Text

Proposed Tech Spec Change PCOL-83/08,enhancing Clarity & Providing Consistency W/Plant Design & Operation,Per 821020 Confirmation of Action Ltr
	ML20076G568
Person / Time
Site:	Grand Gulf
Issue date:	06/09/1983
From:	; MISSISSIPPI POWER & LIGHT CO.
To:	;
Shared Package
ML20076G565	List: ML20076G561; ML20076G568;
References
	NUDOCS 8306140729
	Download: ML20076G568 (72)
	v • d • e

_

PROPOSED CHANGE TO THE OPERATING LICENSE NPF-13 PCOL-83/08 Mississippi Power & Light (MP&L) requests that the operating license for Grand Gulf Nuclear Station (GGNS) (NPF-13) be amended as detailed below.

These proposed changes, as discussed below, are provided for Nuclear Regulatory Commission (NRC) review and approval per 10CFR50.90.

1. (GGNS - 487 item 2)

SUBJECT:

Technical Specification Table 2.2.1-1, page 2-4.

DISCUSSION:

Technical Specification Table 2.2.1-1 specifies the Reactor Ver al Dome Pressure-High setpoint as less than or equal to 1093 psig and the allowable value as less than or equal to 1080 psig. The setpoint and allowable value should be changed to less than or equal to 1064.7 psig and less than or equal to 1079.7 psig, respectively.

JUSTIFICATION: General Electric Company (NSSS supplier) design specifications state the setpoint value as less than or equal to 1064.7 psig with an allevable value of less than or equal to 1079.7 psig.

General Electric has reviewed the proposed changes and concurs with the new values.

SIGNIFICANT HAZARDS CONSIDERATION:

The proposed change is administrative in that it corrects an error in the previously identified values for these parameters.

This corresponds to NRC example (i) of changes which are not significant hazards considerations. No significant increase in the probability of consequences of an accident previously evaluated is created nor is the possibility of a new or different kind of accident from any accident previously evaluated introduced. No reduction in safety margin is created since the setpoint is proposed to be changed in a conservative direction and the margin between the setpoint and allowable valve remains constant. Therefore, this change constitutes no significant hazards considerations.

8306140729 830609 PDR ADOCK 05000416 p

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j TABLE 2.2.1-1 i

i REACTCR PROTECTION SYSTEM INSTRUMENTATION SE1 POINTS

'M.L0lRIOLE h

VALUES TRIP SETPOINT F'TIONAL WIIT E

1.

Intemediate Range Monitor, Neutron Flux-High i 120/125 divisions i 122/125 divlsfens e

of full scale of full scale N

N ege Power Range Monitor:

< 20K of RATED 2.

l 7

< 15% of RATED a.

Neutron FIva-Nigh, Setdown

~ THERMAL POWER

~ THERMAL POWER w

b.

Flow Blased $1mulated Thermal Power-High

< 0,66 W+48E, with

< 0.66 W+51%, with a maximum of

1) Flow Biased a maximue of

~

< 111.0% of RATED

< 113.0E of RATED

~ THERMAL POWER

2) Migh Flow Clamped

~ THERMAL POWER

< 120E of RATED

< 118% of RATED

~ THERMAL POWER c.

Neutron Flex-High

~ THERMAL POWER NA NA

< g/019.7 I

T d.

Insperative

/457 psig 3.

Reetter Vessel Steam Dome Pressure - High i1 psig

> 11.4 inches above

> 10.8 inches above 4.

Reacter Vessel Water Level - Low, Level 3

~ instrument zero"

~ instrument zero*

< 53.5 inches above

< 54.1 inches above 1

5.

Ileetter Vessel Water Level-Migh. Level 8

~ instrument zero*

j 5.

Mein Steam Line Isoletten Valve - Closure 1 6% closed i 75 closed 1

l 7.

Main Steam Line Radiatten - High 5 3.0 x full power i 3.6 n full power background backgrowns',

O j

1 1.73 psig i 1.93 psig h

l 8.

Drywell Pressere - Nigh

< 6 5 of f911 scale Scram Discherge Volume Water Level - High 1 6dt of full scale h

9.

> 40 psIg

> 0 psis l

10. Tertine Step Valve - Closure

11. Tertine Centeel Valve Fast Closure,

_ 41 psig

_ 44.3 psig g

j Trip 011 Pressure - Low NA 4

NA i

12. Reacter Mode Swltch Shutdown Position h

MA MA 2

13. Manual Scram

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i "5ee Bases Figure 8 3/4 3-1.

t i

I I

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2. (GGNS - 569, 322)

SUBJECT:

Technical Specifications as follows:

Specification Page Specification Page Table 3.3.2-1 (Note *)

3/4 3-14 3.6.6.3 (Action a.2 & b) 3/4 6-53 3.3.2 (Action 21.b) 3/4 3-14 3.7.1.1 (Note *)

3/4 7-1 Table 4.3.2.1-1 (Note *)

3/4 3-23 3.7.1.3 (Note *)

3/4 7-4 Table 3.3.7.1-1 (Note **)

3/4 3-57 3.7.2 (Note *)

3/4 7-5 t

Table 4.3.7.1-1 (Note **)

3/4 3-59 3.7.2 (Action b.2) 3/4 7-5 i

3.6.6.1 (Note *)

3/4 6-46 3.8.1.2 (Note *)

3/4 8-9 3.6.6.1 (Action b) 3/4 6-46 3.8.1.2 (Action a) 3/4 8-9 3.6.6.2 (Note *)

3/4 6-47 3.8.2.2 (Note *)

3/4 8-14 3.6.6.2 (Action) 3/4 6-47 3.8.2.2 (Action a) 3/4 8-14 3.6.6.3 (Note *)

3/4 6-53 3.8.3.2 (Note *)

3/4 8-17 3.8.3.2 (Action a.1 & b.1) 3/4 8-18 i

DISCUSSION:

The Technical Specifications listed above contain statements defining operational conditions and/or actions involving the handling of irradiated fuel assemblies in the primary or secondary containment. Various notes define operational conditions where operability of systems and ccmponents providing a protective function during a fuel handling accident is required. Action statements specify restrictions to be applied due to equipment inoperability.

