ML20081C242

Proposed Tech Specs Re Limiting Conditions for Operation, Surveillance Requirements & Administrative Requirements

ML20081C242
Person / Time
Site:	Fort Calhoun
Issue date:	03/09/1984
From:	OMAHA PUBLIC POWER DISTRICT
To:
Shared Package
ML20081C225	List: ML20081C222 ML20081C239 ML20081C242
References
RTR-NUREG-0737, RTR-NUREG-737 NUDOCS 8403120195
Download: ML20081C242 (14)

Text

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T TECHNICAL SPECIFICATIONS TABLE OF CONTENTS Pay DEFINITIONS ........................................................ T

.................... 1-1 1.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM Safety Limits - Reactor Core ......................... 1-1 1.1 Safety Limit, Reactor Coolant System Pressure ....... 1-4 1.2 1.3 Limiting Safety System Settings, Reactor Protective System ............................................. 1-6 LIMITING CONDITIONS FOR OPERATION ........................... 2-0 2.0 2.0.1 General Requirements ........................ 2-0 2.1 Reactor Coolant System ............................... 2-1 2.1.1 Operable Components ......................... 2-1 2.1.2 Heatup and Cooldown Rate .................... 2-3 2.1.3 Reactor Coolant Radioactivity ............... 2-8 2.1.4 Reactor Coolant System Leakage Limits ....... 2-11 2.1.5 Maximum Reactor Coolar.t Oxygen and Halogens concentrations ............................ 2-13 2.1.6 Pressurizer and Steam System Safety Valves .. 2-15 2.1.7 Pressurizer Operability .................... 2-16a 2.1.8 Reactor Coolant System Vents ................ 2-16bl 2.2 Chemical and Volume control System ................... 2-17 Emergency Core Cooling System ........................ 2-20 2.3 Containment Cooling .................................. 2-24 2.4 Steam and Feedwater Systems .......................... 2-28 2.5 2.6 Containment System ................................... 2-30 Electrical Systems ................................... 2-32 2.7 Refueling Operations ................................. 2-37 2.8 Radioactive Matericla Release ........................ 2-40 2.9 2.10 Reactor Core ......................................... 2-48 2.10.1 Minimum conditions for Criticality ......... 2-48 2.10.2 Reactivity Control System and Core Physics I Parameter Limits .......................... 2-50 2.10.3 In-Core Instrumentation ..................... 2-54 ng 2.10.4 Power Distribution Limits ................... 2-56 g

2.11 Containment Building and Fuel Storage Building Crane . 2-58 2.12 Control Room Systems ................................. 2-59 m Nuclear Detector Cooling System ...................... 2-60 2.13 I* 2.14 Engineered Safety Features System Initiation Instru-mentation Settings ................................. 2-61

@ Instrumentation and Control Systems .................. 2-65 2.15 2-71 2.16 River Level .---~m.~~.~~~~~~~~~~--.

35 Ita. 2.17 Miscellaneous Radioactive Material Sources ........... 2-72 Shock Suppressors (Snubbers) ......................... 2-73 2.18 2.19 Fire Protection System ............................... 2-89 Steam Generator Coolant Radioactivity ............... 2-96 2.20 2.21 Post-Accident Monitoring Instrumentation ............. 2-97 l Acendment No. 32,38,52,54.'57,67 i ATTACHMENT A

TABLE OF CONTENTS (Cont'd)

Page 5.9 Reporting Requirements ................................ 5-10 5.9.1 Routine Reports ............................... 5-10 5.9.2 Reportable Occurrences ........................ 5-12 5.9.3 Special Reports ............................... 5-15 5.9.4 Unique Reporting Requirements ................. 5-15 5.10 Records Retention ..................................... 5-18 5.11 Radiation Protection Program .......................... 5-19 5.12 Environmental Qualifications .......................... 5-20

