ML20062C247
| ML20062C247 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 10/24/1978 |
| From: | Ziemann D Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20062C243 | List: |
| References | |
| NUDOCS 7811070307 | |
| Download: ML20062C247 (10) | |
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UNITED STATES y#
' g NUCLEAR REGULATORY COMMisslON I,'
. I WASHINGTON. D. C. 20555
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CONNECTICUT YANKEE ATOMIC POWER COMPANY DOCKET NO. 50-213 HADDAM NECK PLANT AMENDMENT TO FACILITY OPERATING LICENSE Amendment 'No. 29 License No. DPR-61 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application by Connecticut Yankee Atomic Power Company (the licensee) dated September 29, 1978, as supported by the analyses submitted by letters dated September 22 and October 20, 1978, complies with the standards and require-ments of the Atomic Energy Act of 1954, as amended (the Act),and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the license amendment, the provisions of the Act, and the rules and regulations of the Comission; 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 4
comon 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 Comission's regulations and all applicable requirements have been satisfied.
l 2.
Accordir. gly, Facility Operating License No. DPR-61 is hereby amended by deleting paragraph C.(3) in its entirety, by changing the Technical Specifications as indicated in the attachment to this license amendment and by amending paragraph 2.C.(2) to read l
as follows:
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"(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 29, are hereby
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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 C0f911SSION J !3 g __
x Qr rm. w DennisL.ZiemanNChief Operating Reactors Branch #2 Division of Operating Reactors
Attachment:
Changes to the Technical Specifications Date of Issuance: October 24, 1978
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ATTACHMENT TO LICENSE AMENDMENT NO. 29 FACILITY LICENSE NO. DPR-61 DOCKET NO. 50-213 Revise the Appendix A Technical Specifications by deleting the following pages and inserting the enclosed pages. The revised pages contain the captioned amendment number and vertical lines reflecting the area
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of change.
Remove Pages Insert Pages 1-6 1-6 2-5 2-5 2-7 2-7 3-10b 3-10b 3-16 3-16 3-17 3-17 3-17a
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TABLE 1.1 OPERATIONAL MODES REACTIVITY
% RATED AVERAGE COOLANT MODE CONDITION,Keff THERMAL POWER
- TEMPERATURE 0
t 1.
POWER OPERATION 20.99
> 5%
1350 F 2.
STARTUP
>.97 3 5%
23500F 3.
HOT STANDBY 5.0.97 0
1,350'F 0
4.
HOT SHUTDOWN 0
350 F > Tavg> 2000F 3.0.97 5.
COLD SHUTDOWN 5.0.97 0
$ 200*F 6.
REFUELING **
50.92 0
5 140*F k
- Excluding decay heat.
- Reactor vessel head unbolted or removed and fuel in the vessel.
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Amendment No. 29
2.4-MAXIMUM SAFETY SETTINGS - PROTECTIVE INSTRUMENTATION Applicability:
Applies to trip settings for instruments monitoring reactor power and reactor coolant pressure, temperature, and flow.
Objective:
To provide for protective action in the event that the principle process variables approach a safety limit.
Specification:
Protective instrumentation trip settings shall be as follows:
Four Reactor Coolant Three Reactor Coolant Pumps Operating Pumps Operating (1) Pressurizer Pressure
- E 2300 psig
' < 2300 psig
<86%ofranky
< 86% of range (2) Pressurizer Level *
(3) Variable Low Pressure *** 3,17.4(Tavg+1.38AI)-8850 3,17.4(Tavg+1.38aI)-8850 (4) Nuclear Overpower **
ji109% of rated power
- [ 84% of rated power (5) Low Coolant Flow ***
> 90% of normal loop flow 3,90% of normal loop flow (6) Reactor Coolant Loop Valve-Tempoerature
- E 200F
< 200F Interlock (7) High Steam Flow 110% of full load steam 110% of full load steam flow flow j
May be bypassed when the reactor is at least 1.5%Ak suberitical.
A The nuclear overpower trip is based upon a symmetrical core' power distri-bution.
If any asymmetric power distribution should occur, resulting in the power in any quadrant being 10% greater than the average core power as indicated by the neutron detectors and loop AT measurements, the nuclear overpower trip on all channels shall be reduced one percent for each percent deviation greater than the above 10%.
If it is determined that asymmetric power distributions exceed 30%, the reactor will be shut down.
