ML20087N309

From kanterella
Revision as of 12:27, 25 September 2022 by StriderTol (talk | contribs) (StriderTol Bot change)
(diff) ← Older revision | Latest revision (diff) | Newer revision → (diff)
Jump to navigation Jump to search
Resubmitted Application to Amend License NPF-11,revising Unit 1 Tech Specs to Reflect Changes Incorporated Into Unit 2 Tech Specs.Original Submittal of Change Request NPF-11/84-02 on 840113 Rejected by NRC
ML20087N309
Person / Time
Site: LaSalle Constellation icon.png
Issue date: 03/22/1984
From: Schroeder C
COMMONWEALTH EDISON CO.
To: Harold Denton
Office of Nuclear Reactor Regulation
Shared Package
ML20087N310 List:
References
8335N, NUDOCS 8404030326
Download: ML20087N309 (15)
v • d • e


Text

'N Commonwe:lth Edison

) One First Nationil Plata. Chicago, liknots

^J3 ~7 Address Reply to. Post Off>ce Box 767 Ng' Chicago, Illinois 60690 March 22, 1984 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555

Subject:

LaSalle County Station Unit 1 Revised Request for Change to W F-ll Technical Specifications Regarding Upgrade to Reflect Unit 2 Technical Specification Changes NRC Docket No. 50-373 Reference (a): LaSalle County Station Unit 2 license WF-18 dated December 16, 1983 (b): C. W. Schroeder letter to H. A. Denton dated January 13, 1984.

Dear Mr. Denton:

The purpose of this letter is to resubmit the following change in technical specifications for LaSalle County Station Unit 1.

Change request WF-11/84 Revise Unit 1 Technical Specifications to reflect changes incorporated into Unit 2 Technical Specifications.

This change was originally submitted by Refernce (b). The NRC rejected Reference (b) and requested that Commonwealth Edison Company resubmit this request with additional information.

This proposed change is addressed in Attachment A and has received Onsite and Offsite review and approval.

Commonwealth Edison Company has reviewed this revised amendment request and has determined that no significant hazard consideration exists.

Our review is documented in Attachment B. s Pursuant to 10 CFR 170, this change reflects one example of a Class II amendment. This change has recently (12-16-83) been reviewed, approved tnd issued by the NRC on LaSalle County Station Unit 2, WF-18. Although it does not qualify for a Class I amendment because the review is not exactly concurrent with Unit 2 the NRC staff has been aware of Commonwealth Edison Company's intention to upgrade the Unit 1 Technical Specifications to reflect changes made in the Unit 2 specifications. As such, a remittance of $1,200.00 was provided in Reference (b). ,

c-8404030326 840322 1  % i PDR ADOCK 05000 ( hl P

l l

l H. R. Denton March 22, 1984 l To the best of my knowledge and belief the statements contained herein are true and correct. In some respects these statements are not based on my personal knowledge but upon information furnished by other Commonwealth Edison and contractor employees. Such information has been reviewed in accordance with Company practice and I believe it to be reliable.

Commonwealth Edison Company is notifying the State of Illinois of our request for this amendment by transmittal of a copy of this letter and its attachments to the designated State official.

If there are any questions concerning this matter please contact this office.

Enclosed please find three (3) signed originals and twelve (12) copies of this letter and the attachments. The number of copies to be submitted was discussed with Dr. A. Bournia on March 22, 1984.

Very truly yours, Wrh%

C. W. Schroeder Nuclear Licensing Administrator 1m-Enclosures cc: Dr. A. Bournia - 1 Hand Deliver NRC Resident Inspector - LSCS G. Wright - State of Illinois SUBSCRIEED ~AND' SWORN to before me this 421rdday of Mutv] i , 1984 v twn $f 0- on$2d Notary Public 8335N

ATTACHMENT A March 22,1984 LAEALLE COUNTY STATION UNIT 1 TECHNICAL PECIFICATION CHANGE REQUEST

SUBJECT:

Update of Unit 1 Technical Specifications to the Issued Unit 2 Technical Specifications for LaSalle County Station

REFERENCES:

1) Standard Technical Specifications for General Electric Boiling Water Reactors - NUREG 0123 revision 3.
2) C. W. Schroeder letter to H. R. Denton dated July 22, 1983.
3) C.. W. Schroeder letter to H. R. Denton dated August 23, 1983.
4) C. W. Schroeder letter to A. Schwencer dated September 18, 1982.
5) .A. Schwencer letter to D. L. Farrar dated April 11, 1983.
6) C. W. Schroeder letter to H. R. Denton dated October 14, 1983, LaSallE County Station Unit 1 In-plant SRV Test - Final Evaluation Report.
7) C. W. Schroeder letter to H. R. Denton dated November 7, 1983, Primary Containment Vent and Purge Isolation Valve Closure Times.
8) C. W. Schroeder letter to H. R. Denton dated November 23, 1983, Fire Protection.
9) Federal Register Notice 48FR31611 - Final rule on licensed operator staffing at nuclear power plants.
10) C. W. Schroeder letter to H. R. Denton dated August 31, 1983.
11) C. W. Schroeder letter to H. R. Denton dated Dctober 13, 1983, Bypass Valve Position / Flow Determination. -

BACKGROUND The Operating License (NPF-18) for LaSalle County Station Unit 2 was issued on December 16, 1983. The Technical Specifications for LaSalle County Unit 2 differed somewhat from that issued and as amended through Amendment 15 for Ldialle County Unit 1. Commonwealth Edison committed to update the Unit 1 Tcchnical Specifications to that issued for Unit 2 where the changes did not involve specific design differences. This submittal includes those differences as identified by Commonwealth Edison. I

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

DISCUSSION A. The following is a list of the changes to the Unit 1 Technical Specification. Thic list is based on the Unit 2 Technical Specification as issued and on the current Unit 1 Specification:

1. Page XIX through XXIII - added list of Tables and Figures.
2. Page 1-9, *** footnote states moved instead of coupled.
3. Single recirculation loop operation is folded into the body of the Tech Specs, pages 2-1, 2-4, B 2-1, B 2-4, 3/4 2-1, 3/4 2-3, 3/4 2-4, 3/4 3-53, 3/4 4-1, la, 2, 3, B 3/4 1-2, 8 3/4 2-1, B 3/4 2-3, B 3/4 4-1.
4. Page 3/41 revised 4.1.1.c to allow 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> instead of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
5. Page 3/41 added word " withdrawn" to action b.l.a)l).
6. Page 3/4 1 added footnote to allow startup to perform test if necessary.
7. Pages 3/4 1-6, 8, 9, 14 - control rods specifications 3.1.3.2. ,

3.1.3.4, 3.1.3.5., and 3.1.3.7 have added "3.0.4 not applicable".

8. Pages 3/4 3-4, 5, 41 - deleted startup test setpoint verification footnote.
9. Pages 3/4 3-11,14 - added footnote (1) to allow bypass of delta T instruments for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
10. Page 3/4 3 revised reactor water cleanup ambient and differential temperature setpoints,
11. Pages 3/4 3-15,16,17 - deleted startup test setpoint verification footnote.
12. Pages 3/4 3-18, 19 - added greater than or equal to 5 second time delay reference.

l 13. Pages 3/4 3-39, 3/4 2-4, 5 - added change to allow operation if EOC-RPT inoperable.

14. Page 3/4 3 revised APRM calibration frequency to semi-annually.
15. Pages 3/4 3-60, 63 - added footnote for common systems.
16. Pages 3/4 3-72 ar.d 1/4 9 added footnote to require S/N greater than 2 for source range to be less than 3 cps.