'5.13 Secondary Watcr Chemistry ............................. 5-20 5.14 Systedes Integrity ..................................... 5-21 5.15 Iodine Monitoring ..................................... 5-21 5.16 Sampling and Analysis of Plant Effluents .............. 5-21 l INTERIM SPECIAL TECHNICAL SPECIFICATIONS .................... 6-1 6.0 6.1 Limits on Reactor Coolant Pump Operation .............. 6-1 6.2 Use of a Spent Fuel Shipping Cask .................... 6-1 6.3 Auxiliary Feedwater Automatic Initiation Setpoint ..... 6-1 6.4 Operation With Less Than 75% of Incore Detector Strings Operable .................................... 6-1 iii Amendment No. 57, )$, $3, $$, $$, $7, 73 l

E 2.0 LIMITING CONDITIONS FOR OPERATION Reactor Coolant System (Continued) 2.1 2.1.8 Reactor Coolant System Vents Applicability Applies to the status of the reactor coolar.i gas vent system. This specification is applicable while in modes 1, 2 , or 3.

Objective To ensure capability of venting non-condensible gases from the reactor coolant system, the following gas vent system requirements must be met:

(1) At least one reactor coolant system vent path con-sisting of at least two valves in series powered from emergency buses shall be OPERABLE and closed at each of the follouing locations:

a. Reactor vessel head.

b. Pressurizer steam space.

(2) With one of the above reactor coolant system vent paths inoperable, startup and/or power operation may continue provided power is removed from the valve actuators of all the inoperable valves- re-store the inoperable vent path to OPERABLE status within 30 days or be in HCT STANDBY within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

(3) With both of the above reactor coolant system vont paths inoperable, maintain the inoperable vent path closed with power removed from the valve actu-ators of all the inoperable valves in the inoper-able vent paths and restore at least one ot the vent paths 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

_! be in HOT STANDBY within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUT-DOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

Basis The purpose of this specificatica is to ensure a method and system is available to remove steam and/or non-con-densible gases from the' reactor coolant system, which may inhibit core cooling during natural circulation. The Power Operated Relief Valves are not to be considered a vent _ path for the purpose of this specification.

2-16b

2.0 LIMITING CONDITIONS FOR OPERATION 2.21 Post-Accid,ent Monitoring Instrumentation

', Applicability Applies to post-accident monitoring instrun.entation not in-cluded as part of the Reactor Protective System or Engineered Safety Features. This specification is applicable while in modes 1, 2, and 3.

O[ sctive Tc assure that instrumentation necessary to monitor plant parameters during post-accident conditions is operable or that backup methods of analysis are available.

Specifications Post-accident instrumentation shall be operable as provided in Table 2-9. If the required instrumentation is not oper-able, then the appropriate action specified in Table 2-9 shall be taken.

Basis Post-accident monitoring instrumentation provides infor-

. mation, during and following an accident, which is. considered helpful to the operator in determining.the plant condition.

It is desirable that this instrumentation be operable at all times during operation of the plant. However, none of the post-accident monitors are required for safe shutdown of the plant nor are any control or safety actions initiated by the monitors.

In general, the post-accident monitors provide wide range cap-abilities for parameters which are beyond the range of normal protective and control instrumentation. They also provide re-mote sampling and analysis capability to reduce personnel ex-posure under post-accident conditions. Because the infor-

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mation necessary to assess the effect of an accident (i.e.,

-core damage) can be obtained from other sources and by manual methods, it is not necessary that the post-accident monitors be operable at all times.

