGULATORY COMMISSION E
WASHINGTON, D. C. 2 455 a
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VIRGINIA ELECTRIC AND POWER COMPANY DOCKET NO. 50-280 SURRY POWER STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE P.:ndment No. 50 License No. DPR-32 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Virginia Electric and Power Company (the licensee) dated February 26, 1979, cmiplies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in confomity with the application, the provisions of the act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and, E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
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. 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to the license amendment, and paragraph 3.B of Facility Operating License No. DPR-32 is hereby amended to read as follows:
B.
Technical Specifications The Technical Snecifications contained in Appendices A and B, as revised through Amencent No. 50
, are her eby incorporated in the license.
The licensee sh:ll operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
.ubA A. Schwencer,' Chief Operating Reactors Branch #1
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Division of Operating Reactors
Attachment:
Changes to the Technical Specifications Date of Issuance: July 25,1979 803 30%
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VIRGINIA ELECTRIC AND POWER COMPANY DOCKET N0. 50-281 SURRY POWER STATION, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 49 License No. DPR-37 1.
The Nuclear Regulatory Comn'ssion (the Commiss an) has found that:
A.
The application for ar endment by Virginia Electric and Power Company (the licenseel dated February 26, 1979, complies with the standards and "_ct-ecents of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will 0;erate in confomity with the application, the provisions of the act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized b.y this amendment can be conducted without endangering the health and safety of the public, and (11) that such activities will be conducted in compliaIce with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and, E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
803 309
. 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to the license amendment, and paragraph 3.B of Facility Operating License No. DPR-37 is hereby amended to read as follows:
B.
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 49, are hereby incorporated in the license. The licensee shall operate the facility in acconf ance with the Technical Specifications.
3.
This license anendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION 44tb4(
t A. Schwencer, Chief Operating Reactors Branch #1 Division of Operating Reactors
Attachment:
Changes to the Technical Specifications Date of Issuance: July 25, 1979 803 310
ATTACHMENT TO LICENSE AMENDMENT NOS.50 AND 49 FACILITY OPERATING LICENSE NOS. DPR-32 AND DPR-37 DOCKET NOS. 50-280 AND 50-281 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are '.dentified by amendnent nunber and contain vertical lines indicating the area of change.
Renove Insert 3.12-2 3.12-2 3.12-9 3.12-9 3.12-10 3.12-10 3.12-11 3.12-11 3.12-12 3.12-12 3.12-13 3.12-13 3.12-16 3.12-16 6.3-2 5.3-2 803 3;y
TS 3.12-2 culations and physics data obtained during. Unit Startup and subsequent operation, will be permitted.
c.
The shutdown cargin with allowance for a stuck control rod assembly shall be grecter than or equal to 1.77% reactivity under all steady-State cperation :onditions, neept f o r p'.iy s i c s tests, fr m.
cero to full power, including ef fects of axial power distribution.
The shut-down margin as used here is defined as the amount by which the reactor core would be suberitical at hot shutdown conditions (T 5547 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 suberitical, 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 reamtivity changes, shr.ll not be lower than the insertion limit fer zero power.
5.
Deleted 6.
Insert;on limits do not apply durir.g physics tests or during periodic exercise 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 f
worth, inserted.
D000 07:P"ip Amendment No. 50 Unit 1 I UWs ladQidf 3l2 Amendment No.49 Unit 2 s
TS 3.12-9 of 3.12.C.1 and 3.12.C.2 shall not apply and the reactor may t enain critical f or a period not to exceed two hours provided inmedic.ta at:entien s dirc:ej toward.aiing the nacaseary repairs.
In the event the affected assemblies cannot be returned to service within this specified period the reacter will be brought to hot shutdown conditions.
4 The provisions of 3.12.C.1 and 3.12.C.2 shall not apply during physics tests in which the assemblies are intentionally misaligned.
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5.
If an inoperable full-length rod is located belou the 200 step level and is capable of being tripped, or if the f all-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-;
- apply, 6.
If
.1 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 beco=es 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 saf ety analysis, the unit power level shall be reduced to an Amendment No. 50, Unit 1
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TS 3.12-10 analytically det.amined pa rt power level which is consistent with the safety analysis.
D.
If the rcactar is opera- 'T; above 73% of rated ; c Jer 11th one excora nuclear channel out of service, the core quadrant power balance shall be determined.
1.
Once per day, and 2.
Af ter a chan;;e in power level greater than 10% or raore 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 cuadrant) 2.
Core exit thermocouples (at least four per quadrant)
E.
Incoerable Red Positic a Indicator Channels 1.
If a rod position indicator channel is out of service then:
a.
For operation between 507. and 100*: of, rated power, the position of the RCC shall be checked indirectly by core ins trumenta t ion (excore decector and/or thermocouples and/or movable incore detectors) every shif t or subsequent to motion, of the non-indicating rod, exceeding 24 steps, whichever occurs first.
b.
During operation below 507. 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.
I old{ dy,j@jy F.
Misal1:ned or Droceed Control Rod r,
1.
If the Rod Position Indicator Channel is functional and the associated full length control rod is more than 803 314 Amendment No. 50, Unit 1 Amendment No. 49, Unit 2
TS 3.12-11 15 inches out of alignment with its bank and cannot be realigned, then unless the hot channel factors are shown M be within design limits as specified in Section 3.12.3.1 within S neurs, power shall be reduced so as not to exceed 75'.' of permitted power.
2.
To inc rease power above 75'; of rated power uith a full length control rod more than 15 inches out of alignment with its bank an analysis shall first be =ade to determine the hot channel factors and tha resulting allowable power level based on Section 3.12.B.
