ML20212A066

Proposed Tech Specs Re Steam Line Rupture Detection/ Isolation Sys

ML20212A066
Person / Time
Site:	Fort Saint Vrain
Issue date:	12/19/1986
From:	PUBLIC SERVICE CO. OF COLORADO
To:
Shared Package
ML20212A044	List: ML20212A039 ML20212A056 ML20212A066
References
TAC-60421, NUDOCS 8612220270
Download: ML20212A066 (20)

Text

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ATTACHMENT 2 PROPOSED CHANGES N

0612220270 861219 PDR ADOCK 05000267 P PDR

Fort St. Vrain #1 Technical Specifications Amendment #

Page 4.4-4 Specification LCO 4.4.1 Table 4.4-2 INSTRUMENT OPERATING REQUIREMENTS FOR PLANT PROTECTIVE SYSTE LOOP SHUT 00WN MINIMUM DEGREE PERMIS-MINIMUM OF SIBLE TRIP OPERABLE RE00N- BYPASS

& FUNCTIONAL UNIT SETTING CHANNELS DANCY CON 0!TIONS

!a. Deleted '

lb. Deleted Ic. Deleted Id. Deleted le. Deleted if. Deleted 2a. Deleted 2b. Deleted 2c. Deleted 2d. Deleted 3a. Loop 1 Shutdown ------- ---

2 Logic 1 None 3b. Loop 2 Shutdown -----------

2 Logic 1 None 4a. Circulator 1A and 18 Circulators 2 1 None Shutdown - Loop 1A and IB Shutdown logic Shutdown 4b. Cf reulator IC and 10 Cfreulators 2 1 None Shutdown - Loop IC and 10 Shutdown Logic Shutdowa i

Fort St. Vrain #1 Technical specifications knendment # <

Page 4.4-6a e

spectficatten LCO 4.4.1 Table 4.4-3 (Continued)

PLANT Pia n.iiVE SYSTD8. TRIP SETPOINTS - CIRCULATOS TRIP TR!p ALLOWASLE No. PUNCTIONAL UN!T SETPOINT VALUE 10e. Steam Leek Detection s 52.3 degrees F s 52.8 degrees F Turtine Sutiding per minute per minute Laos 1 rate of rise rate of rise

• 10h. Steam Leak Detection s 52.3 degrees F s 52.8 degrees F Reester Sutiding -

per minute per minute

• Loop 1 rate of rise rate of rise loc. Steam Leek Detection s 52.3 degrees F s 52.8 degrees F Turtine Sutiding per etnute per minute Loop 2 rate of rise rate of rise 10d. Steam Leak Detection s 52.3 degrees F s 52.8 degrees F Reacter Building per minute per minute Loop 2 rate of rise rate of rise I

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i Fort St. Vrain #1

+ Technical Specifications Amendment #

Page 4,4-6b

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g.1 Specification LCO 4.4.1 Table 4.4'-3 (Continued) *(q PLANT PROTECTIVE SYSTEM. OPERABILITY REQUIREMENTS - CIRCULATOR TRIP TOTAL CHANNELS MINIMUM FUNCTIONAL NO. OF TO CHANNELS ACTION NO. UNIT CHANNELS ' TRIP ' OPERABLE APPLICABILITY (NOTE 2) 10a. Steam Leak 4 . ' 2 i 3 Note 1 1,2,3,4l5 Detection  !

s Turbine Building Loop 1 -

s 10b. Steam Leak 4 2 3> 'l' , N te 1 1,2l,'3,4,5 Detection N ,

Reactor Building '

Loop 1 10c. Steam Leak 4 2 3' Note 1 1,P ,3,4 ; 5 -

Detection Turbin.e Building Loop 2  !.

10d. Steam Leak 4 2 3, Note 1 1,2,3,4,5' Detection .

Reactor Building . g .

Loop 2 '

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s Note:

1. The reactor shall not be operated at power (above 2*4 rated thermal power) except as provided oy these'tequirements and their associated ACTION statements.

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2. The requirement to shutdown affected helh:m circulators, provided in the introduction to Specific.ition 4.4.1, does not apply.