p

17. Pages 3/4 3-81, 3/4 11-13, 3/4 11 revised radioactive effluent reporting requirements.
18. Page 3/4 3 specified that isolation of the off gas system was required only during channel calibration.
19. Page 3/4 4 there is no requirement for immediate scram upon loss of both recirc pumps.
20. Page 3/4 4 revised tolerance from +1% to + or - 1% for SRV's.
21. Page 3/4 4 revised tolerance from + or - 10 to + or - 50 psig.
22. Page 3/4 4 revised withdrawal times for reactor vessel material specimens.
23. Pages 3/4 5-3, 4, 5 - water tight doors specifications for ECCS corner rooms have been added.
24. Page 3/4 5 revised HPCS delta P setpoint.
25. Page 3/4 5 deleted footnote for startup test setpoint verification.
26. Pages 3/4 6-2, 3 - revised calculation method of MSIV leakage rate limit.
27. Pages 3/4 6-8, 9 - revised tendon action statements and surveillance and table clarifications.
28. Pages 3/4 6-15 and 3/411 added limitation on using standby gas treatment for purging the primary containment.
29. ' Pages 3/4 3-70, 3/4 6-16,17,18 - deleted SRV test footnote.
30. Table 3.6.3-1. (Pages 3/4 6-24, 26, 27, 28, 32, 34)
a. (PCIS valves) added # (3.0.4 not applicable) to various valves.
b. Revised butterfly valve times after first refuel outage. (Also some VQ velve times have changed immediately.)
c. 1Cm238 & 2CK)24A have been deleted.
d. 1E12-F099B added.
31. Page 3/4 7 deleted footnote allowing crosstie of 250 volt batteries.
32. Page 3/4 7 revised fire pump parameters and fire suppression water system pressure.
33. Page 3/4 7 (Table 3.7.7-1) revised area temperatures.
34. Pages 3/4 7-27 throutf 45 - revised total snubber spec. - No Table.
35. Page 3/4 7 deleted calibration requirement for Main Turbine Bypass System and deleted valve position requirement. Also changed applicability to l 25% power. Deleted startup test footnote.
36. Diesels:
a. Page 3/4 8 added explanation to 2A inoperable action f. to prevent excessive 1A testing when system inoperable.
b. Page 3/4 8 deleted old surveillance item 6.
c. Page 3/4 8 deleted starts on stored air surveillance (old item 13).

I

37. Page 3/4 8 changed 'and/or' to 'or' for 0 and 1A during shutdown.
38. Page 3/4 8 revised equipment needed for Unit 2 Division 1 AC.
39. Page 3/4 8 revised 'and/or' to 'or' for AC during shutdown.
40. Pages 3/4 8-14, 15, 17, - deleted Unit 2 Division 1 DC sources and deleted ability to crosstle.
41. Page 3/4 8 revised 'and/or' to 'or' for DC sources during shutdown.
42. Page 3/4 8 added drywell hoists and cranes to drywell circuits to be deenergized and deleted them from page 3/4 8-24.
43. Page 3/4 8 added "3.0.4 not applicable" to thermal overload bypass specification.
44. Page 3/4 8 deleted valve IVQ041 from thermal overload bypass table, j 45. Page 3/4 8 revised requirements to functionally test the RPS EPA's to only during cold shutdowns greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
46. Page 3/411 deleted P-32 from liquid waste sampling Table 4.11.1-1.
47. Page 3/411 revised requirement (sampling).
48. Page 3/412 revised number of sample locations (Table 3.12.1-1) .
49. Page B 3/4 0 added clarification of time use during required shutdowns.
r.m
50. Page B 3/4 5 revised HPCS pump flow basis.
51. Page B 3/4 6 clarified tendon surveillance basis.
52. Pages B 3/4 7-3, 4, 5 - revised snubber basis.
53. Pages B 3/4 11-1, 3 - clarification to radioactive effluent basis.
54. Page 5 corrected drywell free volume.
55. Page 6 revised corporate management figure.
56. Pages 6-13,14 - new shift manning Table for two units.
57. Pages 6-28, 29 - added footnote for common PCP and ODCM.

t I

l i

B. The following is a list of pages on which the change was a minor administrative change (eg. typo; adding comma, parenthesis, etc; wording change for clarification; etc):

1. Pages II,.VIII, XV,
2. Pages 3/4 1-4, 1-11, 1-19.
3. Pages 3/4 3-1, 3/4 3-58, 3/4 3-82, 3/4 3-83, 3/4 3-84.
4. Pages 3/4 4-13, 3/414, 3/417, 3/4 23, 3/4 24.
5. Page 3/4 5-8.
6. Pages 3/4 6-5, 3/4 6-11, 3/4 6-19, 3/4 6-20, 3/4 6-21, 3/4 6-33, 3/4 6-35, 3/4 6-36, 3/4 6-37, 3/4 6-38, 3/4 6-40, 3/4 6-41.
7. Pages 3/4 7-14, 3/4 7-17, 3/4 7-18, 3/4 7-22, 3/4 7-24.
8. Pages 3/4 8-1, 3/4 8-5, 3/4 8-7, 3/4 8-9, 3/4 8-16.
9. Pages 3/4 9-16, 3/4 9-17.
10. Page 3/4 11-9.
11. Pages 3/412-1, 3/412-4.
12. Pages B 3/4 1-1, B 3/4 1-5.
13. Page B 3/4 3-6.
14. Page B 3/4 12-1.
15. Pages 6-3, 6-20.