2-97 i

r TABLE 2-9 Post-Accident Monitoring Instrumentation Operating Limits Minimum i: * -

Operable Instrument Channels Action

1. ' Containment Wide Range Radiation 2 (a)

-Monitors (RM-091A & B)

2. Wide Range Noble' Gas Stack Monitor RM-003L (Noble Gas Portion Only) 1 (a)

' RM-063M (Noble Gas Porti.on Only) 1 (a)

' RN-063G (Noble Gas Portion Only) 1 (a)

3. Main Steam Line Radiation Monitor 1 (a)

(RM-064) i 4. Containment Hydrogen Monitor 2 (b)(c)

(VA-81A & B) 5.- Containment Water Level 1 (d)

Narrow Range-(LT-599 & LT-600)

Wide Range-(LT-387 & LT-398) 2 (b)(c)

6. Containment' Wide Range Pressure 2 (b)(c)

(a) With'the number of OPERABLE channels less than required by the minimum channels operable requirements, initiate the pre-planned alternate method of monitoring the- appropriate para-meter (s):within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, and p

1.- either restore the inoperable channel (s) to OPERABLE status within 7 days of the event, or

2. prepare and submit a special report to the Commission pursuant to specification.5.9.3 within-14 days follow-ing the event outlining the action taken, the_cause of '

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the inoperability, and-the plans and schedules for re-

storing the system to' OPERABLE status.

6 (b) With.one channel inoperable, restore-the inoperable monitor '

to OPERABLE. status'within 30 days or_be in at least HOT SHUT-DOWN:within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

.(c) With both channels inoperable, restore at least one channel

'to OPERABLE status within 72_ hours or be in at least HOT SHUTDOWN within the next'12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

d.

With the number of OPERABLE channels less than required by (d) the minimum channels: operable requirements, operation may continue until the next cold shutdown, at which time the re-

-quired channel (s)-shall be made operable.

4 2-98

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TABLE 3-3 -

MINIMUM FREQUENCIES FOR CHECKS, CALIBRATIONS AND TESTING .

OF MISCELLANEOUS INSTRUMENTATION AND CONTROLS .

Surveillance Channel Description Function Frequency Surveillance Method

1. Primary CEA Position a. Check S a. Comparison of output data with secondary -

Indication System CEAPIS.

b. Test M b. Test of power dependent insertion limits, ~

deviation, and sequence monitoring 14 systems.'t

c. Calibrate R c. Physically measured CEDM position used to verify system accuracy. Calibrate CEA postion interlocks.

2. Secondary CEA Position a. Check S a. Comparison of output data with primary Indication System CEAPIS.

b. Test M b. Test of power dependent insertion limit, deviation, out-of-sequence, and overlap monitoring systems. 14 "4

'w

c. Calibrate R c. Calibrate secondary CEA position indi-cation system and CEA interlock alarms.

3. Area, Process, and a. Check D a. Normal readings observed and internal Post-Accident Radi- test signals used to verify instrument ation Monitors operation.

b. Test M b. Detector exposed to remote operated radi-ation check source or test signal.

c. Calibrate R c. RM-063L, M, and H and RM-064 - One time a

factory calibration is acceptable pro-vided linearity solid sources are used to check the integrity of the detectors.

RM-091A and B - In situ calibration by electronic signal substitution is accept-able for all range decades above 10 R/hr. In situ calibration for at least one decade below 10 R/hr shall be by means of calibrated radiation source.

All other monitors - Exposure to known radiation source.

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TABLE 3-3.(Continusd) +

MINIMUM FREQUENCIES FOR CHE'CKS, CALIBRATIONS AND TESTING OF MISCELLANEOUS INSTRUMENTATION AND CONTROLL 3 Surveillance

' !) . Channel Description Function Frequency Surveillance Method ,

'8"

25. Containment-Purge Iso- -Check M Verify valve position using control

@ lation Valves (PCV- room indication. .

'

• 742A,B,C,&D)

E

26. Containment Hydrogen a. Check M a. Comparison of readings from re-

$ Monitors (VA-81A&B) dundant channels.

b. Test g b. Calibrate span /zero using sample gas and check flow rates.

I

c. Calibrate R c. Calibrate using known signals ap-plied to sensors.

w 27. . Containment Water Level

d. Narrow Range (LT-599 a. Check M a. Compare independent level readings.