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Fort St. Vrain #1 Technical Sp:;cifications Amendment #

Page 4.4-6c Table 4.4-3 (Continued)

ACTION STATEMENTS (for 10a, 106, 10c and 10d Only)

ACTION 1 - With only 7 OPERABLE channels in either building or in both buildings, operation at power may continue provided the inoperable channel is placed in bypass within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or reduce power to below 2% within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The channel shall be returned to OPERABLE status within the following 7 days.

ACTION 2 - With only 6 OPERABLE channels in either building or in both buildings, place the inoperable channels in bypass and reduce power to below 2% within the next 12

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hours. Operation at power may continue if at least 7 OPERABLE channels in both buildings are placed in service.

ACTION 3 - With inoperable channels or loops other than as provided in ACTION 1 and 2 above, reduce power to below 2% within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

ACTION 4 - With any one valve actuated by SLRDIS or electrical wiring and circuits used to actuate that valve inoperable, restore the valve and/or associated components to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or reduce power to below 2% within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

ACTION 5 - With two or more valves actuated by SLRDIS or electrical wiring and circuits used to actuate those valves inoperable, restore the valves and/or associated components to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or reduce power to below 2% within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Fort St. Vrain #1

. Technical Specifications Amendmeest #

Page 4.4-8 Soestftcatten LCO 4.4.1 NOTES FOR TABLES 4.4*1 T1441128 4.4-4 (a) See Specificasten L5553.3 for trip setting.

(t) 1%e theressouples from each leep, total of feve, constitute one chesnel. For each channel, two theressouples must be operetle to at least one operating leep for that chamael to be contidored operatie.

(c) With one arteary coolant high lesel metsture mentter tripped trips of either leep primary coolant solsture sentters will cause, full scree. Hence, number of operable channels (1) etavs states numer required to cause scree (0) oguais one, the statem degree of redundamsy.

(d)Seth 400 volt buses 1A and 1C Tess of voltage for no longer thee

, 35 sesends.

(e) One enemmel coaststs of one unoerveltage relay free oesh of the two 400 volt buses (two underveltage relays per chamael). These relays fat) been which is the dtractlen required to inttiate a seras.

(f) The inesorable channel must be in the tripped conditten, unless the trip of the channel will cause the protective acttee te occur. Failure to trip the inesoratle channel requires taking the appropriate corrective actten as Itsted on ca ges 4.4-1 and 4.4-2 within the spectffed time limit.

(g)RWP bypass permitted if the bypass aise causes assectated single caannel scras.

(h) Permissible Oypass Condittens:

I. Any circulater buffer seal malfunction.

II. Loop het reneet header h1gh activ1ty.

III. As stated in LCO 4.g.2.

(j) Celeted.

(k) One operable heltum etreulator inlet thermoccuole in an coerable loop is required for the enannel to me constcored coeraole.

(e) Low Power RWP bistacle resets at 44 after reactor cwor inf ttally enceses 55.

(n) Power range RWP bistables automatically reset at 105 after reactor power t s cecreased from greater than 3C%. The Rh8 may De manually reset :stween 105 and 305 power. '

(D) Item 74. must :e accomoanied by item 7c for loco 1 shutdown.

Itse 7b. must :e accompanted my f tem 7c for loco 2 snutdown.

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Fort St. Vrain #1 Technical Speciffcations Amendment #

Page 4.4-9 l NOTts FOR TA8LES 4.4-1 through 4.4-4 (Centtaved) r) Separete fastrumentatten is provided se each ctreulater fee this functional unit. Only the affected heltua streulater shall be shut requirements down are not met.within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> if the indicated s) Deleted.

t) A primary coolant dew point metsture monttee shall not he constdered operaale unless the folieving condittees are met:

1) Reacter Powee Ranee Mfatem Sammte Pfew Startus to 25 1 sec/sec. ~

> 25 - 55 5 ses/ses.

> 55 - 205 15 sec/ses.

> 205 - 355 30 sec/sec.

> 355 -1005 50 scc /ses.

2) M1nteus flow of ttee 1) is alarmed in tre control rees and the alare is set in accordance with tse power ranges spectffed.