4 These notes and statements are inconsistent throughout the Technical Specifications. The Technical Specifications are confusing due to terminology differences and it is not clear i

when the specification applies. The notes and statements should be revised to eliminate confusion and to accurately define the requirements. Specific changes that should be made are as follows:

1.

Technical Specification Table 3.3.2-1, note designated

"*", page 3/4 3-14.

i Change note to read "When handling irradiated fuel in the primary or secondary containment and during CORE ALTERATIONS and operations with a potential for draining the reactor vessel."

2.

Technical Specification 3.3.2, Action Statement 21.b, page 3/4 3-14 i

Insert the word " primary" so that the statement reads

"... irradiated fuel in the primary containment and..."

l 3.

Technical Specification Table 4.3.2.1-1, Note designated

"*", page 3/4 3-23.

Insert the words " primary or" so the statement reads

"... irradiated fuel in the primary or secondary containment..."

F3(3651)

4 Technical Specification Table 3.3.7.1-1, Note designated

"**", page 3/4 3-57.

Insert the words "prinary or" so the statement-reads

"...being handled in the primary or secondary containment."

5.

Technical Specification Table 4.3.7.1-1, Note designated

"**" page 3/4 3-59.

Insert the words " primary or" so the statement reads

"...being handled in the primary or secondary containment."

6.

Technical Specification 3.6.6.1, Note designated "*" page 3/4 6-46.

Insert the words " primary or" so the statement reads

"... handled in the primary or secondary containment and..."

7.

Technical Specification 3.6.6.1, Action Statement b, page 3/4 6-46.

Insert the words " primary or" so the statement reads

"... irradiated fuel in the primary or secondary containment..."

8.

Technical Specification 3.6.6.2, Note designated "*" page 3/4 6-47.

Insert the words " primary or" so the statement reads

"... handled in the primary or secondary containment and..."

9.

Technical Specification 3.6.6.2, Action Statement, page 3/4 6-47.

Insert the words "pr.imary or" so the statement reads

"... irradiated fuel in the primary or secondary containment..."

10.

Technical Specification 3.6.6.3, Note designated "*" page 3/4 6-53.

Insert the words " primary or" so the statement reads

"... being handled in the primary or secondary containment..."

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11.

Technical Specification 3.6.6.3, Action Statement a.2 and b, page 3/4 6-53.

Insert the words " primary or" so the statement reads

"... irradiated fuel in the primary or secondary containment..."

12.

Technical Specification 3.7.1.1, Note designated "*" page 3/4 7-1.

Replace the words " Auxiliary Building or Enclosure Building" with the words " primary or secondary containment."

13.

Technical Specification 3.7.1.3, Note designated "*" page 3/4 7-4.

Replace the words " Auxiliary Building or Enclosure Building" with the words " primary or secondary containment".

14 Technical Specification 3.7.2, Note designated "*" page 3/4 7-5.

Insert the words " primary or" so the statement reads

"...being handled in the primary or secondary containment".

15. Technical Specification 3.7.2, Action Statement b.2 page 3/4 7-5.

Insert the words " primary or" so the statement reads

"... irradiated fuel in the primary or secondary containment..."

16. Technical Specifications:

a.

3.8.1.2, note designated "*", page 3/4 8-9.

b.

3.8.1.2, Action Statement a, page 3/4 8-9.

c.

3.8.2.2, note designated "*", page 3/4 8-14.

d.

3.8.2.2, Action Statement a, page 3/4 8-14.

3.8.3.2, note designated "*", page 3/4 8-17.

e.

f.

3.8.3.2, Action Statements a.1 and b.1, page 3/4 8-18.

Replace the words " Auxiliary Building or Enclosure Building" with the words " primary ~pr secondary containment".

JUSTIFICATION: FSAR subsections 15.7.4 and 15.7.6 present an analysis of the Design Basis Fuel Handling Accident. Those systems and their associated components which provide a protective function during this event are the secondary containment, containment-i and drywell ventilation isolation, fuel handling ~ area ventilation isolation, standby-gas treatment, and the control room ventilation system.- The analysis assumes an irradiated

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fuel assembly is dropped from the fuel handling equipment onto irradiated fuel in the reactor vessel or the spent fuel pool; i

- however, the analysis is also applicable to dropping a fuel L

assembly onto fuel stored in the primary containment storage i

racks. Primary containment integrity is.not required during fuel handling operations since the secondary containment 4

performs the protective function; however, the containment and i

drywell ventilation isolation' system and components do provide a protective function since~the ventilation system represents an additional release path to the environment.

4 l

The notes and statements discussed above define operability requirements and conditions for those systems and components 4

that provide a protective function during a fuel handling accident; however, the terminology is inconsistent.

t

. Some statements refer to containment and it is not clear which containment is referenced, (primary or secondary or both).

Some statements refer to secondary containment and core alterations; however, " core alterations" does not include stored fuel in the primary centainment. Other statements refer to handling operations in the Auxiliary Building and Enclosure Building. These areas are both considered secondary containment; however, associated action statements refer to handling operations in the primary containment. Thece inconsistencies should-be corrected to eliminate confusion and j -

properly define the requirements. Justification for specific changes is provided below:

l i

1.

Technical Specification Table 3.3.2-1, note designated

"*", page 3/4 3-14.

This table delineates the operability requirements of isolation actuation instrumentation and includes both t

primary and secondary isolation instrumentation for 4

ventilation valves; however, the note requiring operability during-fuel handling operations does not specifically state that operability is required when i

handling irradiated fuel in secondary containment. FSAR p

subsection 15.7.4 provides an analysis of a Fuel Handling Accident in secondary. containment. This analysis assumes j.

an automatic isolation of the ventilation system on detection of a high radiation. signal from' the exhaust fuel storage pool. sweep instrumentation.