h ATTACHE NT B SIGNIFICANT HAZARDS CONSIDERATION Commonwealth Edison has evaluated the proposed Technical Specification Amendment and determined that is does not represent a significant hazards consideration. Based on the criteria for defining a signficant hazards consideration established in 10 CFR 50.92, operation of LaSalle County Station Unit 1 in accordance with the proposed amendment will not:

1) Involve a signficant increase in the probability or consequences of an accident previously evaluated because the changes involved in this request represent previously reviewed and approved issues which are already present in the Technical Specification' issued for LaSalle County Station thit 2, These issues are discussed in detail below.
2) Create the possibility of a new or different kind of accident from any accident previously evaluated because no new changes not already incorporated in the issued Unit 2 Technical Specifications are included in this change.
3) Involve a significant reduction in the margin of safety because all the changes are as previous issued in the Unit 2 Technical Specifications.

These changes are acceptable and meet the above criteria for the reasons listed below. The justifications refer to items previously refered to under Discussion Section A

1) Editorial change only.
2) The revised footnote allows additional control rod testing required by the

~ Technical Specifications in addition to coupling, such as RPIS, RMCS etc.

This is required for example after maintenance prior to startup with the one rod out interlock in effect. With only one rod out shudown margin is i maintained. This therefore, does not effect the margin of safety.

l 3) Single recirculation loop operation was approved under Unit 1 license (tPF-ll) Amendment #11 and this revision simply moves the requirements

into the applicable Technical Specification sections.
4) This extended time was suggested by NRR and provides the time necessary to allow the surveillence to be performed.

L 5) This change was part of LRG-I Tech Spec issues and corrects action b.l.A) l, since if the inoperable control rod is not withdrawn, its safety function is met and need not be considered when determining separation from other withdrawn inoperable control rods. The limit of a maximum of 8 control rods inoperable is still maintained.

l i

I

'jy l 6) This surveillance requirement requires that control rod density be "less .

than or equal to 505." This would require allowing the reactor to startup l to perform this test. The addition of this footnote provides for allowance to startup'and perform the test within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after achieving the required conditions. This allowance is similar to that for Safety / Relief Valves (4.4.2.1.b) .

7) These changes were part of LRG-I Tech Spec issues and allow for startup.

(Mode Changes) with various control rod parameters not meeting the LCO if the Tech Spec required action is followed. With the Technical i

!- Specification action taken the plant is maintained within the required

(

margin of safety. The limit of a maximum 8 control rods inoperable is still maintained.

8) The reference to startup testing is deleted based on completion of the Startup Test Program. Completion-documented in reference 2. l 1
9) The leak detection system for the main steam area consists of ambient P temperature and delta temperature sensors to alarm and isolate the system on temperature values which sense steam leaks of 1% to 10% of rated steam flow (LaSalle County FSAR,Section 7.3.1.3.2.4.3.).. The Technical i Specification requires the delta and ambient temperature sensors to be operable in Modes 1, 2,' and 3.

I The Reactor Building ventilation fans are required to be shutdown

> periodically for filter changes during plant operation especially during L the spring planting season.(dusty conditions) and during secondary

containment isolation damper cycling. Upon restart of the ventilation fans, even with available auxiliary ventilation operating, there is a very high probability of a Group I isolation and Reactor Scram on high delta
  • temperature due to hot air swept out and cold air being brought in during fan restart. The plants' engineered safety features and ECCS systems have i been unnecessarily challenged during minor routine operations such as these.

.This change allows bypass 'of the MSIV delta-temperature leak detection trip channels for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The ambient temperature trip channels are.

still required to be operable which would provide continuous: leak idetection protection.. The short 4-hour bypass of delta temperature leak detection protection is more than outweighed by the reduction of l , challenges to .the' plants' engineered safeguards and ECCS systems as well as .the maintenance of a clean atmosphere in the Reactor Building.

