ML19309E069
| ML19309E069 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 04/11/1980 |
| From: | Early P POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| To: | Schwencer A Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8004180350 | |
| Download: ML19309E069 (10) | |
Text
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POWER AUTHORITY OF THE STATE OF NEW YORK to CoLUMeus CIRCLE NEW YORK N. Y. loo 19 (2123 397.6200 GEOR,c,E y MERRY OPERATING OFFICER TRUSTEES JOHN W. SOSTON JOHN 3.DYSON Rasics 7 a osRacTOR OF POWER OPER4TIONS GEORGE L. ING ALLg I
JOSEPH R. SCHMIEDER vics cuanRMAN h
PRES DENT HIEF RICH ARD M. FLYNN Aprk1 11, 1980 LEROY W. SBNCLAIR aO=ERT L MILLONIl IPN-80-39 "c3="Pfi1"!E'***
FREDERICK R. CLARK
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Director of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D.
C.
20555 Attention:
Mr. Albert Schwencer, Chief Operating Reactors Branch No. 1 Division of Operating Reactors
Subject:
Indian Point 3 Nuclear Power Plant Docket No. 50-286 Confirmatory Order - Sixty Day Actions
Dear Sir:
l to this letter provides the Authority's responses to the sixty day action items of the NRC Confirmatory Order (Interim Actions), dated February 11, 1980.
Very truly yours
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Paul Early Vic Presiden and Ass stant Chief Engineer-Projects cc:
Mr. T. Rebelowski, Resident Inspector U. S. Nuclear Regulatory Commission P. O. Box 38 Buchanan, New York 10511 Mr. William J. Cahill, Jr.,Vice President Consolidated Edison Company of New York 4 Irving Place New York, New York 10003 Aoo/
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8004180 3 1
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i ATTACHMENT 1 i
i RESPONSES TO CONFIRMATORY ORDER i
APPENDIX A ITEMS t
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POWER AUTHORITY OF THE STATE OF NEW YORK l
INDIAN POINT 3 NUCLEAR POWER PLANT DOCKET NO. 50-286 APRIL l'1, 1980 i
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l A.7 Require that all reactor operators and senior reactor operators conduct simulator training and in-plant walk through of the following emergency procedures.
The in-plant walk-throughs shall be completed prior to the next reactor startup following issuance of the Order, or within thirty days of the date of issuance, whichever occurs first.
3 Those reactor operators and senior reactor operators who l
have not received simulator training within the past three months on these items shall be given such simulator training within 60 days of the date of the Order:
a.
Plant or reactor startup to include a range wherein L
reactivity feedback from nuclear heat addition is noticeable and heat up rate is established b.
Manual control of steam generator level and/or feed-water during startup and shutdown c.
Any significant (10%) power change using manual rod control L
d.
Loss of Coolant (i) including significant PWR steam generator leaks (ii) inside and outside containment (iii) large and small, including leak rate determination (iv) saturated reactor coolant response (PWR) e.
Loss of core coolant flow / natural circulation i
f.
Loss of all feedwater (normal and emergency) r g.
Station blackout h.
Anticipated Transients Without Scram (ATWS) i 1.
Stuck open relief valve on secondary side j.
Intersystem LOCA i
RESPONSE
The Authority has completed the simulator training required by
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items A.7.a through A.7.j above.,
l C.1 Review the steady state steam generator operating level to l
determine the optimum steady state level for the purpose of maximizing dryout time with due consideration for overfilling.
The results of this study shall be provided to the NRC.
RESPONSE
The steady state generator operating level was chosen based on analyses and setpoint-type studies.
This level was optimized with respect to Class I transients, such as load swings and load rejec-tions, and Chapter 14 FSAR safety analyses.
Many other factors enter into *he selection of this normal operating level such as mass avail. ale for discharge following a secondary pipe rupture, moisture carryover considerations, and steam generator overfilling.
Since all of the above were considered in the optimization of the steam generator normal water level, any change (increase) in the normal water level will of course, cause a departure from optimum.
l More detailed information, with regard to the effect of a change in steam generator level on steam generator dryout time, core un-covery time and moisture carryover is provided below.
j STEAM GENERATOR DRYOUT CALCULATION It should be noted that an increase in nominal steam generator level (i.e., mass) is not the prime consideration in calculation of steam generator dryout time.
A more important consideration is post-trip mass at a low level setpoint, which is the steam gen-erator mass that is used in dryout calculations.
A steam generator dryout calculation computes the time that is required to dissipate the liquid inventory in the steam generator below the low level setpoint due to decay heat generated in the core.