FSAR. subsection.15.7.6 provides an analysis of a fuel handling accident in primary. containment. This analysis assumes automatic isolation of the containment and drywell ventilation system as well as isolation of the secondary containment ventilation system. Isolation is initiated by i

the' containment and drywell ventilation exhaust radiation-i_

high instrumentation and the fuel handling area l

- ventilation exhaust radiation-high high instrumentation,

^

respectively.

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g,-tw--

-**tt+*e--r--*-wr"*-d"-a*tr--'

The note in this table refers to both primary and secondary isolation instrumentation and should be revised to identify fuel handling operations in either the primary or secondary containment.

2.

Technical Specification 3.3.2, Action 21.b, page 3/4 3-14 Action Statement 21 applies to only the containment and drywell ventilation exhaust radiation high instrumentation which isolates primary containment ventilation only. For clarity, the statement should be revised to specify primary containment.

3.

Technical Specification Table 4.3.2.1-1, note designated

"*", page 3/4 3-23.

This table delineates the surveillance requirements for isolation actuation instrumentation. This is the same instrumentation identified in Table 3.3.2-1.

As discussed above operability is required during fuel handling operations in primary or secondary containment; therefore, the note should state the applicable conditions the same as in table 3.3.2-1.

4.

Technical Specification Table 3.3.7.1-1, note designated

"**", page 3-57.

This table delineates the operability requirements of radiation monitoring instrumentation which actuates protective functions. The accident analysis in FSAR subsection 15.7.6 assumes isolation of the secondary containment ventilation system and SGTS initiation in response to a fuel handling accident in the primary containment. Therefore, the note should include handling operations in the primary containment.

5.

Technical Specification Table 4.3.7.1-1, note designated

"**", page 3-59.

This table delineates surveillance requirements for the instrumentation in Table 3.3.7.1-1; therefore, the note should be changed to require surveillance when operability is required as discussed in item 4.

6.

Technical Specification 3.6.6.1, note designated "*", page 3/4 6-46.

FSAR subsections 15.7.4 and 15.7.6 provide an analysis of a fuel handling accident. The analysis includes accidents within primary containment as well as accidents within secondary containment. Primary containment integrity during fuel handling operations is not required (except ventilation)'since secondary containment restricts release. The applicability note in the specification should include fuel handling operations in the primary containment.

F7(3651) l

7.

Technical Specification 3.6.6.1, Action Statement b, page 6-46.

Secondary containment integrity is also required during irradiated fuel handling within the primary containment as discussed in item 6 above. The specification should be revised to include this requirement.

8.

Technical Specification 3.6.6.2, Note designated "*", page 3/4 6-47.

As discussed in item 6 above, isolation capabilities are required in mitigating a fuel handling accident in primary containment. The note should be revised to include this requirement.

9.

Technical Specification 3.6.6.2, Action statement, page 6-47.

The specification should be revised to include irradiated fuel handling operations as discussed in item 8 above.

10.

Technical Specification 3.6.6.3, Note designated "*", page 3/4 6-53.

FSAR subsections 15.7.4 and 15.7.6 provide an analysis of a fuel handling accident within primary or secondary containment.

In both cases operation of the Standby Cas Treatment System is assumed. The note should be revised to require operability while handling irradiated fuel in the primary containment as well as in the secondary containment.

11.

Technical Specification 3.6.6.3, Action Statement a.2 and b, page 3/4 6-53.

As discussed in item 10 above, operability is required during the handling of irradiated fuel in the primary containment; therefore, action requirements should be revised to require suspension of such activities when the SGTS becomes inoperable.

12.

Technical Specification 3.7.1.1, note designated "*", page 3/4 7-1.

This note refers to handling irradiated fuel within the auxiliary building or enclosure building. This terminology is inconsistent with other technical specifications which refer to handling irradiated fuel in the primary or secondary containment. The intent of this specification is to assure electrical power to protective systems in the event of a fuel handling accident. As discussed previously, the fuel handling accident analysis presented in FSAR subsections 15.7.4 and 15.7.6 essumes operability of ventilation isolation systems, radiation F8(3651)

monitoring systems, and the Standby Gas Treatment System; therefore, the note should refer to primary or secondary containment rather than the building designations.

13.

Technical Specification 3.7.1.3, Note designated "*", page 3/4 7-4.

As discussed in item 12, the note should refer to primary or secondary containment rather than the building designations.

14.

Technical Specification 3.7.2, Note designated "*", page 3/4 7-5.

This specification delineates the operability requirements of the control room emergency filtration system. FSAR subsection 6.4 states the control room ventilation system limits exposure to operating personnel to within the guidelines of 10CFR50, Appendix A and general design criterion 19 during any of the design basis accidents.

This system is required to be operable during the fuel handling DBA; however, the note refers only to ha'ndling fuel in the secondary containment. For a fuel handling accident inside the primary containment, the containment and drywell ventilation system is another release path that must be considered; therefore, the note should also specify irradiated fuel handling operations in the primary containment.

15.

Technical Specification 3.7.2, Action statement b.2, page 3/4 7-5.

As discussed in item 14 above, the handling of irradiated fuel in the primary containment also presents an exposure hazard; therefo'e, the action should require suspension of r

these activities.

16.

Technical Specifications as follows:

a.

3.8.1.2, note designated "*", page 3/4 8-9.

b.

3.8.1.2, Action statement a, page 3/4 8-9.

c.

3.8.2.2, note designated "*", page 3/4 8-14.

d.

3.8.2.2, Action statement a, page 3/4 8-14.

e.

3.8.3.2, note designated "*", page 3/4 8-17.

f.

3.8.3.2, Action statements a.1 and b.1, page 3/4 8-18.

In all of these Technical Specifications, the terminology used is inconsistent with other Technical Specifications.