10) During required testing to determine the adequacy to these setpoints for

' Unit 1 tit was found that these values were not conservative. They were

( administratively corrected to the'new setpoints and NRR informed to place L

'the revised setpoints-in the Unit 2 Tech Spec and that Unit 1 would be L corrected during this Tech Spec change submittal.

V o

I s

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

.11) The reference to startup testing is deleted based on completion of the

'Startup Test Program.

12) This change is based on the requirement to add time delay re'ays (Unit 1 license condition 2.c.(30).(b)) and is in accordance with reference (1).
13) This change is in accordance with reference (1) with the actual data

-submitted under reference (3).

14) This change is in accordance with reference (1).
15) Editorial change only.
16) This is a clarification requested by NRR based on previous approval of NPF-ll Ammendment 2.
17) This change based on previous submittal (reference (4)) with comments from NRR.
18) The Off Gas Post Treatment Radiation Monitoring System as described in FSAR Sections 11.5.2.1.2 and 11.3.2.1.1.9 is used to provide an automatic isolation function on a High High High radiation level. This function closes the discharge valve from the Off Gas System to the station vent stack. This valve is located downstream of the after filters and upstream of the station vent stack. (See Figure 11.3-1.)

Due to the location of this valve in the Off Gas System, it cannot be closed when the Off Gas System is in operation without affecting main condenser vacuum and Off Gas System operation. Since surveillences for similar type primary containment isolation functions are not required to test the actual isolation valve automatic closure except during scheduled plant shutdowns on an 18-month frequency, it is reasonable to test all functions of the post treatment radiation monitor except the actual valve closure as part of the Quarterly Channel Functional Test. The verification of automatic isolation function will then be verified at j least once per 18 months as part of the channel calibration.

l 19) This changed based on reference (1).

20) Typo change based on reference (1)
21) This change allows for reasonable instrumentation / testing accuracy and is licensee supplied number based on reference (1). This tolerence does not i effect the ability to ensure that the pressure isolation valves leakage rate is within required limits.

i l' ' 22) The revised withdrawl times were requested by NRR and are based on current regulations.

l i

l

\

23) These changes were made at NRR request (reference (5)).
24) The revised setpoint is more conservative than the original setpoint and corrects a typographical error.
25) The reference to startup testing is deleted based on completion of the Startup Test Program.
26) This change based on reference (1).
27) These changes are based on technically correct actions and are required due to allowance provided for in the surveillence requirements.
28) This change is an added requirement to increase the assurance that the Standby Gas Treatment System will perform its required function. This change was requested by NRR.
29) The SRV Test footnote is no longer required since the test is completed and results submitted to NRR (reference (6)).
30) a. The addition of "3.0.4 not applicable" to specific valves which are not required to be operable for other specifications or where the other specification contains "3.0.4 not applicable " statement allows startup under the action statement. With the valve closed the requirement for ,

primary containment integrity is maintained with the valve inoperable.

b. This is a requirement of mR (reference 7).
c. Valves ICM0238 and 1CM024A are not containment isolation valves and '

receive no isolation signals. The system outboard of these valves is required to be at containment pressure during isolation periods since other valves ICM0218, ICt022A, 1CM025A, and IC W26B open automatically upon reciept of an isolation signal (footnote (h) to Tech Spec table 3.6.3-1).

d. Added valve 1E12-F0998 - valve inadvertantly left off of original table.
31) This change is required by footnote in present Unit 1 specification.
32) This change required by mR (reference 8).

' 33) The revised area temperature table is in accordance with reference (1).

The reduced temperature requirements are based on equipment operability and not upon nominal HVAC design values and personnel comfort.

34) The revised snubber specification without the snubber table was a change requested by CECO and others. This Specification was provided by NRR.
35) This change is based on reference (1). A correction to the applicability which is in error in reference (1) was also included. The footnote was deleted based on Startup Test completion. (See reference 11).
36) a. The added note is simply a clarification of Action F. If all the loads which the diesel generator feeds are declared inoperable than the associated diesel (1A) need not be run every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. The required actions which pertain to the equipment declared inoperable still apply.

