ML14175A583
| ML14175A583 | |
| Person / Time | |
|---|---|
| Site: | Robinson  (DPR-023) |
| Issue date: | 04/11/1979 |
| From: | Schwencer A Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML14175A584 | List: |
| References | |
| DPR-23-A-036 NUDOCS 7905070125 | |
| Download: ML14175A583 (13) | |
Text
UNITED STATES NUCLEAR RI:GULATORY COMMISSION WASHINGTON, D. C. 20555 CAROLINA POWER AND LIGHT COMPANY DOCKET NO. 50-261 H. B. ROBINSON STEAM ELECTRIC PLANT UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 36 License No. DPR-23
- 1. The Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment by Carolina Power and Light Company (the licensee) dated March 6, 1979, complies 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 conformity 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 satisified.
'7905070 12<5
-2
- 2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B. of Facility Operating License No.
DPR-23 is hereby amendedto read as follows:
"B. Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 36, are hereby incorporated in the license. The licensee shall 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 A. Schwencer, Chief Operating Reactors Branch #1 Division of Operating Reactors
Attachment:
Changes to the Technical Specifications Date of Issuance: April 11, 1979
ATTACHMENT TO LICENSE AMENDMENT NO. 36 FACILITY OPERATING LICENSE NO. DPR-23 DOCKET NO. 50-261 Revise Appendix A as follows:
Remove the following pages and insert identically numbered revised pages:
Pages ii 3.10-2 3.10-4 3.10-8 3.10-9 3.10-11 3.10-13 3.10-14 3.10-19 5.3-2
Section Tilpa 3.10.5' Deleted 3.10.6 Inoperable Control Rods 3.10-8 3.10.7 Power Ramp Rate Limits 3.10-9 3.10.8 Required Shutdown Marsins 3.10-9 3.11 Movable in-Core Instrumentation 3.11-1 3.12 Seismic Shutdown 3.12-1 3.13 Shock Suppressors (Snubbers) 3.13-1 3.14 Fire Protection System 3.14-1 3.14.1 Fire Detection Instrumentation 3.14-1 3.14.2 Fire Suppression Water System 3.14-1 3.14.3 C02 Fire ?otectior System 3.14-2 3.14.4 Fire Hose Stations 3.14-2a 3.14.5 Fire Barrier Penetration Fire Seals 3.14-3 4.0 Surveillance Requirements 4.1-1 4'.1 Operational Safety Review 4.1-1 4.2 Primary System Surveillance 4.2-1 4.3 Primary System Test:ig Following Opening 4.3-1 4.4 Containment Tests 4.4-1 4.4.1 Operational Leakage Rate Tests 4.4-1 4.4.2 Isolation Valve Tests 4.4-4 4.4.3 Post Accident Recirculation Beat Removal System 4.4-4 4.4.4 Operational Surveillance Program 4.4-5 4.5
£mergency Core Cooling, Contaiment Cooling and Iodine Removal Systems Tests 4.5-1 4.5.1 System Tests 4.5-1 4.5.2 Component Tests 4.5-2 4.6 Emergency Power System Periodic Tests 4.6-1 4.6.1 Diesel Generators 4.6-1 4.6.2 Diesel :uel Tanks 4.6-2 4.6.3 Station Batteries 4.6-2 4.7 Secondary. Steam and Power Conversion System 4.7-1 4.8 Auxiliary Feedwater System 4.8-1 4.9 Reactivity Ano=alies 4.9-1 4.10 Radioac:ive Effluents 4.10-1 4.11 Reactor Core 4.11-1 4.12 Refueling Til:er Systems 4.12-1 4.13 Shock Suppresors (Snubbers) 4.13-1 4.14 Fire Protection System
- 4. ~4 5.0 Design Features 5.1-1 5.1 Site 5.1-1 5.2 Contatment 5.2-1 5.2.1 Reactor Contair ant 5.2-1 5.2.2
?enetrations 5.2-1 5.2.3 Containment System 5.2-2 5.3 Reactor 5.3-1 5.3.1 Reactor Core 5.3-1 5.3.2 Reactor Cool-a: System 5.3-2 5.4 Fuel Storage 5.4-1 5.5 SeIsic Design 5.5-1
-ii Amendment No. 36
3.10.1.5 Except foroysics tests, if a full-lengtu4 ntrol rod is more than 15 inches out of alignment with its bank, then within two hours:
- a.