There specifications assure electrical power to systems and Jnstrumentation that is required to function during a fuel handling accident within primary or secondary l

l F9(3651)

containment as discussed in FSAR subsections 15.7.4 and e

15.7.6.

The notes and action statements listed above should be revised to refer to primary and secondary containment rather than the Enclosure and Auxiliary Building.

SIGNIFICANT HAZARDS CONSIDERATIONS:

Evaluation of no significant hazards considerations for these proposed changes are provided below:

1.

Technical Specification Table 3.3.2-1, note designated

"*", page 3/4 3-14.

This proposed change is purely administrative in nature in that it clarifies the term "in containment" to mean both primary and secondary containment. This is the broadest interpretation of this term possible and is thus most limiting. Based upon NRC examples (i) and (ii) of Standards for Determining Whether License Amendments Involve No Significant Hazards Considerations, this change constitutes no significant hazard.

2.

Technical Specification 3.3.2, Action 21-b, page 3/4 3-14 Action Statement 21 applies strictly to a primary containment function associated with isolation resulting from a Containment and Drywell Ventilation Exhaust Radiation - liigh Signal. The proposed change clarifies this scope. As such it is purely administrative as described in NRC example (i) of amendments not likely to involve significant hazards considerations, this change does not constitute a significant hazard consideration.

3.

Technical Specification Table 4.3.2.1-1, note designated

"*", page 3/4 3-23.

This proposed change clarifies the surveillance requirement to be consistent with the operability change proposed in item 1.

It is purely administrative and based upon the justification provided by item 1 constitutes no significant hazards considerations.

4.

Technical Specification Table 3.3.7.1-1, note designated

"**", page 3-57.

This proposed change expands the operability requirements for the Control Room Radiation Monitor and the Fuel Handling Area Ventilation Exhaust Radiation Monitor to include fuel handling operations in the primary containment. This change, therefore, constitutes an additional restriction not presently included in the Technical Specifications. As such, it corresponds to NRC F10(3651)

example (ii) of amendments not likely to involve significant hazards considerations. Therefore, this change does not constitute a significant hazards consideration.

5.

Technical Specification Table 4.1.7.1-1, note designated

"**" page 3/4 3-59.

This proposed change expands OPERATIONAL CONDITIONS for which surveillance is required and clarifics the surveillance requirements to be coneistent with the operability change proposed in item 4.

It corresponds to NRC example (ii) mentioned in item 4 above and based upon the justification provided in item 4 constitutes no significant hazards considerations.

6.

Technical Specification 3.6.6.1, note designated "*", page 3/4 6-46.

This change expands the OPERATIONAL CONDITIONS for which this LCO is applicable to include the handling of fuel in the primary containment.

It corresponds to NRC example (ii) of amendments not likely to involve significant hazards considerations in that it constitutes an additional restriction not presently included in the Technical Specifications therefore, this change does not constitute a significant hazards consideration.

7.

Technical Specification 3.6.6.1, Action Statement b, page 3/4 6-46.

This proposed change adds an additional requirement to suspend handling of irradiated fuel in the primary containment in the event of loss of secondary containment integrity during Operational Condition *. This change is cesigned to address the fact that primary containment integrity, with the exception of ventilation, is not required during fuel handling operations. Integrity is provided by the secondary containment. As such, this change constitutes an additional restriction not presently included in the Technical Specifications.

It corresponds to NRC example (ii) of amendments not likely to involve significant hazards considerations. Therefore, this changes does not constitute a significant hazards cotisideration.

8.

Technical Specification 3.6.6.2, Note' designated "*", page 3/4 6-47, s

Thischangeexpandstheoperationalconditionsherwh'ich this LCO is applicable to include the handling of fuel in the primary containment.

It corresponds to NRC cxample (ii) of amendments not lik_ely to involve significEnt

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hazards considerations in that it constitutes an additional restriction not presently included in the Technical Specifications. Therefore, this change does not constitute a significant hazards consideration.

9-Technical Specification 3.6.6.2, Action Statement, page 6-47.

This change expands restricted operation to include handling of irradiated fuel in the primary containment.

For reasons identical to item 8 above, this change does not constitute a significant hazards consideration.

10.

Technical Specification 3.6.6.3, Note designated "*", page 3/4 6-53.

This change expands the OPERATIONAL CONDITIONS for which this LCO is applicable to include the handling of fuel in the primary containment.

It corresponds to NRC example (ii) of amendments not likely to involve significant hazards considerations in that it constitutes an additional restriction not presently included in the Technical Specifications. Therefore, this change does not constitute c significant hazards consideration.

11.

Technical Specification 3.6.6.3, Action Statement a.2 and b, page 3/4 6-53.

These changes expand operational restrictions to include handling of irradiated fuel in the primary containment.

For reasons identical to 10 above, this changes does not constitute a significant hazards consideration.

12.

Technical Specification 3.7.1.1, note designated "*", page 3/4 7-1.

This proposed change established consistency in nomenclature used to identify areas where the FSAR accident analyses have shown that safety is dependent upon appropriate power available to required safeguards features. This is a purely administrative change corresponding to NRC example (i) of amendments not likely to involve significant hazards considerations. Therefore, this change does not constitute a significant hazards consideration.

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13.

Technical Specification 3.7.1.3, Note designated "*", page 3/4 7-4 1

This proposed change establishes clarification for the purpose of nomenclature consistency identical to item 12

~

above. For the reasons discussed in item 12, this change I

1s does not constitute a significant hazards consideration.

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14. Technical Specification 3.7.2 Note designated "*", page 3/4 7-5.

The proposed change expands the operability requirements for the Control Room Emergency Filtration System to include fuel handling operations in the primary containment. This change therefore constitutes an additional restriction not presently included in the Technical Specifications. As such, it corresponds to NRC example (ii) of amendments not likely to involve significant hazards considerations. Therefore, this change does not constitute a significant hazards consideration.