Therefore, raising the low level setpoint will increase the post-trip mass and increase the steam generator dryout time.
The current Indian Point Unit 3 low level setpoint results in a steam generator dryout time of 34 minutes.
An increase in the low level setpoint of 5% of the narrow range span increases the liquid mass by 2950 lbs., and the dryout time by about 1.9 minutes (5.6%).
Table 1 provides the detailed calculation results, which are based on best estimate decay heats.
In addition, if reactor trip is assumed to occur at the normal operating level, an increase in the normal operating level would result in a commensurate increase in the dryout time.
The IP3 steam generator dryout time of 34 minutes, compared to about 3 minutes for TMI, allows considerable time for operator action, in the event it is required.
Operators at TMI took on the order of eight minutes to realign valves and obtain auxiliary feedwater flow. _
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CORE UNCOVERY TIME Based on generic Westinghouse analysis, time required to uncover the core after dryout of the steam generators, is about 30 minutes. Thus, the total time available to ensure that an adequate heat sink exists and to prevent uncovering of the core, is about 64 minutes.
The additional time available, due to an increase in steam generator level of 5% of narrow range is an insignificant 3%.
MOISTURE CARRYOVER The steam generator operating water level has an important effect l
on moisture carryover margin.
This is because of the general trend to increase moisture carryover with an increase in water level above the nominal value.
This trend has been observed at a Westinghouse plant operating at slightly below full power conditions.
Data ob-tained from this plant indicates that a water level increase of approximately 5% of span results in a 15% to 25% increase in moisture carryover.
Since Indian Point Unit No. 3 is currently operating near the upper limit of moisture carryover, it can be concluded that an increase in nominal operating level will result in excessive moisture delivery to the turbine.
CONCLUSION i
Present steam generator levels, both normal operation and low level trip, have been reviewed with regard to increasing them.
The effect of such a change has been shown to be insignificant for Indian Point Unit 3, with respect to increasing operating action time available.
However, such a level increase would lead to potential operating difficulties and turbine damage due to moisture carryover.
l Table 1 - Indian Point Unit 3 Steam Generator Dry Out Calculational Results l
Liquid Inventory at 30% Narrow Range Level 70,870 lbs.
Steam Generator Dryout time for 30% low level trip 34.2 min.
Liquid Inventory increase for increase in low 2,950 lbs.
Level Setpoint by 5%
(i.e., to 35%)
Steam Generator Dryout time for 35% low level trip 36.1 min.
Increase in dryout time for 5% level increase 1.9 min.
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C.2-Evaluate possible co-impregnation of the charcoal in the plant's air effinent filtration systems with K1 & I2 and an amine such as TEDA (triethylene - diamine) to improve the iodine removal capability of these systems.
The results of this review shall be submitted to the NRC.
RESPONSE
An evaluation was performed to determine if there is sufficient justification for replacement of the existing effluent filtration system charcoal beds with co-impregnated charcoal.
The charcoal currently in use at Indian Point 3 is impregnated with 5% by weight of KI3 The only amine which is commercially available at present for use in nuclear plant charcoal adsorbers is TEDA.
Consequently, the evaluation was based on the comparative methyl iodide removal efficiency of the Indian Point 3 charcoal impregnated with KI3 versus charcoal co-impregnated with KI and TEDA.
In addition all charcoal manufacturers contacted stated that of the various types of co-impregnation available (including KI + I2 and I2 + TEDA),
KI + TEDA provides the best methyl iodide removal efficiency.
Indian Point 3 uses two grades of KI3 impregnated charcoal - Mine Safety Appliance type 85851 and type 463563.
Based on the perform-ance data gathered, there is no difference in methyl iodide removal efficiency between type 463563 and that of charcoal co-impregnated with KI and TEDA.
Type 85851 impregnated charcoal provides a methyl iodide removal efficiency of approximately 2% less than co-impregnated charcoal.
This small difference in efficiency does not warrant replacement of approximately half of the charcoal beds at Indian Point 3.
However, as future charcoal replacements are made, the Authority plans to install new charcoal co-impregnated with KI and TEDA.
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3 C.3 Evaluate effects on plant systems stability if power is reduced as much as 50%, treating power as a parameter.
(For example, the effects on the feedwater flow automatic control).
RESPONSE
The IP-3 plant has operated at various power levels for prolonged periods of time during the initial start-up testing program.
Since start-up, operation at reduced loads has occurred on several occasions as a result of equipment outages, testing, back pressure considerations, etc.
It has been found that the plant can operate for prolonged periods in a stable condition at 50% power with the feedwater flow control in the automatic mode.