Correct the situation, or
- b.
Determine by measurement the hot channel factors and appl Specification 3.LO.2.1, or
- c.
Limit power to 70 percent of rated power for three-loop operation.
3.10.1.6 Insertion limits do not apply during physics tests or during periodic bxercise of individual rods. However, the shutdown margin indicated in Figure 3.10-2 must be maintained except for the low power physics test to masure control rod worth and shutdown margin.
For this test the reactor may be critical with all but one full length control rod inserted.
3.10.2 Power Distribution Limits 3.10.2.1 At all times except during low power physics tests, the hot channel factors defipqed in the basis anst meet the following limits:
Q (Z) < (2.20/?) X K(Z) for P >.5 FQ (Z) < (4.40) X K(Z) for P <.5 al.
pN
< 1.55 (1 + 0.2(1-P))
where P is the fractin of licensed power at which the core is operating, K(Z) is the function given in Figure 3.10-3, and Z is the core height location of FQ.
3.10-2 Amendment No. 36
3.10.2.1.2 The predetermined power level at which ADnS initiation is required is given by the relation SAPDMS < 1.435 x 0.94 Fxy 3.10.2.1.3 Fxy shall.be determined for the unrodded core plane regions away from fuel support grids, located between a core plane elevation 2.0 feet from the top of the core and a core plane elevation 2.0 feet from the bottom of the core with no control rod inserted more than 2.0 feet into the core. This determina tion shall be made from the movable incore detector maps specified in 3.10.2.3.
3.10.2.2 If either measured hot channel factor exceeds these values the reactor power shall be reduced so as not to exceed a frac tion of the design value equal to the ratio of the FN or FNI limit to measured value, whichever is less, and the high neutron flux trip setpoint shall be reduced by the same ratio.
If subsequent incore mapping cannot, within a 24-hour period, demonstrate that the hot channel factors are met, the over power AT and overtemperature AT trip setpoints shall be similarly reduced.
3.10.2.3 Following initial loading and at regular monthly intervals thereafter, power distribution maps using the movable detector system, shall be made to confirm that the hot channel factor limits of Specification 3.10.2.1 are satisfied. For the purpose of this confirmation:
- a.
The measurement of total peaking factor, FMeas shall be Q
increased by three percent to account for manufacturing tolerances and further increased by five percent to account for measurement error.
3.10-4 Amendment No. 36
3.10.4 Rod Drop Time 3.10.4.1 The drop time of each control rod shall be not greater than 1.8 seconds at full flow and operating temperature from the beginning of rod motion to dashpot entry.
3.10.5 Deleted 3.10.6 Inoperable Control Rods 3.10.6.1 A control rod shall be deemed inoperable if (a) the rod is misaligned by more than 15 inches with its bank, (b) if the rod cannot be moved by its drive mechanism, or (c) if its rod drop time is not met.
3.10-8 Amendment No. 36
3.10.6.2 No more than one inonerable control rod s1 be permitted power operation.
3.10.6.3 If a full length control rod cannot be moved by its mechanism, boron concentration shall be changed to compensate for the with drawn worth of the inoperable rod such that shutdown margin equal to or greater than shown on Figure 3.10-2 results.
3.10.7 Power Ramp Rate Limits 3.10.7.1 During the return to power following a shutdown where fuel assemblies have been handled (e.g., refueling, inspection),
the rate of reactor power increase shall be limited to 3 per cent of full power in an hour between 20 percent and 100 percent of full power. This ram; rate requirement applies during the initial startup and may apply during subsequent power increases depending on the maximum power level achieved and length of opera tion at that power level.