15. Technical Specification 3.7.2, Action Statement b.2, page 3/4 7-5.

This change expands operational restrictions to include handling of irradiated fuel in the primary containment.

For reasons identical to 14 above, this change does not

]

constitute a significant hazards consideration.

l

16. Technical Specification as follows:

a. 3.8.1.2, Note designated "*", page 3/4 8-9.

b. 3.8.1.2, Action Statement a, page 3/4 8-9.

c. 3.8.2.2, note designated "*", page 3/4 8-14.

d. 3.8.2.2, Action Statement a, page 3/4 8-14

e. 3.8.3.2, note designated "*", page 3/4 8-17.

l

f. 3.8.3.2, Action Statement a.1 and b.1, page 3/4 8-18.

All of these proposed changes involve clarifications for the purpose of nomenclature consistency identical to items i

12 and 13 above.

For the reasons discussed in these j

items, these changes do not constitute a significant hazards consideration.

NOTE:

Technical Specification page changes marked with a PCOL number j

and circled are changes that were previously submitted to the

?

NRC.

4 1

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(GGNS-56b322)

INSTRWENTATION Ta1LE 3.3.2-1 (Continued)

ISOLATT T ACTUAT10N INSTRUMENTATION i

ACTION Be in at least NOT SHUTDOW within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and ACTION 20 within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

l l

Close the affected system isolation valve (s) within one hour ACTION 21 In OPERATIONAL CON 0! TION 1, 2, or 3 be in at least HOT or:

SHUTDOW within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD a.

pr ea.cy within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

l In Operational Condition *, suspend CORE ALTERATIONS,.5M 5 3

handling of irradiated fuel in the containment and l v e m 2.

b.

operations with a potential for draining the reactor vessel.

Restore the manual initiation function to OPERA 8L 4

48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least NOT SHUTDOW within the ne M

ACTION 22 j;

and in COLD 5HUTDOW within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

c,-. j 8e in at least STARTUP with the associated isolati within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or be in at least NOT SHUTDOW within 12 ACTION 23 y (2 and in COLD SHUTDOW within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

_t 8e in at least STARTUP within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ,

[(ACTION 24 Establish SECONDARY CONTAINMENT INTEGRITY with the standb w s. f 2jj,7ACTIDN25 treatment system operating within one hour.

Restore the manual initiation function to OPERA 8LE statu within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or close the affected system iso

~} fMg ACTION 26 4fg yey Close the affected system isolation valves within one hour d.I W~g ACTION 27 and declare the affected system inoperable.

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Lock the affected system isolation valves closed within one hour f ev< ~

7 eE S ACTION 28 and declare the af facted system inoperable.

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=i When handling irradiated fuel in the3contai6 ment and during CORE n -g ALTERATIONS and operations with a potential for draining the E1Q*

During CORE ALTERATIONS and operations with a potential fce

[.h. g; See Specification 3.6.4, Table 3.6.4-1 for valves in each valve group.

reactor vessel, Sir

,e A channel may be placed in an inoperable status for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months fo es, p ((a)required surveillance without placing the trip system in the tripped b)

~~E,f f(a dition provided at least one other OPERABLE channel in the sam 3

h,q. *J 1s monitoring that parameter.

Also actuates the standby gas treatment sy oe 7 J.f e (c)

N tM t '; :y:t NNOM ht*U (d) mode of operation.

me:S :nd/ r te d:c:::h :-trt:

(e),* :

(fT Alsol tips and isolates the mechanical vacuum pumps.

A channel is OPERA 8LE ff 2 of 4 instruments in l

Also actuates secondary containment ventilation isolation dampe (g) g$

(h) ti:g:. 'valv633-F004 Sf 5-FDC8:4G33#ffl valves per Table 3.6.6.2-1.

i' g

l h es. A n kar/

?

AiA.emb N. 56Estem h :t ::

Closes only RWCU s

1. v G s Tre beat f'y >h a e '.s."D1F"' "#' "'1%

N.

(1) g)

sRAnoTLAMi

TABLE 4.3.2.1-1 (Continted) es ISOLATION ACTUATION INSTRUMENTATION SURVElttANCE REQUIREMENTS OPERATIONAL 8

CHANNEL G;

CHANNEL FUNCil0NAL CHANNEL CONDITIONS IN WNICH CHECK TEST CALIBRA110N SURVElttANCE REQUIRED TRIP FUNCTION i

5.

RNR SYSTEM ISOLATION RHR Equipment Room Ambient Temperature - High 5

M R

1,2,3 a.

b.

RNR Egulpment Room l

a Temp. - High 5

M R

1,2,3 Reactor Vessel Water Level -

5 M

R 1, 2, 3 c.

tow, Level 3 i

f d.

Reactor Vessel (RNR Cut-in Permissive) Pressure - High 5

M R

1,2,3 w) i 5

M R

1,2,3 j

Drywell Pressere - High Y

e.

M '}

MA 1,2,3 I

O f.

Mensal Initiation MA i

I n.;

y *r "When handling terediated fuel in thegseccadary containment and during CORE ALTERATIONS and l

with a potential for draining the reactor vessel.

p

- When reactor steam pressure 3,1945 psig and/or any turbine stop valve is open.

During CORE ALTERATION and operations with a potential for draining the reactor vessel.

i All other G

i initiation switches shall be tested at least once per 18 months during shutdown.

Q l

circuitry associated with manual initiation shall receive a CHANNEL FUNCTIONAL TEST at least once (a) Manus) per 31 days as part of circuitry required to be tested for automatic system isolation.

7 l

p (b) Each train or logic channel shall be tested at least every other 31 days.

e 1

Os m

M V

l 1

I

h TABLE 3.3.7.1_1 (Continued)

RA0IATION MONITORING INSTRUMENTATION h@

MINIMUN CMANNELS APPLICABLE ALARM / TRIP MEASUREMENT RANGE

- ACTION OPERA 8LE CON 0!TIONS SETPOINT e,

INSTRWWITATION

!E E

10. Area Monitors Fuel Mand 11ag Area

3 a.