3) The aattent t,emperatures indicated by both temporary thermecouples asunted on the flow sensers in penetrations 81 and 83 are less than 185'F.

a) Fixed alarms of 1 sec/sec and 75 sec/sec are operaale.

Fort St. Vrafn #1 Technical speciffcations

Amendment #

Page 4.4-12 Mast Eleatrical Systes power loss requires a scras to prevent any ^ n .-

_ flow etsas:ches from occurring. A 30-second delay is provided following a power loss before the scrae is initiated to allow the emergency diesel generater to start. If it does start, the scram is avetded.

Two-Loom Troutie. Operation on one loop at a mestaus of about 55 power may continue following the shutdown of the other loop (unless proceded by scras as in the case of high meisture.)

Onset of trouble in the remaining loop (two-1,oep trouble) results in a scras. Trouble is deffned as a signal which normally inttfates a loop shutdown. Similarly, simultaneous '

thutdown signals to both loops result in shutdown of one of the two loops only and a reactor scras. However, actuation of both steam Line Rupture Detection / Isolation System (SLR0!$) toops, effectively shuts cown both loops because it senos an actuation legte signal to all four circulator tria logie channels. The consequences of a two-loop shutdown and subsecuent loss of forced ctreulation have been analyzed and found to be acceptable. The consequences are bounded by an inter uptten of forced circulation cooling accident described in FSAR Section I

14.4.2.2. Safe Shutdown Cooling.

Hioh Temperature in tne pipe cavity would Indicate the presence of an undetected steam leak. A steam leak or pipe aupture under the PCRV within tre succort ring would also be detectacle in the cine cavity, therefore only one set of sensors and logic ts recutred to monitor botn areas. The setootnt has :een set above the temocrature tnat would to expectcJ to occur t

+n the nice i

cavity if the steam less were detected.

• Fort St. Vrain #1 Technical Specifications Amendment #

Page 4.4-12a b) Loon Shutdown Inputs Shutdown of Both Circulators is a loop shutdown input which is necessary to insure proper action of the reactor protective (scram) system (through the two-loop trouble scram) in the event of the loss of all circulators and low feedwater flow.

The remaining loop shutdown inputs are equipment protection items which are included because their malfunction could prevent a scram due to loss of the two-loop trouble scras input.

c) Circulator Shutdown Inputs l Circulator Shutdown Inputs (except circulator speed hioh on water turoines) are equipment protection items which are tied to two loop trouble through the loop shutdown system. These items are included in Table 4.4-3 because a malfunction could prevent a scram due to loss of the two loop trouble scras input.

Circulator speed high on water turbines is included to assure continued core cooling capability on loss of steam drive.

Steam Leak Detection in the Reactor Buf1 dine is required for equipment qualification of Safe Shutdown Cooling Systems. The ALLOWA8LE VALUE is set at s 52.8 degrees F per minute rate of rise in order to prevent exceeding the harsh environment temperature profile to which the safs shutdown electrical equipment is qualified, per the requirements of 10CFR50.49. A setpoint calculation analysis performed per ISA Standard 567.04

• and RG1.105 results in the stated ALLOWABLE VALUE and TRIP SETPOINT as specified in the LCO and this basis. The TRIP SETPOINT has been established with sufficient margin between the technical specification limit for the process variable and the nominal TRIP SETPOINT to allow for 1) inaccuracy of the instruments; 2) uncertainties in the calibration; 3) instrument drift that could occur during the interval between calibrations; and 4) inaccuracies due to ambient temperature changes, vibration and other environmental conditions. The TRIP SETPOINT is set at s 52.3 degrees F per minute rate of rise until such time as the drift characteristics of the detection system are better understood from actual plant operating experience and the assumptions used in the setpoint analysis are verified.

SLRDIS design incorporates two panels. each with its own set of sensors for the Reactor and Turbine Buildings and dual logic trains in each panel. The SLRDIS design preserves the single failure concept. A single failure will neither cause nor prevent SLRDIS actuation in the event of a high energy line break. The probability of an inadvertent actuatten is extremely small due to the matrix log'ic employed for circulator trip and valve actuation. The SLRDIS panels are referred to as " loops";

however. due to the way the outputs of the panels are combined to provide protective action and satisfy the , single failure concepts the SLRDIS loops do not correspond to primary or secondary loops.