@ & LT-600)

b. Calibrate R b. Known signals applied to sensors.

Wide Range (LT-387 a. Check M a. Observe normal reading and simulate

& LT-388) full scale reading.

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b. Calibrate R b. Known signals applied to sensors.

28. Containment Wide Range a. Check , M a. Compare independent pressure read-Pressure Indication ings.

b. Calibrate R b. Apply known pressure to sent4 ors.

Q - Quarterly S - Each Shift D - Daily

. M - Monthly A - Annually.

R - 18 Months P - Prior to each startup if not performed within previous week.

PM - Prior to scheduled cold leg cooldown below 300*F; . monthly whenever temperature remaina below 300*F and reactor vessel head is installed.

_- TABLE 3-5 . -

(Continusd) .

USAR Section Test Frequency Reference (Continued) Automatic and/or manual ini- At least once per plant 3a.-10c. 4.

tiation of the system shall operating cycle.

n be demonstrated.

g: 11. Containment Cooling 1. Demonstrate damper action. 1 year, 2 years, 5 years, 9.10

.& Iodine Removal Fuse- and every 5 years there- ,

able Linked Dampers 2. Test a spare fuseable link. after.

ce g 12. Fuel Elements Visually inspect fuel elements During each refueling out- 3

• removed from the reactor. age.

I Diesel Generator Calibrate. During each refueling out- 8.4.3

• 13.

Under-Voltage Relays age.

M

14. Motor Operated Safety Verify the contactor pickup value During each refueling out-Injection Loop Valve at <85% of 460 v. age.

Motor Starters (HCV-Y 311, 314, 317, 320, ls 327, 329, 331, 333,

" 312, 315, 318, 321)

15. Pressurizer Heaters Verify control circuits operation During each refueling out-for post-accident heater use. age.

16. Spent Fuel Pool Re- Test neutron poison samples for Intervals of 1, 2, 4, 7, gion 1 Racks dimensional change, hardness 11, 15, 20, and 25 years change, and neutron attenuation after ir.s talla tion.

change. ,

17. Reactor Coolant Gas 1. Verify all manual isolation During each refueling out-Vent System valves in each vent path are age just prior to plant in the open position. start-up.

2. Cycle each automatic valve During each refueling out-in the vent path through at age.

least one complete _ cycle of full travel from the control room. Verification of valve cycling may be determined by observation of position indi-cating lights.

3. Verify flow through the re- During each refueling out-  !

actor coolant vent system age.

vent paths.

c 5.0 ADMINISTRATIVE CONTROLS 5.14 Systems Integrity

, A program to reduce leakage from systems outside containment that would or could contain highly radioactive fluids during a serious transient or accident to as low as practical levels shall be inplemented. This program shall include the follow-ing:

1. Provisions establishing preventive maintenance and periodic visual inspection requirements, and

2. Integrated leak test requirements for each system at a frequency not to exceed refueling cycle intervals.

5.15 Iodine Monitoring A program which will ensure the capability to accurately de-termine the airborne iodine concentration in vital areas under accident conditions shall be implemented. This program shall include the following:

1. Training of personnel,

2. Procedures for monitoring, and

3. Provisions for maintenance of sampling and analysis equipment.

5.16 Sampling and Analysis of Plant Effluents A program which will ensure the capability to obtain and analyze radioactive iodines and particulates in plant gaseous effluents under accident conditions shall be in-plemented. The program shall include the following:

1. Training of personnel,

2. Procedures for sampling and analysis, and

3. Provisions for maintenance of sampling and analysis equipment.

I FORT CALHOUN 5-21 Amendment No. 57 h

o .