Monitors

-2 3

10 to 10 mR/hr 72 1)

New Fuel 1

(e) 12.5 mR/hr/NA Storage Yavlt

-2 3

<2.5 mR/hr/NA 10 to 10 mR/hr 72 2)

Spent. Feel 1

(f)

.6 M. to ~@

E

$4.,. e Are 1

(S) 2.5 mg,/gA W

10 to 10 mR/hr 72 At all times 10.5 mR/hr/NA k

l 3) cye b.

Control Room 1

1 Radiation Monitor gm l

'a.

p,:

,,f of-With RH4 heat exchangers in aperation, When irradiated fuel is being handled in thegecondary containment.

Any required change to l

Y Final Setpoint to be determined during startup test program.

5 i

this setpoint shall be submitted to Commission within 90 days after test complet on.

Initial setpuint.

l 8-

-='i

, or f

l (a) Trips system with 2 channels upscale-Mp M7 or one channel upscale and one ch I

dwasede bn. u;4;

- p:u:Q:

8 2 channels '-_

d '-.d ow xde.,

j (3) Isolates containment /drywell purge penetrations.

(c) With irrodisted fuel in spent fuel storage pool.(d) Also iselstes the second i

l (e) With feel in the new fuel storage vault.

(f) With feel in the spent fuel storage pool.

l (3) *.A f \\.. h me,,, w rge nee.

L ocl.sose.J o:p s y

j l

h LovP"*dt W.W.y nde opscJe WW =ed noe dwescJe, e,-

' " ' ## " d ' * "

E

"~

b

+N s.

2 vIiv)s, 2, (GGAls-52 B) s l

CGGNS-56Tp322)

=

TABLE 4.3.7.1-1 RADIATION MONITORING INSTRUNENTATION SURVEILLANCE REQUIREMENTS O

OPERATIONAL E

CHANNEL CONDITIONS FOR CHANNEL FUNCil0NAL CHANNEL WHICH SURVEILLANCE E

INSTRUMENTATION CHECK _

TEST CAllBRATION NEQUIRED e

g U

1.

Component Cooling Water Radiation 5

M R

At all times Moniter g

2.

Standby Service Water System Radiation Monitor 5

M R

1, 2, 3, and*

3.

Offgas Pre-treatment Radiation Monitor 5

M R

1, 2 4

Offgas Post-treatment Radiation Monitor 5

M R

1, 2 Q

5.

Carbon Bed Vault Radiation Monitor 5

M R

1, 2 m

7.

M 'I R

1, 2, 3, 5 and**

Y 6.

Control Room Ventilation Radiation I

S (a

Monitor 7.

Coatainment and Drywell Ventilation Exhaust Radiation Monitor S

M R

At all times k

y V

~

8.

Fuel Mandling Area Ventilation 5

M R

1, 2, 3, 5 and**

G4 Fu J.

Radiation Monitor N

9.

Fuel Mandling Area Pool Sweep 5

M R

(b) v Enhavst Radiation Monitor

10. Area Monitors a.

Fuel Handling Area Monitors 1)

New Fuel Storage Vault 5

M R

(c)

~2)

Spent Fuel Storage Pool 5

M R

(d) g.

5 M

R At all times

- p 3)

v. we h3c. Area.

n g

(e) k

( b. ' Control Room Radiation Monitor with uns neat exchangers in operation.

e, ;wr **

'A p

7 t

When treadiated fuel is beIng handled in the secondary containment.

(a) The CMRMMEL FUNCTIONAL TEST shall demonstrate that control room annunciation occurs if any a

i conditions entst.

Instrument indicates measured levels above the alare/ trip setpoint.

i 1.

i 2.

Circuit failure.

3.

Instrument indicates a downscale failure.

4.

Instrument controls not in Operate mode.

(b) With irradiated feel in the spent fuel storage pool.

i l

(c) With fuel in the new fuel storage vault.

We (d) With fuel in the spent fuel storage pool.

f (e) %fi4k{wel IN4e drfer 5torage arem.

4}

5-

z. @ms-m m)

CONTAINMENT SYSTEMS I

3/4.6.6 SECONDARY CONTAINMENT SECONDARY CONTAINMENT INTEGRITY L941 TING CONDITION FOR OPERATION SECONDARY CONTAINMENT INTEGRITY shall be maintained.

3.6.6.1 APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3 and *.

l ACTION:

Without SECONDARY CONTAINMENT INTEGRITY:

In OPERATIONAL CONDITION 1, 2 or 3, restore SECONDARY CONTAINMENT INTEGRITY within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least NOT SHUTDOWN within the a.

next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOW within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , ry n p,~ional Condition

, suspend handling of irradiated fuel in perat b.

In secondary containment, CORE ALTERATIONS and operations with a th The provisions of potential for draining the reactor vessel.

Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS SECONDARY CONTAINMENT INTEGRITY shall be demonstrated by:

4.6.6.1 Verifying at least once per 31 days that:

a.

All Auxiliary Building and Enclosure Butiding equipment 1.

hatches and blowout panels are closed and sealed.

The door in each access to the Auxiliary Sutiding and Enclosure 2.

Building is closed, except for routine entry and exit.

All Auxiliary Building and Enclosure Building penetrations not 3.

capable of being closed by OPERABLE secondary containment automatic isolation dampers / valves and required to be closed during accident conditions are closed by valves, blind flanges, or deactivated automatic dampers / valves secured in position.

b.

At least once per 18 months:

Verifying that one standby gas treatment subsystem will draw down 1.

the secondary containment to greater than or equal to 0.25 inches of vacuum water gauge in less than or equal to 120 seconds, and Operating one standby gas treatment subsystem for one hour and 2.

maintaining greater than or equal to 0.25 inches of vacuum u ter 9auge in the secondary containment at a. flow rate not exceeding 4000 CFM.

a Containment and during "When irradiated fuel is being handled in the CORE ALTERATIONS and operations with a potential for draining the reactor vess GRAND GULF-UNIT 1 3/4 6-46

g, (GG Al5-Src3,322)

CONTA1984ENT SYSTEMS SECONDARY CONTAlletENT AUTOMATIC ISOLATION DAMPERS / VALVES e

LIMITING CONDITION FOR OPERATION The secondary containment ventilation system automatic isolation 3.