In addition any valid failure would still require adjustment of the testing frequency in accordance with Table 4. 8.1.1.2-1.

b. This changed in accordance with Generic Letter 83-30.
c. The diesel starts on stored air requirement is fullfilled during the preoparational test program at the minimum allowable air reciever pressure. Additional periodic testing is not necessary and adds several additional starts to the diesels. In discussions with NRR this old requirement was deleted.
37) This change is based on the need to have only one of the Division 1 or Division 2 Emergency diesel generators available during shutdown periods.

The words and/or were easily misinterpreted and this change is simply a clarification as to what is required.

38) This change was requested as part of reference (3).
39) This change is based on the need to have only 1 of the Division 1 or Division 2 Emergency diesel generators available during shutdown periods.

The words and/or were easily misinterpreted and this change is simply a clarification as to what is required.

40) This change was required by footnote *on page 3/4 8-15.
41) This change is based on the need to have only 1 of the Division 1 or Division 2 Emergency diesel generators available during shutdown periods.

The words and/or were easily misinterpreted and this change is simply a clarification as to what is required.

42) LSCS Technical Specification Basis for Electrical Equipment Protective Devices 3/4.8.3 states " Primary containment electrical penetrations and

, penetration conductors are protected by either de-energizing circuits not i required during reactor operation or demonstrating OPERABILITY of primary l and bacup overcurrent protection circuit breakers by periodic surveillance."

Since the primary containment electrical penetrations and penetration conductors are protected by either method discussed above it is reasonable to move the drywell crane and hoist circuits from Tech Spec Section

3.8.3.2 to 3.8.3.1 and protect these circuits by de-energizing the circuits when primary containment is required. These circuits are not required during reactor operation.

l

43) If a thermal overload bypass circuit is inoperable and administratiave control is taken as required by the action statement, startup (mode changes) should not be restricted since the system will still provide any required safety function. .
44) Valve IVQ041 was inadvertantly included on this Table. 'Although not required by the valve function, thermal overload bypass protection for this valve was maintained. This valve receives no automatic signals and valve position dues not effect operation of safety related systems.
45) The revised surveillance requirement prevents additional challanges to the plant safety system since to perform the surveillance requires deenergizing the RPS bus in question thereby allowing a single failure to cause a reactor scram. In addition shutdown of reactor aux 111 aries (eg Reactor "ater cleanup, Reactor Water Sampling etc.) is required to perform surveillance due to isolation valve closure required by the loss of an RPS bus.
46) This change based on previous submittal (reference 4) with comments from tBR.
47) This change allows relaxation of sampling requirements when the DOSE EQUIVALENT I-131 concentration in the primary coolant does not increase by more than a factor of 3 and the noble gas monitor shows that effluent activity has not increased more than a factor of 3. If these conditions are met there is assurance that the change in thermal power had no marked effect upon the fuel and therefore increase sampling is not warrented.
48) This change reflects actual installed sample locations as defined in the Offsite Dose Calculation Manual (ODCM).
49) This change was provided by NRR and is based on current NRR position on shutdown time allowances when an LCO is not met.
50) This change based on FSAR Table 6.3.2 for HPCS pump flow for reflooding requirements (reactor vessel at low pressure).
51) This is a minor clarification which adds requirements for an Special Report required by this Specification.
52) The revised snubber specification without the snubber table was a change requested by CECO and other. This Specification was provided by tER.
53) These changes are minor editorial changes only.
54) Administrative correction to the drywell net free. volume based on FSAR Table 6.2-1.
55) Administrative change only.
56) The revised shift manning Tables is for two Unit operation this table is the same as that issued for Unit 2 for two Unit operation. This table is in accordance with reference (9).
57) Editorial change only.

Based on the preceding discussion, it is concluded that the proposed change clearly falls within all acceptable criteria with respect to the system or component, the consequences of previously evaluated accidents will not be increased and the margin of safety will not be decreased. Therefore, based on tne guidance provided in the Federal Register and the criteria established in 10 CFR 50.92(e), the proposed change does not constitute a significant hazards consideration.

Retrieved from "https://kanterella.com/w/index.php?title=ML20087N309&oldid=3495596"