ML19210A528
| ML19210A528 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 01/16/1976 |
| From: | Arnold R METROPOLITAN EDISON CO. |
| To: | |
| Shared Package | |
| ML19210A523 | List: |
| References | |
| NUDOCS 7910300567 | |
| Download: ML19210A528 (7) | |
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umnM@'7b METROPOLITAN EDISON COMPANY JERSEY CENTRAL POWER & LIGHT COMPANY AND PENNSYLVANIA ELECTRI: COMPANY THREE MILE ISLAND NUCLEAR STATION UNIT 1 Operating License No. DPR-50 Docket No. 50-289 Technical Specification Change Request No. 31 This Technical Specification Change Request is submitted in support of Licensee's request to change Appendix A to Operating License No. DPR-50 for Three Mile Island Nuclear Station Unit 1. As a part of this request, proposed replacement pages for Appendix A are also included.
METROPOLITAN EDISON COMPANY n
7 /
By M Vice Pregtdent-Gene afion Sworn and subscribed to me this day of ,
Notary Public
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METROPOLITAN EDISON COMPANY JERSEY CENTRAL POWER & LIGHT COMPANY AND PENNSYLVANIA ELECTRIC COMPANY THREE MILE ISLAND NUCLEAR STATI0'.' UNIT 1 Operating License No. P?R-50 Docket No. 50-289 Technical Specification Change Request No. 31 This Technical Specification Change Request is submitted in support of Licensee's request to change Appendix A to Operating License No. DPR-50 for Three Mile Island Nuclear Station Unit 1. As a part of this request, proposed replacement pages for Appendix l. are also 2 ncluded.
METROPOLITAN EDISON COMPANY By /s/ R. C. Arnold Vice President-Generation Sworn and subscribed to me this 16th day of January , 1976 Lawrence L. Lawyer Notary Public 1487 319
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION IN THE MATTER OF DOCKET NO. 50-289 OPERATING LICENSE NO. DPR-50 METROPOLITAN EDISON COMPAhT This is to certify that a copy of Tunnical Specification Change Request No. 31 to Appendix A of the Operating License for Three Mile Island Nuclear Station, Unit 1, dated January 16, 1976, and filed with the U.S. Nuclear Regulatory Commission January 16, 1976, has this 16th day January, 1976, been served on the chief executives of Londonderry Township, Dauphin County, Pennsylvania, and of Dauphin County, Pennsylvania, by deposit in the United States Mail, addressed as follows:
Mr. Weldon B. Arehart, Chairman Mr. Charles P. Hoy, Chairman Board of Supervisors of Board of County Commissioners of Londonderry Township Dauphin County R.D. #1, Geyers Church Road Dauphin County Courthouse Middletown, Pennsylvania 17057 Harrisburg, Pennsylvania 17120 METROPOLITAN EDISON COMPAh7 (l
By h [b Vice Preifident-Generation 1487 520
THREE MILE ISLAND NUCLEAR STATION UNIT 1 (TMI-1)
Operating License No. DPR-50 Docket No. 50-289 Technical Specification Change Request No. 31 The licensee requests that the attached changed pages replace pages 2-5, 2-6, 2-9, and table 2.3-1 of the existing technical specifications.
Reasons for Proposed Change Metropolitan Edison is currently involved in a program with Babcock and Nilcox to improve the capability of the TMI-l plant to withstand a loss of electrical load (LOEL) from 100% power without tripping the reactor.
The ability to ride through a L0EL without tripping the reactor is important for several reasons. In the first place, if the reactor trips the plant cannot resume power generation for a considerable length of time following the load loss. In order to ensure continuity of electric power supplied to the public it is desirable for the plant to be able to pick up load again as soon as possible af ter the initial interruption. In the second place, if the reactor trips following a separation of the station from the grid, emergency sources of power must be relied upon to run vital station auxiliaries. If reactor trip is prevented, the plant can continue to supply all auxiliaries in the normal fashion.
In order to improve the ability of the TMI-l plant to avoid reactor trip on a loss of load, Babcock and Wilcox has reccmmended several relatively minor plant modifications. For the most part, the modifications are aimed at increasing the effectiveness of the secondary plant steam relief systems and improving the response of the steam generator controls. In addition to these changes, however, B&W has recommended an increase in the high reactor coolant pressure reactor trip setpoint.
The attached supplementary safety analysis justifies a revised trip setting of 2405 psig. Accordingly, Met-Ed ultimately intends to increase the high pressure reactor trip setting from 2355 psig to 2405 psig on a permanent basis.
In addition, in order to preserve the margin between the high pressure trip setting and the setting of the pressurizer code safety valves, Met-Ed intends to increase the set pressure of the safety valves, at the same time the permanent change to the high pressure trip setting is made. Preliminary study indicates that unnecessary conservacism can be removed from the calculation which was used to determine the current safety valve setting of 2435 psig, with the result that a revised setting of 2500 psig can be implemented without comp romis ing the plant's protection from overpressure. A final analysis justifying an increased safety valve setpoint will be completed in the near future, at which time a request for the revised safety valve setting and a permanent reactor trip setting of 2405 psig will be submitted. If approved, these changes will be made during the refueling outage scheduled to commence in February of this year.