6.2 dampers / valves shown in Table 3.6.6.2-1 shall be OPERABLE with isolation times less than or equal to the times shown in Table 3.6.6.2-1.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3 and *.

ACTION:

With one or more of the secondary containment ventilation system automstic isolation dampers / valves shown in Table 3.6.6.2-1 inoperable, maintain at least one isolation damper / valve OPERABLE in each affected penetration that is open, and within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months either:

J Restore the inoperable damper / valve (s) to OPERABLE status, or a.

Isolate each affected penetration by use of at least one deactivated b.

automatic damper / valve secured in the isolation position, or Isolate each affected penetration by use of at least one closed c.

manual valve cr blind flange.

Otherwise, in OPERATIONAL CONDITION 1, 2 or 3, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTOOWN within the l following24houg,ionalCondition

, suspend handling of irradiated Otherwise, i[condary containment, CORE ALTERAT]DNS and perat fuel in the>se The provisions of potential for draining the reactor vessel.

Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS Each secondary containment ventilation system automatic isolation 4.6.6.2 f? per/ valve shown in Table 3.6.6.2-1 shall be demonstrated OPERABLE:

Prior to returning the damper / valve to service after maintenance, repair or replacement work is performed on the damper / valve or its associated a.

actuator, control or power circuit by cycling the damper / valve through at least one complete cycle of full travel and verifying the specified isolation time.

During COLD SHUTDOWN or REFUELING at least once per 18 months by ve that on a containment isolation test signal each isolation damper / valve b.

actuates to its isolation position.

By verifying the isolation time to be within its limit when tested c.

pursuant to Specification 4.0.5.

l n%er nw

'When irradiated fuel ks being handled in the[secon[a GRAND GULF-UNIT 1 3/4 6-47

2. (cas-3ea m)

CONTAlemENT SYSTEMS

\\

STAND 8Y GAS TREATMENT SYSTEM L!NITING CONDITION FOR OPERATION 3.5.6.3 Two independent standby gas treatment subsystems shall be OPERABLE.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3 and *.

ACTION:

With one standby gas treatment subsystem inoperable, restore the a.

inoperable subsystem to OPERABLE status within 7 days, or:

1.

In OPERATIONAL CONDITION 1, 2 or 3, be in at least NOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

pr,% ev y *v InOperationadondition

, suspend handling of irradiated fuel 2.

in thersecondary containment, CORE ALTERATIONS and operations with a potential for draining the reactor vessel. The provi-sions of Specification 3.0.3 are not applicable.

g b.

With both standby gas treatment subrystems inoperable in Operational Condition *, suspend handling of irradiated fuel in the secondary containment. CORE ALTERATIONS or operations with a potential for draining the reactor vessel. The provisions of Specification 3.0.3.

are not applicable.

SURVEILLANCE REQUIREMENTS 1

4.6.6.3 Each standby gas treatment subsystem shall be demonstrated OPERABLE:

At least once per 31 days by initiating, from the control room, flow a.

through the HEPA filters and charcoal adsorbers and verifying that the subsystem operates for at least 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months with the heaters OPERABLE.

l pri & T **

"When irradiated fuel is being handled in the4 secondary containment and during l

CORE ALTERATIONS and operations with a potential for draining the reactor vessel.

~ GRAND GULF-UNIT 1 3/4 6-53

2. (cans-seg,m) 3/4.7 PLANT SYSTEMS 3/4.7.1 SERVICE WATER SYSTEMS STANDBY SERVICE WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.Y.1.I' Two independent standby service water (SSW) system subsystems shall be OPERABLE with each subsystem comprised of:

One OPERABLE SSW pump, and An OPERABLE flow path capable of taking suction from the associatec a.

SSW cooling tower basin and transferring the water through the RHR b.

heat exchangers. ECCS pump room seal coolers, and associated coolers and pump heat exchangers.

OPERATIONAL CONDITIONS 1, 2, 3, 4, 5 and *.

.A.PPLICABILITY:

ACTION:

In OPERATIONAL CONDITION 1, 2 or 3:

With one SSW subsystem inoperable, restore the inoperable a.

subsystem to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least 1.

HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOW within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

With both SSW subsystems inoperable, be in at least HOT SPUTDOWN i

2.

within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN ** within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , In OPERATIONAL CONDITION 3 or 4 with the SSW subsystem, which is associated with an RHR loop required OPERABLE by Specification 3.4.9.1 b.

or 3.4.9.2, inoperable, declare the associated RHR loop inoperable and take the ACTION required by Specification 3.4.9.1 or 3.4.9.2, as applicable.

In OPERATIONAL CONDITION 4 or 5 with the SSW subsystem, which is associated with an ECCS pump required OPERABLE by Specification 3.5.2, c.

inoperable, declare the associtted ECCS pump inoperable and take the ACTION required by Specification 3.5.2.

In OPERATIONAL CONDITION 5 with the SSW subsystem, which is associated with an RHR system required OPERABLE by Specification d.

3.9.11.1 or 3.9.11.2, inoperable, declare the associated RHR system 3.9.11.1 inoperable and tche the ACTION required by Specification or 3.9.11.2, as applicable.

In Operational Condition 8, with the $$W subsystem, which is associated with a diesel generator required OPERABLE by Specifica-e.

tion 3.8.1.2, inoperable, declare the associated diesel generator inoperable and take the ACTION required by Specification 3.8.1.2.

The provisions af Specification 3.0.3 are not applicable.

saca d ary es.ih 6 eJ pr: w y er When handling irradiated fuel in theg ili si Lildir.ii w hel;;.r; Lit !:;.