- Fort St. Vrain #1 Technical Specifications Amendment #

Page 4.4-12b For each SLRDIS loop, the OPERABILITY requirements and their respective ACTIONS represent good operating practices and

.fudgment for a four channel detection system with a 2 of 4 coincidence trip logic. The fourth channel may be placed in bypass for test and/or maintenance purposes, subject to the ACTION statement restrictions, while preserving a 2 of 3 coincidence logic OPERABLE. The Steam Line Rupture Detection / Isolation System as designed and installed has spare channels available for input. Any of the available channels may be selected for input signal processing provided the surveillances are current on the channels used. The SLRDIS is required to be OPERABLE only at power (above 2% rated thermal ,

power). Analyses with rated reactor power at 2% demonstrate

! that automatic actuation of SLRDIS is not likely to occur during a high energy line break lasting until it is ma oally terminated at one hour following initiation. The temperatures as analyzed in both the reactor and turbine buildings stay well below the temperature for which the equipment is qualified.

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l The ACTION statements for inoperable SLRDIS detection and

! information processing equipment allow one channel in each j building to be inoperable for up to 7 days; a second inoperable i

I channel in either building requires that power be reduced to below 2% within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The 7 day ACTICN time for a single detector channel is acceptable based on preservation of a 2 out of 3 coincidence detection system still in operation. ACTION 3 is applicable to other functions within the SLRDIS instrumentation panel such as loss of power from instrument buses, or other failures in the logic trains and associated electronics. A 12 hour time period in ACTION 3 for inoperability of those associated SLRDIS functions minimizes the time that SLRDIS may operate with limited functional capability. An inoperable valve or associated equipment is allowed for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. High energy line break analysis for environmental qualification assumes the worst-case single active failure. Thus, a single valve inoperable for up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is within the bounds of analysis. When two or more valves and/or associated equipment is inoperable. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is allowed to restore the inoperable equipment. Repairs may be performed while the plant is at power, thus, minimizing thermal cycling of plant and installed equipment.

Fort St. Vrafn #1 Technical Specifications Amendment #

Page 4.4-13 Steam Laak Detectfen- in the furtitne autidf31s retuired for egulpment qualtffcatten of Safe Shutdown Caeltag Systems. Thus, the lietts and teasts are the same as dfscussed in the basis for

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staas leak detestfen 1a the reacter bei1dfag, d) Red Wfthdraw Frohtbtt Ineuts Startup Channel Countrate-Lew is provided to prevent centrol red withdraus) and reacter startup without adeguato neutree flus fadicatten. The trip logic fs selected to be above the background neise level.

Linear Channel (95 Power) directs the operator's attenties to either a dounscale failure of a power range channel er taproper posittening of the I.S.S.

Linear Channel (305 Power) is provided to prevent centrol red wtthdrawal if reactor power escoeds the !.$.3. ifatt for the

" Low Power

• positien.

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- Fort St. Vrain #1 Technical Sp;cifications Amendment #

Page 5.4-2 Calibration frequency of the instrument channels listed in Tables 5.4.1, 5.4.2, 5.4.3, 5.4,4 are divided into three categories:

passive type indicating devices that can be compared with like units on a continuous basis; semiconductor devices and detectors that may drift or lose sensitivity; and on-off sensors which must be tripped by an external source to determine their setpoint. Drift tests by GGA on transducers similar to the reactor pressure transducers (FSAR Section 7.3.3.2) indicate insignificant long term drift. Therefore a once per refueling cycle calibration was selected for passive devices (thermo-couples, pressure transducers, etc.). Devices incorporating semiconductors, particularly amplifiers, will be also calibrated on a once per refueling cycle basis, and any drift in response or bistable setpoint will be discovered from the test program. Drift of electronic apparatus is not the only consideration in determining a calibration frequency: for example, the change in power distribution and loss of detector chamber sensitivity require that the nuclear power range system be calibrated every month. On-off sensors are calibrated and tested on a once per refueling cycle basis.