1487 3.21
Just prior to the refueling outage in Feb:oary, Met-Ed intends to perform a test of the loss of electrical load transient at TMI-1. The purpose of the test is to obtain sufficient data on the LOEL transient to evaluate the ef fectiveness of recent plant modifications in improving plant response to a LOEL. In order to ensure that the test is successful, i.e. , data is obtained during a successful runback from 100% to 15% power without reactor trip, it is desired to raise the high pressure reactor trip setting to 2405 psig just prior to the test. However, it is not practical to increase the pressurizer safety valve setting before tne test. Therefore, Met-Ed requests approval for a change in the trip setting for the purpose of the test. A request for permanent increased settings of both the reactor trip and the cede safety valves will be submitted in the near future, to be implemented during the refueling outage as discussed above.
Safety Analysis Justifying Change The attached supplementary safety analysis provides justification for the revised high pressure reactor trip setpoint of 2405 psig. The analysis indicates th a t , with the revised trip setting, for the end of life conditions which will exist for the LOEL test, reactor coolant system pressure as well as other critical plant parameters are held below the safety limits for the most limiting accidents treated in the TMI-l FSAR. (It also shows that for all conditions which will be encountered in fuel cycle 1, critical parameters will be maintained below these limits.) Note that the accidents analysed in Section 14.1.2.3 of the FSAR, Rod Withdrawal from Rated Power, and Section 14.1.2.4, Moderator Dilution, have less severe consequences and, therefore, have not been specifically analyzed in the attachment. The Loss of Electric Power treated in FSAR Section 14.1.2.8 also has less severe consequences and requires no further analysis.
Since the pressurizer code safety valve setpoint must remain at the current setting of 2435 psig during tne LOEL test, it is conceivable that some postulated accident could cause lif ting of the safety valves prior to the reactor coolant pressure reaching the revised high pressure reactor trip setting. Our analysis of the consequences of such an accident indicate that no hazard to public safety would result. However, to provide additional assurance in this regard, special precautions will be taken during the period of time the trip setting is raised to protect against any possible adverse consequences of lifting the safety valves prior to high pressure trip.
With regard to the LOEL transient itself, increasing the reactor trip setting to 2405 psig will not present a risk of reactor coolant system overpressure during the test. Data obtained from previous full power load rejections show that the pressure reached following the LOEL did not appreciably overshoot the present trip point (2355 psig). Raising the high pressure trip setting to 2405 psig can increase the peak pressure by at most 50 psi (the amount of the proposed increase in the trip setting) . Therefore, the maximum reactor coolant pressure will still be limited to well below the plant design pressure.
1487 522
2.3 LD:ITI!:G SAFEri F73TE." SE"rII:GS , PROTECTIO:! I::ST.RG'E:ITATIC::
Applicability Applies to instruments conitoring reactor power, reactor pcVer inbalance, reactor coolant cysten pressure, reacter coolant outlet tenperature, flow, nu=ber of pumps in operation, and high reactor building pressure.
Objective To provide automatic protection action to prevent any ec bination of process variables from exceeding a safety limit.
Snecifications 2.3.1 The reactor protection system trip setting limits and the permissible bypasses for the instrument channels shall be as stated in Table 2.3-1 and Figure 2 3-2.
2.3.2 For the 2h hour period prior to shutdown for refueling the high pressure trip setpoint may be increased to 2h05 psig provided that the follcuing precautions are taken during this period.
- a. If pressurizer level exceeds 315 inches as indicated by the high-high level alars, the reactor shall be manually tripped.
- b. If the reactor drain tank pressure exceeds 15 psig, the reactor shall be manually tripped.
! 2 2-5
hs es 3e reactor protecticn system censists of four instrucent channels to eni .or each of several selected plant conditions thich will cause a enetc r nt'p if any one of these conditions deviates from a pre-selected opera'.ing r cp to the degree that a safety limit may be reached.
The trip settin6 limits for protecticn system instrr.entatic n rce liste.. i-Table 2 3-1. The safety analysis has been based upon these protectior. /st:
instrumentation trip set points plus calibration and instrumentation errors.
[uclear Overpower A reactor trip at high power level (neutron flux) is provided to prevent da= age to the fuel cladding from reactivity excursions too rapid to be detected by pressure and temperature measurements.
During normal plant operation with all reactor coolant pumps operating, reactor trip is initiated when the reactor power level reaches 105 5% of rated power.
Adding to this the possible variation in trip set points due to calibration and instrument errors, the maximum actual power at which a trip would be actu-ated could be 112%, which is more conservative than the value used in the safety analysis (l).