DISCUSSION, JUSTIFICATION, AND SIGNIFICANT HAZARDS CONSIDERATIONS 1 The proposed amendment vill establish Limiting Conditions for Oper-ation and surveillance requirements for the following systems install-ed as a result of post-TMI requirements and guidance provided by ,

NUREG-0737. l Reactor Coolant System Vents:

Additional wording is proposed in Section 2.0 of the Technical Speci-fications concerning Limiting Conditions for Operation which requires certain actions to be taken in the event of inoperability of specific The new specifi-portions of the reactor coolant gas vents system.

cation requiresta vent path consisting of at least two valves in series and powered from emergency buses be operable and closed at both the reactor vessel head and the pressurizer steam space. Provi-sions are made in this specification to allow plant operation co con-If the tinue up to 30 days with one of these vent paths inoperable.

vent path operability cannot be restored within that period of time, the reactor must be placed in hot standby within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and cold shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. The specification also states that if both of the reactor coolant system vent paths are inoperable, power operation may continue up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> as long as the vent paths are closed and power has been removed from the valve actuators of all of the inoperable valves in the inoperable vent paths. If one of the vent paths cannot be restored to operability within the 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, the reactor must be placed in hot standby con-dition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and a cold shutdown condition within the fol-lowing 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. With the exception of the time allowed to reach a hot standby condition, this proposed specification is consistent with the recommended Limiting Conditions for Operation delineated by the NRC in Generic Letter 83-37 on November 1, 1983. The time allowed to reach a hot standby condition was proposed to be 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> instead of the recommended 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. The proposed 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> limit is consistent with existing Limiting Conditions for Operation time requirements and will allow for a safer, more controlled power decrease within a reasonable amount of time.

An addition is being proposed to Section 3.0 of the Technical Specifi-cations concerning surveillance requirements. The proposed changes establish surveillance requirements for the reactor coolant gas vents These surveillance requirements en-system on a refueling frequency.

sure that all manual isolation valves in each vent path are verified to be in the open position prior to plant startup from a refueling outage, all automatic valves in the vent paths are cycled through at least one complete cycle of full travel from the control room during a refueling outage, and a flow test is performed on the system during each refueling outage.- These surveillance requirements are con-sistent with the requirements delineated in Generic Letter 83-37, with the following exception:

ATTACHMENT B t

u_

1 j

The manual isolation valves in each vent path are not required by  ;

the proposed specifications to be locked inDue the open to the position, location but l only to be verified in the open position. these valves, locking the valves and/or physical shape of some of All of these valves are located is considered to be unfeasible.

the containment which is locked and inaccessible, for all inside The e:aall amount of practical access that purposes, during power operation.is allowed to the containment during p is strictly controlled and for extremely short periods of time.

The District will provide further assurance these valves are open by requiring that the surveillance be performed simultaneously with reactor c.oolant system lineups just prior to plant startup.

The District believes this is adequate to ensure these manual iso-lation valves remain in the open position and will not be inad-vertently closed.

Post-accident monitoring instrumentation, including:

1. Containment Wide Range Radiation Monitor

2. Wide Range Noble Gas Stack Monitor

3. Main Steam Line Radiation Monitor 4.- Containment flydrogen Monitor

5. Containment Water Level Monitor

6. Containment Pressure Monitor Channel operability requirements for the post-accident radiation monitors are provided in proposed Table 2-9 and are generally con-sistent with the recommendations provided in Generic Letter 83-37, dated November 1, 1983. Where alternate means of monitoring are available, the plant may continue operating with less than the re-but must submit a special re-quired number of operable channels; of inoperability, the alter-port natetoactions the NRCtaken, outlining the causeand the plans for restoring the system to oper-This applies to items 1, 2, and 3 above. The Techni-able status. include range and cal Specification guidelines provided by engineered Only the NRC safety feature detpoint for radiation monitors. in and reactor protective system setpoints are presently included Therefore, because no control or

, the Technical Specifications. initiated by the post-accident monitors, the safety functions are range and setpoints for these in themonitors willSpecifications.