The Surveillance Requirements for the Steam Line Rupture Deter. tion / Isolation System instrumentation in Table 5.4-3 include provisions for CHANNEL CHECK, CHANNEL CALIBRATION, CHANNEL FUNCTIONAL TEST and an ACTUATION LOGIC TEST. The frequency of CHANNEL CALIBRATION, REFUELING, is consistent with the interval for testing and calibrating similar detectors (heat sensitive cabling used for fire detection). The manufacturer of the instrumentation recommends an 18 month interval for test / calibration of the electronics portion of the Steam Line Rupture Detection / Isolation System, thus, the CHANNEL FUNCTIONAL TEST is specified for that interval. The ACTUATION LOGIC TEST is specified for a REFUELING interval. The ACTUATION LOGIC TEST verifies proper operation of the SLRDIS Detection and Logic Racks from a simulated rate-of-rise input signal through and including actuation of the output logic relays. Time response of the SLRDIS Detection and Logic Racks is verified to be equal to or less than 7.1 seconds as assumed in the high energy line break analysis. The potential for an inadvertent actuation during testing suggests that logic testing be performed only when the plant is in SHUTDOWN. Thus, the surveillance requirements are specified for REFUELING but not to exceed 18 months. The SLRDIS control unit includes a supervision system that continuously and automatically monitors critical circuitry and internal components, and alarms SLRDIS trouble conditions to the operators.

Fort St. Vrain #1 Technical Specifications Amendment #

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i Fort St. Vrain #1 Technical Sp;cifications knendment #

Page 5.4-10a Table 5.4-3 (Continued)

CIRCULATOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL ACTUATION FUNCTIONAL CHANNEL CALIBRA- FUNCTIONAL LOGIC NO. UNIT CHECK TION TEST TEST ASPLICABILITY 8a. Steam Leak D R(a) R(a) R(b) At Power

• Detection Turbine Building Loop 1 8b. Steam Leak 0 R(a) R(a) R(b) At Power

• Detection Reactor Building Loop 1 8c. Steam Leak D R(a) R(a) R(b) At Power

• Detection Turbine Building Loop 2 8d. Steam Leak D R(a) R(a) R(b) At Power

• Detection Reactor Building loop 2 Notes to lines 8a. through 8d. above only:

(a) The calibration / test consists of verifying the rate of rise setpoint and checking for opens and shorts in the sensor cable.

(b) The SLRDIS Detection and Logic racks shall be verified to have a response time less than or equal to 7.1 seconds when a simulated rate-of-rise trip input signal is used to actuate the output relay logic.

R - At least once per Refueling cycle, not to exceed 18 months.

Applicable only above 2% RATED THERMAL POWER.

0 - Daily.

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• ATTACHMENT 3 SIGNIFICANT HAZARDS CONSIDERATION 1

SIGNIFICANT HAZARDS CONSIDERATION <

I. EVALUATION Based upon PSC Safety Analysis Report (EE-EQ-0014), Steam Line Rupture Detection / Isolation System (SLRDIS), it is concluded that the SLRDIS is capable of performing its intended function to detect and isolate major secondary coolant line ruptures of high energy steam pipe lines of the secondary cooling system without operator intervention. The resulting harsh environments from a SLRDIS terminated leak are less severe than the harsh environment previously established on the basis of the operator manually terminating the leak at 4 minutes. These operator terminated leaks previously established the harsh environments used for Fort St. Vrain equipment qualification.

The SLRDIS replaces the existing steam pipe rupture detection system. SLRDIS is designed to detect steam leaks in cither the reactor or turbine buildings and isolate those leaks to preserve or maintain an environment in which electrical equipment is qualified. This change in steam leak detection and automatic action provides more inclusive coverage of potential steam leaks and results in no change in the radiciogical consequences. For certain steam leaks, the SLRDIS System f.as a slower response than the existing sy s ter., but analysis demonstrates that equipment qualification is naintained. Thus, the existing steam pipe ruptuee detection f.ystem will no longer be required.

Manual operator intervention for isolating HEles in the feedwater, condensate and extraction steam systems and those line breaks not isolated by SLRDIS is adequate to assure that the resulting temperature profiles are enveloped by that to which the equipment will be qualified.