Plant operating history has shown that the secondary side stability and efficiency decrease as plant load decreases.
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C.4 Submit a schedule to implement the ATWS instrument modifi-cation justified in accordance with the Westinghouse analytical results contained in the letter from T.N. Anderson to S.H. Hanauer in MS-TMA-2182 dated December 30, 1979.
RESPONSE
The analytical results contained in the December 30, 1979, Westinghouse letter indicate that Indian Point 3 can withstand the consequences of the postulated ATWS events.
There are only two functions that are needed to mitigate the consequences of the most severe ATWS events prior to proceeding to long term shutdown conditions.
These functions are the actuation of the auxiliary feedwater system and the tripping of the main turbine for those events that result in a potential loss of heat sink such as the loss of lead or the loss of main feedwater.
Presently these functions:
are obtained via systems which are postulated to be unavailable during an ATWS, so another method of guaranteeing auxiliary feed-water initiation and main turbine trip will be installed.
The new method referred to as AMSAC (Alternative Mitigating Systems Actuation Circuitry) will be independent of the trip system and unaffected by a common mode failure in the reactor protection system.
Actuation of the auxiliary feedwater system and tripping of the main turbine will be accomplished for those transients identified in the December 30, 1979 Westinghouse letter which are applicable to Indian Point 3.
We believe that the AMSAC instrument modification can be imple-j mented using the following schedule January 1981 Complete Review and Engineering Design April 1981 Complete Purchasing of Equipment April 1982 Complete Delivery of Hardware Fall 1982 Complete Installation During Cycle 4/5 Refueling Outage.
The above schedu:e is based on the assumptions and criteria used in the Westinghouse letter of December 30, 1979, referenced above.
i Should there be any sifnificant changes in the scope of the ATWS instrument modification, as finally approved by the NRC, then this schedule may be extended. I
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C.5 Examinn methods of catablishing the highest reliability for tho l
gas turbines and submit the results to the NRC.
The licensee specifically shall:
1 (1) Provide details of gas turbine controls, modes of operation, and other relevant information; (2) Evaluate possible improvements to the starting and running reliability of the gas turbines; e
(3) Evaluate and initiate actions which will ensure that a gas turbine can be brought on line within one hour after loss of off-site power; l
(4) Determine how gas turbine power can be provided to Indian i
Point Unit 3; and (5) Evaluate the limitation that Indian Point Unit 2 not be operated if the gas turbines are out-of-service.
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RESPONSE
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Indian Point Unit No. 3 is not equipped with gas turbines, therefore, the responses to items C.5. (1), (2), (3) and (5) will be provided by Consolidated Edison since Indian Point Unit No. 2 is equipped with gas turbines.
With regard to item C.5.(4) the followint response is provided:
Similar to Indian Point Unit 2, gas turbine power can be provided to Indian Point Unit 3 from any of the three gas turbines via either of the two 13.8 Kv underground feeders l
or two 13.8 KV over head feeders which connect off-site power to the unit.
Maximum flexibility of routing is provided by interties at the Buchanan Substation (138 KV and 13.8 i
KV buses) and at the Indian Point site (138 Kv site l
switchyard and gas turbine substation 6.9 KV bus tie).
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C.6 Establish an on-site group reporting to offsite management.
The function of the group shall be to examine plant operating characteristics, NRC bulletins, Licensing Information Service advisories and other appropriate sources which may indicate areas for improving plant safety.
Where useful improvements can be achieved, the group shall also develop and present detailed recommendations for revised procedures, equipment modifications or other improvements.
RESPONSE
An On-Site Safety Review Group has been established to perform the functions indicated above.
This group consists of three (3)
Con Ed employees and three (3) Authority employees.
Each group of three consists of one (1) Senior Engineer with approximately 5 to 7 years brhnicd experience and 2 other individuals each with approximately 1 to 2 years bedrdcal experience.
t The group will operate as a single committee of 6 with the Chairman of the committee alternating between the Consolidated Edison Senior Engineer and the Authority Senior Engineer.
The Senior Engineer shall report to the Power Authority Senior Vice President-Nuclear Generation and the Consolidated Edison Vice President, Power Generation.
Reports and recommendations issued by the on-site committee shall be jointly distributed between Consolidated Edison and the Authority.
Events occurring on either unit will be reported to both management organizations, i
Tasks applicable to both plants shall be reviewed by the full 6-member committee with subsequent plant specific actions monitored by the respective plant group of 3.
The Authority will have a group of 3 in place by April 11, 1980.
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Detailed procedures governing the operation of the group will be developed in conjunction with Consolidated Edison.
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