Technical be included in the operating manuals and not Where satisfactory alternate methods are not available, the plant required may continue operating temporarily with less than the number of operable channels; but must restore the inoperable channels to operable status within the specified timeThis period or be applies E in at least hot shutdown Thecondition within wide range 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.pressure and containment to items 4, 5, and 6.

level monitors are used only for information and are environmental-ly and seismically qualified (with the exception of the contain-ment wide range pressure recorders which are seismically qualified only); therefore, operation of the plant for 30 days with only one i

channel is justified. The '2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> action to be taken in case of

- failure of both monitors is consistent with the requirements for the containment hydrogen monitor. The action specified for the the narrow range containment water level monitor requires The that narrow system be repaired during the next cold shutdown.

range containment water level monitor is utilized primarilyab- for containment sump pump control and as an aid to help detect normal leakage inside containment. The District believes plant operation should not be contingent upon operability of this moni-toring system to the point where a special shutdewn would be re-quired to repair the system, as it offers no useful post-accident information wh_ich could not be readily obtained by alternate methods (i.e.,' containment sump inventory could be approximated by observing the time it takes to pump down the sump).

Several changes are proposed to Section 3.0 of theTable Technical Speci-3-3, Item fications concerning surveillance requirements.

3, is modified to include post-accident radiation monitors.

This applies to items 1, 2, and 3. Surveillance requirements for con-tainment hydrogen monitors are provided in Table 3-3, Item 26. A monthly channel check will be performed by sampling with each moni-tor and comparing readings. A quarterly channel operability test will be performed for calibrating with a sample gas and verifying flow rates. ;libration of channels will occur at refueling. The

, containment hydrogen monitoring system (CHMS) is maintained in a standby condition. Alarms are provided to alert the operators of channel failures while the CHMS is in this standby condition. To perform a channel check of this system, both trains must be started, containment isolation valves opened, and other necessary equipment actuated (i.e., ca tylist gas applied to the system).

Under norm 31 operating conditions, the minute levels of hydrogen concentration in containment will be undetectable resulting in zero readings at the CHMS panels. The District believes the small amount of useful information to be gained from such a check does not warrant daily operation of this system; therefore, the Dis-tri'ct proposes a monthly channel check to verify proper system It should be operation and a quarterly channel operability test.

noted also that both channels have 100% redundancy and are class lE systems.

In Section 5, a new paragraph 5.'16 entitled " Sampling and Analysis of Plant Effluents" is proposed. This paragraph contains the pro-posed requirement concerning post-accident sampling of plant gaseous effluents.

Significant !>azards. considerations include:

(1) This proposed-amendment does not involve a significant in-crease in the probability or consequences of an accident pre-viously evaluated. The subject systems were designed and in-stalled to mitigate the consequences of a postulated acci-dent. The proposed changes to the Technical Specifications I- -

4- ,

merely require actions which will ensure these systems are

- - capable of performing their intended functions. These pro-posed changes do not alter the design, surveillance require-ments, or operability requirements of any system presently addressed in the existing Technical Specifications.

(2) The proposed amendment does not create the possibility of a new or different kind of accident from any accident previous-ly evaluated. The proposed changes are intended,to provide assurance that the above referenced systems are capable of performing their intended functions. Therefore, the changes ,

.are conservative in nature and are not expected to create the possibility of an unanalyzed accident. These changes do not alter the design, operability requirements, or survnil-lance requirements of any other system which is presently covered by the existing Technical Specifications.

. (3) The proposed amendment does not involve a reduction in a margin of' safety. As mentioned previously, the proposed changes merely assure that newly installed systems perform their intended functions. The change doea not alter the 4

design, operability requirements, Increasing or surveillance require-or establishing ments of.any other plant system.

operability requirements or surveillai.ce requirements of a system does not reduce existing margins of safety.

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I JUSTIFICATION OF FEE CLASSIFICATION j

The proposed amendment is deemed to be Class III, within the mean- -

ing of 10 CFR 170.22, in that it. involves a single safety concern and has been deemed not to involve a significant hazards consider-ations.

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ATTACHMENT C

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