1. Consequences of other accidents analyzed in the FSAR were examined for adverse impact as a result of the installation of the SLRDIS. Design Basis Accident No. 2,

" Rapid Depressurization/ Blowdown Accident", was determined to have one assumption invalidated in that the SLRDIS could prevent initiation of forced circulation cooling at 5 minutes into the accident. Reanalysis of the accident determined that forced circulation cooling could be delayed for at least 60 minutes without exceeding the conservative FSAR temperature for onset of fuel particle failure of 2900 degrees F. a temperature well below that at which rapid fuel deterioration is expected to occur. This is more than ample time for the operator to restore forced circulation cooling.

- - - - - - ~ - . - - , - _ . - . - - . - , - , , - - - - - , -- - . , - , ,

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2. The potential for the SLRDIS to create new or different types of accidents not previously analyzed was examined.

The conclusion was that SLRDIS actuation initiated by a high  !

energy line break, fir,es or primary ::oolant leaks may result  !

in an interruption of forced circulation cooling. It was '

further concluded that sufficient information is currently available in conjunction with new information available from the SLRDIS for the operator to properly diagnose and recover from the event by re-establishing forced circulation cooling within 60 minutes. The 60 minute forced circulation recovery is the most limiting recovery time and is associated with DBA-2. For all other accidents, 90 minutes is adequate time to restore forced circulation cooling prior to reaching 2900 degrees F. Surveillance testing or faults associated with a single panel of the SLRDIS will not cause SLROIS actuation.

3. A review was conducted to determined if any margins of safety defined in the basis for a Technical Specification or in the .75AR were significantly decreased. It was concluded that the SLRDIS isolating the secondary coolant system only causes a temporary interruption of forced circulation cooling in both loops. A recovery methodology exists and a recovery procedure will be developed to reestablish forced circulation cooling within the most limiting time associated with the DBA-2 accident - 60 minutes. Thus, it is concluded that the margin of safety is not significantly reduced.

II. CONCLUSION Based on the above evaluation, it is concluded that operation of Fort St. Vrain in accordance with the proposed changes will not (1) involve a significant increase in the probability or consequences of an accident previously evaluated, (2) create the possibility of a new or different kind of accident from any accident previously evaluated, or (3) involve a significant reduction in any margin of safety.

Therefore, this change will not create an undue risk to the health and safety of the public nor does it involve any significant hazards consideration.

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4 Attachment 4 EXAMPLES OF SUPPLEMENTAL CHANGES TO TECHNICAL SPECIFICATIONS THAT DID NOT REQUIRE ADDITIONAL NOTICE ,

The Technical Specification amendment process often deals with issues that cannot be satisfactorily resolved with a single' amendment submittal. Concerns are raised that require additional changes to supplement the initial submittal, and these are often not made the subject of an additional notice for public comment.

PSC considers the changes reflected in the SLRDIS Technical Specifications in Attachment 2 to this letter, to be clarifications and enhancements. They are not significantly different from the specifications that were covered by the NRC's original determination of no significant hazards cor. sideration, and PSC requests that they not be made the subject of additional notice. Examples of similar

, amendment proposals from other licensees are as follows:

1. Pennsylvania Power and Light Company, Susquehanna Steam Electric Station, Units 1 and 2, Amendment Number 56 to NPF-14 and A;nendment Number 26 to NPF-22.

In this issue, two Technical Specification Change requests were submitted after the publication of the proposed In the SER dated

determination of no significant hazards.

April 11, 1986, the NRC explicitly stated that the

! supplemental submittal was essentially covered by the original notice of proposed action- and need not be the subject of an additional notice.

2. Pennsylvania Power and Light Company, Susquehanna Steam Electric Station, Units 1 and 2, Amendment Number 60 to NPF-14 and Amendment Number 30 to NPF-22.

In this issue, a revised submittal was made for clarification, after the publication of the proposed determination of no significant hazards consideration and no additional notice was required.

3. Commonwealth Edison Company, LaSalle County Station, Units 1 and 2, Amendment Number 39 to NPF-11 and Amendment Number 21 to NPF-18.

In this issue, a second submittal was made by the licensee, after the publication of the proposed determination of no significant hazards consideration, and an additional notice was not required.

_ _ .