Specifically, this requirement can be removed for reactor power levels below a power level P (20 percent
<P <100 percent) provi.ded that the plant has operated at or above power level P for at least 72 cumulative hours out of any seven-day operating period following the shutdown.
3.10.7.2 The rate of reactor power increases above the highest power level sustained for a: least 72 cumulative hours during the preceding 30 cumulative days of reactor power operation shall be limited to 3 percent of full power in an hour. Alternatively, reactor power increase can be accomplished by a single step increase less than or equal to 10 percent of full power followed by a maximum ramp rate of 3 percent of full power in an hour beginning three hours after the step increase.
3.10.8 Required Shutdown Margins 3.10.8.1 When the reactor is in the hot shutdown condition, the shutdown margin shall be at leist that shown in Figure 3.10-2.
3.10-9 Amendment No. 36
shutdown margin. The specified control rod insertiofflimits meet the design basis criteria on (1) potential ejected control rod worth and peak ing factor, (2) radial power peaking factors, F,, and (3) required margin shutdown.
The various control rod banks (shutdown banks, control banks) are each to be moved as a bank; that is, with all rods in the bank within one step (5/8 inch) of the bank position. Position indication is provided by two methods:
a digital count of actuation pulses which shows the demand position of the banks, and a linear position indicator (LVDT) which indicates the actual rod position.(2) The 15-inch permissible misalignment provides an enforceable limit below which design distribution is not exceeded. In the event that an LVDT is not in service, the effects of a malpositioned control rod are observable on nuclear and process information displayed in the control room and by core thermocouples and in-core movable detectors. The determination of the hot channel factors will be performed by means of the movable in-core detectors.
The two hours in 3.10.1.5 are acceptable because complete rod misalign ment (control rod 12 feet out of alignment with its bank) does not result in exceeding core safety limits in steady state operation at rated power and is short with respect to probability of an independent accident.
If the condition cannot be readily corrected, the speci fied reduction in power will ensure that design margins to core limits will be maintained under both steady state and anticipated transient conditions.
The intent of the test to measure control rod worth and shutdown margin (Specification 3.10.1.6) is to measure the worth of all rods less the worth of the worst case for an assumed stuck rod; that is, the most re active rod. The measurement would be anticipated as part of the initial startup program and infrequently over the life of the plant, to be associated primarily with determinations of specials interest such as end of life cooldown, or startup of fuel cycles which deviate from normal 3. 10- 1 1 Amendment No. 36
area W the fuel rod and eccentric oof the gap between pallet and clad.
Combined statistically the net affect is a factor of 1.03 to be applied to fuel rod surface heat flux.
- d.
FN, Nuclear Enthalpy Rise Rot Channel Factor, is defined as the ratio of the integral of linear power along the rod with the highest integrated power to the average rod power.
N It should be noted that F is based on an integral and is used as such in AR the DNB calculations. Local heat fluxes are obtained by using hot channel and adjacent channel explicit power shapes which take into account varia tions in horizontal (x-y) power shapes through the core. Thus, the hori zontal power shape at the point of maximum heat flux is not necessarily N
directly related to F N It has been determined by extensive analysis of possible operating power shapes that the design limits on peak local power density and on minimum DNBR at full power are met, provided the values of F and FH in Specification 3.10.2.1 are not exceeded.
For normal operation, it is not necessary to measure these quantities.
Instead, it has been determined that, provided certain conditions are observed, the above hot channel factor limits will be met; these condi tions are as follows:
- a.
Control rods in a single bank move together with no individual rod insertion differing by more than 15 inches from the bank demand position.
- b.
Control rod banks are sequenced with overlapping banks as shown in Figure 3.10-1.
- c.
The control bank insertion limits are not violated.
- d.
Deleted 3.10-13 Amendment No. 36
- e.