- a. Overpower trip based on flow and imbalance The power level trip set point produced by the reactor coolant systen flow is based on a power-to-flow ratio which has been established to accormodate the most severe thermal transient considered in the design, the loss-of-coolant flow accident from high power. Analysis has demonstrated that the specified power to flow ratio is adequate to prevent a DNER of less than 13 should a lev flow condition exist due to any malfunction.
O 6O9 -
1487 324 3-Sa
The power level trip set point produced by the power-to-flow ratio provides both high power level and low flow protection in the event I the reactor power level increases or the reactor coolant flow rate decreases. The power level trip set point produced by the power to flow ratio provides overpower D:IB protection for all modes of pump operation. For every flow rate there is a maximum permissible power level, and for every power level there is a minimum permissible lov flow rate. Typical power level and lov flow rate combinations for the pump situations of Table 2 3-1 are as follows:
- 1. Trip vould occur when four reactor coolant pu=ps are operating if power is 108 percent and reactor flow rate is 100 percent, or flow rate is 92.6 percent pnd power level is 100 percent.
- 2. Trip would occur when three reactor coolant pumps are operating if power is 80.7 percent and reactor flow rate is Th.7 percent or flow rate is 69 2 percent and power level is 75 percent.
3 Trip would occur when one reactor coolant pump is operating in each loop (total of two pumps operating) if the power is 52 9 percent and reactor flow rate is 49 0 percent or flow rate is h5.4 percent and the power level is h9 percent.
For safety analysis calculations the maximum calibration and instrumentation errors for the power level were used.
The power-imbalance boundaries are established in order to prevent reactor thermal limits from being e..ceeded. These ther=al limits are either power
('
peaking kW/ft limits or D:IBR limits. The reactor power imbalance (power in the top half of core minus power in the bottom half of core) reduces the power level trip produced by the pover-to-flow ratio so that the boundaries of Figure 2.3-2 are produced. The pover-to-flow ratio reduces the power level trip and associated reactor-power reactor-power-imbalance boundaries by 1.08 percent for a one percent flow reduction.
- b. Pump monitors The redundant pu p ronitors prevent the minimum core DI!BR from decreasing below 1.3 by tripping the reactor due to the loss of reactor coolant pump (s). The pump monitors also restrict the power level for the number of pu=ps in operation.
- c. Reactor coolant system pressure During a startup accident frem low power or a slow rod withdrc a1 from high power, the system hi6h pressure trip set point is reached before the nuclear overpover trip set poir.t. The trip setting limit shown ir Fi. pre 2.3-1 for high reactor coolant sy., tem pressure (2355*psic) has been established to maintain the system pre.asure be!cv the safety limit (2750 psic) for any desig: t rans ier.t . l
- Except as specified in 2.3.2 1487 325 2-6
TABLE 2.3-1 REACTOR PROTECTION SYSTEM TRIP SE'ITING LIMITS One Reactor Coolant Pump Four Reactor Coolant Pumps Three Reactor Coolant Pumps Operating in Each Loop S'iutdown Operating (Nominal (Nominal Operating Bypass Operating (Nominal Operating Power - 100%) Operating Power - 75%) Power - 49%)
- 1. Nuclear power, Max. 5.0(3) 105.5 105 5 105.5
% or rated power 1.08 times flow minus 1.08 times flow minus Bypassed
- 2. Nuclen r gover based 1.08 times flow minus reduction due to reduction due to on riow(P) and imbal- reduction due to imbalance (s) resce , max. of rated imbalance (s) imbalance (s )
power NA 91% Bypassed 3 Nuclear lover based (S) NA on pump monitors, max.
f,of rated power 2355
- 2355* 1720(h)
? h. Hir,h reactor coolant 2355 *
- cystem pressure, psig, max.
1800 1800 Bypassed 5 Low reactor coolant 1800 system pressure, psig, mi Yl .
(16.25 Tout - 7756)(1) (16.25 Tout - 7756)(1) Bypassed
- 6. Variable low reactor (16.25 Tout - 7756)(1) coolttnt cysten prenaure, psig, min.
619 619 619 7 lleactor coolant temp. 619 F., Max.
A h h h
- 8. liit h iteactor Building h N
6 prescure, psig, max.
(1) ' Pout is in degrees Fahrenheit (F)
- Except as specified in 2.3.2
( ?) ' ten. tor coolant system flow, T
( 3) 'Mn.inistratively controlled reduction set on',, during reactor shutdown
( h ) Automatically set when other segments of the RPS (as specified) are bypassed (5) The pump n.onitors also produce a trip on: (a) loss of two reactor coolant pumps i n < mi- re tetor coolant loop, and (b) loss of one or two reactor coolant pumps during two-pump operation.
2500 P = 2355*
2300
?
E as E 2100 C
?
a.
E 2
31900 f
I; Z
a::
1700 1500 540 560 580 600 620 640 Reactor Outlet Temperature, F
- Except as specified in 2.3.2 PROTECT 10N SYSTEM MAXIMUM ALLOWABLE SET PolHTS THREE MILE ISLAND NUCLEAR STATION UNIT I
, FIGUR E 2.3-1 1487 327