When the reactor power is <10% the overpower trip setpoint is recuced to l
25% of rated power.
l May be bypassed below 10% of rated power.
Basis:
The reactor protective system is desigred and constructed such that no single failure in any of the instrument systems will prevent the desired safety action if an applicable parameter exceed a safety set point.
knendment No. 29 2-5
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and shutdown..It is safe to block this trip below 10%
power since the protection afforded by this trip is not I
required at this low level.
Removal of unnecessary trip signals will reduce the number of spurious trips.
(4) Nuclear Overpower. As explained above, the nuclear over-power reactor trip, in conjunction with the variable low pressure reactor trip, provides overpower, overtemperature protection. The nuclear overpower trip channels will respond first to rapid reactivity insertion rates, detected by the increase in flux, before there are any significant changes in the system process variables. A maximum error of 9% of full power due to set point, instrumentation, and calorimetric determination (see Section 4.3.6 of the FDSA) is considered in establishing the set point.
In order to reduce the time to trip for certain accidents occurring at low power, the overpower setpoint is lowered to 25% when reactor power is below 10%. This low overpower trip would terminate the postulated large steamline break accident from the hot zero power 1
condition. The lower setting for three loop operation j
provides protection at the reduced power level equivalent i
to that provided by the setting for four loop operation at full power. The reduction in setting in the event of an asymmetric power distribution provides protection for the more adverse hot channel factors. The asymmetry is detected primarily by observation of changes in loop AT and neutron detector ion chamber current readings, each of which is displayed at the control board. The loop AT is the difference between the hot leg and cold leg tempera-ture of the reactor coolant, as measured at the steam generator.
Experience with operating reactors indicates i
that power distribution asymmetrics from + 3 to 6% of l
nominal power can be detected by either of the above l
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methods.
If any of the neutron detectors or loop AT measurements differs from the average by.more than 6%, a critical review of core power distributions will be per--
formad to evaluate the significance of the measurement l
from the standpoint of power distribution.
(5) Low Coolant Flow. The low coolant flow reactor trip pro-tects the core against ag increase in coolant temperature resulting from a reduction in coolant flow while the reactor is at substantial power (3). This trip will pre-vent DNB in any loss-of-flow incident, which eliminates the possibility of clad damage. Flow detection in each reactor coolant loop is from a measurement of pressure drop from inlet to outlet of each steam generator. The 90%
low flow signal is high enough to activate a-trip in time to prevent DNB, and low enough to reflect that a loss-of-flow condition truly exists. A maximum instrument and set point error of 5% full flow is considered in determin-ing the set point.
Loss-of-flow protection is also pro-vided by reactor coolant pump breaker and from undervoltage 2-7 Amendment No.'29
BASIS This specification assures that adequate emergency core cooling capacity is available whenever the reactor is critical. Based on the loss of coolant analysis, melting of the cladding is prevented with only one high pressure safety injection pump and one low pressure safety injection (core deluge) pump in operation.
Each of the two trains of emergency core cooling equipment includes these two pumps. With the pumps associated with both trains of emergency core cooling equipment operable, substantial margin exists whenever normal power supplies or both diesel generators are avail-able.
With only one diesel generator operating and the pumps associated with that diesel operable as required in Item (2) of Specification 3.12, the high pressure safety injection pump and the low pressure safety injec-tion pump would be started automatically. When the safety injection pumps are operating on off-site power, the charging pump would be started auto-matically. The RHR pump would be available for manual start for long-term recirculation cooling.
(2)
FDSA Section S.2.8 (3)
D. C. Switzer (CYAPCO) letter to D. L. Ziemann (NRC) dated May 22, 1978.
(4)
D. C. Switzer (CYAPCO) letter to D. L. Ziemann (NRC) dated May 24, 1978.
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3-10b Amendment No. 29
3.10 REACTIVITY CONTROL Applicability:
Applies to control group position during power operation and shutdown margin during subcritical operation (except refueling).
Obj ec tive :
To define control group insertion limits which insure:
(1) An acceptable core power distribution during power opera-tion, (2) A conservative limit on potential reactivity inser-tion for a hypothetical control rod ejection, and (3) Adequate shutdown margins after a reactor trip. To insure that a least 3% SK shutdown margin is available during subcritical operation.
Specification:
A.
Except for low power physics test at or below 10 percent of full power or determination of "just critical" rod positions, operation of the control group banks shall be maintained above the limits shown in Figure 3.10-1.