Axial power distribution control procedures, which are given in terms of flux difference control, are observed.
Flux difference refers to the difference in signals between the top and bottom halves of two-section axcore neutron detectors. The flux difference is a measure of the axial offset which is defined on the difference in power between the top and bottom halves of the core.
For operation at a fraction P of full power, the design limits are met, provided the limits of Specification 3.10.2.1 are not exceeded.
N The permitted relaxation in F with reduced power 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, these hot channel factors limits are met.
The procedures for axial power distribution control referred to above include operator control of flux difference to minimize the effects of xenon redistribution on the axial power distribution during load-follow maneuvers. Basically, control of flux difference is required to limit the difference between the current value of Flux Difference (Al) and a reference value which corresponds to the full power equilibrium value of Axial Offset (Axial Offset -
LI/fractional power).
The reference value of flux difference varies with power level and burnup but expressed as axial offset, it varies primarily with burnup.
The target (or reference) value of flux difference is determined as follows: At any time that equilibrium xenon conditions have been established, the indicated flux difference is noted with control Bank D more than 190 steps withdrawn. This value, divided by the fraction of full power at which the core was operating is the full power value of the tar get flux difference. Values for all other core power' levels are obtained by multiplying the full power value by the fractional power. Since the indicated equilibrium value was noted, no allowances for excore detector error are necessarv and the wnpcified Mwivation of I iq 3.10-14Amendment No. 36
An inoperable rod imposes additional demands on the operator.
The per missible number of inoperable control rods 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 rods upon reactor trip.
Normal reactor operation causes significant pellet cracking and fragmenta tion. Consequently, handling of irradiated fuel assemblies can result in relocation of these fragments against the cladding.
Calculations show that high cladding stresses can occur if the reactor power increase is rapid during the subsequent startup.
The 72-hour period allows for stress relaxation of the clad before the ramp rate requirement is removed, thereby reducing the potential harmful effects of possible pellet or fragment relocation.
The 3 percent limit is imposed to minimize the effects of adverse cladding stresses resulting from part power operation for extended periods of time.
The time period of 30 days is based upon the successful power ramp demon strations performed on Zircaloy clad fuel in operating reactors, resulting in no cladding failures.
References (1)
FSAR Section 14 and WCAP-8243 (2) FSAR Section 7.3 (3) WCAP-8243, Section 4.4.2 (4) WCAP-8243, Section 4.4.3 3.10-19 Amendment NO. 36
5.3.1.5 There are 45 full-length RCC assemblies in the reactor core. The full-length RCC assemblies contain 144 inch length of silver-indium-cadmium alloy clad with the stainless steel.
5.3.1.6 Up to 10 grams of enriched fissionable material may be used either in the core, or available on the plant site, in the form of fabricated neutron flux detectors for the purposes of monitoring core neutron flux.
5.3.2 Reactor Coolant System 5.3.2.1 The design of the Reactor Coolant System complies with the Code requirements. (6) 5.3.2.2 All piping, components and supporting structures of the Reactor Coolant System are designed to Class I requirements.
5.3.2.3 The nominal liquid volume of the Reactor Coolant System, at rated operating conditions, is 9343 cubic feet.(7 References (1) FSAR Section 3.2.3 (2) FSAR Section 3.2.1 (3) FSAR Section 3.2.1 (4) FSAR Section 3.2.3 (5)
FSAR Sections 3.2.1 and 3.2.3 (6) FSAR Table 4.1-9 (7) FSAR Table 4.1-1 (8) "Description and Evaluation of Test Assemblies Containing Gadolinia Bearing Fuel Rods" submitted with letter dated January 5, 1973, fro CP&L to the Director of Licensing.
(9) "Description and Evaluation of Test Assemblies Containing Gadolinia Bearing Fuel Rods -
H. B. Robinson Unit No. 2 Cycle 3" submitted with letter dated March 12, 1974, from CP&L to the Director of Licensing.
5.3-2 Amendment No. 36