Basis The reactivity control corcept assumed fcr operation is that reactivity changes accompanying changes in reactor pouer are compensated by centtol rod assembly motion. Reactivity changes associated with xenon, samarium, fuel deplation, and large changes in reactor coolant renperature (operating temperature to cold shutdown) are compensated for by changes in t!ie 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 reacter 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 assumptiens used in the accident analysis.
In addition, they provide a limit on the maxi =um inserted rod worth in the unlikely event of a hypot' 2tical assembly ejection, and provide for acceptable nuclear peaking factors. The limit may be deter-
=ined on the basis of unit startup and operating data to provide a more realistic limit which will allow for =cre flexibility in unit operation and hb[4f Ob jfj bh Us ydll,ty[p,,
Amendment No. 50, Unit I Amendment No. 49, Unit 2 m
TS 3.12-12 still assure compliance with the shutdown requirement. The maximu= shut-down ear;;in requirement occurs at end of core life and is b~ased on the value used in the analysis of the hypothetical steam break accident. The rod insertien limits are based on end of core life conditions.
The shut-crar n in for the untire cycle 1.,; t h is establishd ct 1.77: : cacti.ity.
m All other accident analyses with the exception of the chemical and vclu=e control system malfunction analysis are based on 1 reactivity shutdown margir.
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 nave been revised to limi: the pctential ejected rod worth in order to account for the effects of fuel densification.
The various control rod assemblies (shutdewn banks, centrol banks A, 3, C and D) are each to be moved as a bank, that is, with all assemblies in the bank within one step (5/3 inch) of the bank position. Position indication is provided by two methods: a digital count of actuatin;; pu'ses which shows the de=and position of the banks and a linear position indicator, Linear Variable Diff erential Transformer, which indicates the actual assembly position. The position indication accuracy of the Linear Differential Transformer is approximately +5% of span (3 5 inches) under steady state cenditions. The relative accuracy of 7
the linear position indicator is such that, with the most adverse errors, an alar is actuated if any two assemblies within a bank deviate by : ore than 14 inches.
In the event that the linear position indicator is not in service, the effects of Amendtront No. 50, Unit 1 i)t F"
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h [j;[;djjj}d Amendment No.49, Unit 2 I
803 316
TS 3.12-13 nalpesitioned centrol red assemblies are cbservable from nuclear and process core thercoccuples and inferration displayed in the Main Centrol Rocn and by Belou SCT' pcwer, ne special tenitoring is required in-core revable detecters.
for ca:positiened control rod assenblics with inoperable red position indicators tn 1, : 1. w. ece-'a - c:;crbi, tire 11rtz:nt E: :a._r e, e.:
_m (full length control rod assembly 12 feet out of alignaent with its bank) opera-tion at SCf steady state power dees not result in exceeding core limits.
The specified control rod assembly drop time is consistent with safety analyses that have been pe-ferred.
An inoperable centrol rod assenbly impeses additional demands on the operators.
The permissible nunber of inoperable control red asse=blies is limited to one in order to limit the magnitude of the operating burden, but such a failure prevent dropping of the operable centrol red asserblies upcn reactor would not trip.
have been chosen as a design basis for fuel performance reitted to Twc criteria fission gas release, pellet te=perature anJ. cladding rechanical prcperties.
C 0 7.
excecd L~:
First, the peak value of fuel centerline te=perature rust not Secend, the cin_=un DNER in the core cust not be les's than 1.30 in ncrral operation or in short ter transients.
In addition tc the above, the peak linear power density, the nuclear enthalpy rise exceed channel factor, and the hot assembly enthalpy rise factor cust net hot their limiting values uhich result from the large break loss of coolant accident analysis based on the ECCS acceptance criteria limit of 22000F on peak clad terperature. This is required to meet the initial conditions assumed for the loss of coolant accident. To aid in specifying the limits on pouer distribution the folicwing hot channel factors are defined.
Amendment No. 50. Unit 1 Amendment No. 49, Unit 2 M%
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803 317
TS 3.12-16 For normal operation, it has been determined that, provided certain condi-tionsareobserved,theenthalpyrisehotchannelfactor,Ffg,11=1: will be cet; these conditions are as rollows:
1.
Contcol rods in a sia31e bank cyce together with no indiviuual rod insertion differing by more than 15 inches from the bank decand position. An indicated misalignment li=it of 13 steps precludes a rod misalign=ent no greater than 15 inches with consideration of maximum instrumentation error.
2.
Centrol rod banks are sequenced with overlapping banks as shown in TS Figures 3.12-1A, 3.:_-13, and 3.12-2.
2.
The rull 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 pcwer between the top and bottom halves of the core.
N The perritted relaxation in FaH with decreasing power level allows radial It has pouer shape changes with rod insertion to the insertion limits.
been determined that provided the above conditions 1 through 4 are observed, this hot ch:nnel factor limit is cet.
OUb 3l8 Amendment No. 50, Unit 1 f
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Amendment No. 49, Unit 2 b' U d UsiM:a w 6 a
TS 5.3-2 3.
Reload fuel will be similar in design to the initial core.
The enrich-
=ent 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 316 oisen rods in th2 form of 12 red clusters, which are located in vacant control rod assembly guide thimbles. The burnable poison rods consist of pyrex clad with -tainless steel.
5.
There are 48 full-lenget. control rod assemblies in the reactor core. The full-lengt' control rod assemblies cont ain a 144-inch length of silver-indium-cadmium alloy clad with stainless steel.
6.
Surry Unit 1, Cycle 4, Surry Uni 2, Cycle 3, and subsequent cores will
=eet the tallowing criteria at all ti=es during the eperatien lif a:ime.
Hot channel factor limits as specified in Secticn 3.12 shall ~:e a.
net.
Amendment No. 50, Unit 1 Amendment No. 49, Unit 2 3
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