B.
If it is determined that a rod has been dropped, an eval-uation of the effect of the dropped rod shall be made to establish permissible power levels for continued operation.
C.
No more than one dropped or one stuck rod shall be per-mitted while the reactor is critical nor shall that rod remain in such condition for more than one full power month of reactor operation.
D.
The maximum worth of any individual control rod in the core at rated power shall not exceed 0.17% AK,' as measured at the beginning of core life.
E.
The maximum worth of any individual control rod in the core with the reactor just critical shall not exceed 0.83% AK, as measured at the beginning of core life.
F.
Except for physics testing, a 3% SK shutdown margin shall g
be maintained during suberitical operation., This shutdown margin may be provided by control rods actually inserted, control rods available to insert (considering a stuck rod),
and/or soluble boron.
Basis:
Specification C limits the time a dropped control rod may be in the core because lower fuel depletion and fission product inventory in the vicinity of t'ne dropped rod, relative to the rest of the core, increases the worth of that rod. The lack of fuel depletion and lack of fission products 3ther than Xenon in the vicinity of a control rod which has been inserted for one full power month will have a negigible effect on the worth of that control rod. Xenon redistribution causes an appreciable increase in the worth of a dropped rod.
The in-creased worth has been measured and found to be acceptable.
3-16 Amendment No.29
'3.10 REACTIVITY CONTROL (continued)
The methods of reactivity control to be used are fully explained in Section 4.2.2 of the FDSA. The control rod program was developed to insure that three major safety considerations are satisfied throughout core life.
They are:
1.
Power distributions (DNB ratios) with equilibrium xenon shall be at least as favorable as those used in the safety analysis and shall be within the limits of Fag and Sepcification 3.17.
2.
Sufficient shutdown margin shall be available to ensure that:
(a) the reactor can be made sufficiently subcritical after a trip from any operating condition, including an allowance for the maximum worth stuck rod, and.(b) the reactor does not return to criticality for any FD3A Chapter 10 postulated accident.
e 3.
Poter.tial ejected rod worths shall nut exceed the limits specified in Section 10.2.7 of the FDSA.
The above safety considerations are satisfied because:
(1) As seen in Section 4.3 of the FDSA, the calculated minimum DNB ratio using the rod program is 3.05 compared with 2.82 using the power distributions considered in the safety analysis.
(2)
Since shutdown margin requirements vary throughout core life as a function of fuel depletion, RCS boron concentration, and RCS Taverage.
the limits of figure 3.10-1 are set to ensure that the shutdown margin after a reactor trip from the most limiting set of reactor conditions meets the design criteria (FDSA Section 4-2-2) of at least 3% 6K with all rods inserted and at least 1% 6K with the highest worth rod stuck out.
In addition, operation within the limits of figure 3.10-1 ensures that sufficient shutdown margin (including a stuck rod) is available to prevent the reactor from returning to criticality during the most limiting FDSA Chapter 10 postulated accident.
(3) Of the three requirements, the third was calculated to be most limiting, and this fonns the basis for Specifications A and D above.
As shown in FDSA Section 10.2.7, the analysis for the rod ejection was quite conservative and a large margin would aist to fuel melting and dispersion.
(4)
Power distribution, control rod worths and shutdown margins will be evaluated prior to initial startup and subsequent startups following refueling. Conformance to the above requirements will be checked at these times and the limit of Figure 3.10-1 adjusted to meet these requirements.
Specification C limits the time a dropped control rod may be in the core because lower fuel depletion and fission product inventory in the vicinity of the dropped rod, relative to the rest of the core, increases the worth of that rod. The lack of fuel depletion and lack of fission products other than Xenon in thevicinity of a control rod which has been inserted for one full power month will have a negligible effect on 3-17 Amendment No. 29
3.10 REACTIVITY CONTROL (continued) the worth of that control rod.
Xenon redistribution causes an appreciable increase in the worth of a dropped rod. The increased worth has been measured and found to be acceptable.
Should a control rod be dropped, no.immediate adverse effects would occur due to automatic load cut-back as described in FDSA Section 7.2.3.
Specification F insures at least 3% 6K shutdown margin is. available when the reactor is subcritical. This margin is required to offset the reactivity addition that would occur during a postulated large steamline break accident.
Re ferences:
1.
FDSA Section 4.2 2.
FOSA Section 7,2,3
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s 3-17a Amendment No. 29 i
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