Whenever both SSW subsystems are inoperable, if unable to attain COL en as required by this ACTION, maintain reactor coolant temperature as low as practical by use of alternate heat removal methods.

3/4 7-1 GRAND GULF-UNIT 1

2, (GGNS-56% 322)

PLANT SYSTEMS ULTIMATE HEAT SINK LIMITING CONDITION FOR OPERATION I

3.7.1.3 Two independent SSW cooling tower basins shall be OPERA 8LE, each with:

A minimum basin water level at or above elevation 130'3" Mean Sea a.

Level, USGS datum, equivalent to an indicated level of 1 87".

b.

Two OPERABLE cooling tower fans.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3, 4, 5 and *.

ACTION:

In OPERATIONAL CONDITION 1, 2, or 3 with one SSW cooling tower basin a.

inoperable, declare the associated SSW subsystem inoperable and, if applicable, declare the HPCS service water system inoperable, and take the ACTION required by Specifications 3.7.1.1 and 3.7.1.2, as applicable.

I In OPERATIONAL CONDITION 4 or 5 with both SSW cooling tower basins b.

inoperable, declare the SSW system and the HPCS service water system inoperable and take the ACTION required by Specifications 3.7.1.1 and 3.7.1.2.

In Operational Condition

with both SSW cooling tower basins inoperable, declare the SSW system inoperable and take the ACTION c.

required by Specification 3.7.1.1.

The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS Two SSW cooling tower basins shall be determined OPERABLE at least 4.7.1.3 once per:

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying basin water level to be greater than or equal a.

i to 87".

l 31 days by starting each SSW cooling tower fan from the control room b.

and operating the fan for at least 15 minutes.

18 months by verifying that each SSW cooling tower fan starts automatically when the associated SSW subsystem is started.

c.

l secookey c,#*i**aT l

pr; - y er "When handling irradiated fuel in thep;dlieri 0;ildig er-Cac h n 2 0;iidi';-

I GRAND GULF-UNIT 1 3/4 7-4

1 2.(GGNS-56%322)

PLANT SYSTEMS 3/4.7.2 CONTROL ROOM EMERGENCY FILTRATION SYSTEM

.1MITING CONDITION FOR OPERATION

.7.2 Two independent control room emergency filtration system subsystems shall be OPERABLE.

APPLICABILITY: All OPERATIONAL CONDITIONS and *.

ACTION:

In OPERATIONAL CONDITION 1, 2 or 3 with one control room emergency a.

filtration subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , b.

In OPERATIONAL CONDITION 4, 5 or *:

With one control room emergency filtration subsystem inoperable, 1.

restore the inoperable subsystem to OPERABLE status within 7 days or initiate and maintain operation of the OPERABLE subsysterr in the isolation mode of operation.

With both control room emergency filtration subsystems inoperable, 2.

suspend CORE ALTERATIONS, handling of irradiated fuel in the gr',~ ry eresecondary containment and operations with a potential for d the reactor vessel.

The provisions of Specification 3.0.3 are not applicable in c.

Operational Condition *.

SURVEILLANCE REQUIREMENTS Each control room emergency filtration subsystem shall be demonstrated

4.7.2 OPERABLE

At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control room, flow through the HEPA filters and charcoal a.

adsorbers and verifying that the subsystem operates for at least 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months with the heaters OPERABLE.

At least once per 18 months or (1) after any structural maintenance en the HEPA filter or charcoal adsorber housings, or (2) following b.

painting, fire or chemical release in any ventilation zone communicating with the subsystem by:

e of I'

1. M h hat with the subsystem operating at 4000 cfm i 105 sting thro filters and g j flow of the system to the p,i charcoal adsorbers, th j"

facilit ss than or equal to subsystem is i

y admitting cold 00P at the system intake _. _

Q rei

=ey er Whenirradiatedfuelisbeinghandledinthefseco=ndarycontainment.

a GRAND GULF-UNIT 1 3/4 7-5

l

~

s 2.

(GGN5-567,522)

ELECTRICAL POWER SYSTEMS k.C. SOURCES - SHUTDOWN LIMITING CONDITION FOR OPERATION i.

As a minimum, the following A.C. electrical power sources shall be 3.'8.1.2 OPERABLE:

One circuit between the offsite transmission network and the ons a.

Class 1E distribution system, and Diesel generator 11 and/or 12, and diesel gener b".

A day tank containing a minimum of 220 gallons of fuel.

1.

A fuel storage system containing a minimum of:

2.

48,000 gallons of fuel each for diesel generators 11 and 12.

a) b)

39,000 gallons of fuel for diesel generator 13.

3.

A fuel transfer pump.

APPLICABILITY: OPERATIONAL CONDITIONS 4, 5 and *.

ACTION:

With all offsite circuits inoperable and/or with diesel generators 11 and/or 12 of the above required A.C. electrical power s a.

'noperable, lier, " ildia; and Cacic,; e " ilig, operations with a V

in tne'A x, potential for draining tie reactor vessel and crane o rei-ary e smJ,., y In addition, when in OPERATIONAL CONDITION 5 with the water level

- w w. r_.,

less than 23 feet above the reactor pressure vessel flange, immediately initiate corrective action to restore the required power sources to OPERABLE status as soon as practical.

With diesel generator 13 of the above required A.C. electrical power sources inoperable, restore the inoperable diesel generator 13 to b.

OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or declare the HPC5 system inopera and take the ACTION required by Specification 3.5.2 and 3.5.3.

The provisions of Specification 3.0.3 are not applicable.

c.

SURVEILLANCE REQUIREMENTS 4.8.1.2 At least the above required A.C. electrical power sources shall be demonstrated OPERABLE per Surveillance Requirements 4.8.1.1.3, except for the requirement of 4.8.1.1.2.a.5.

prWey er sacMey c,J..'.- e 7 When handling irradiated fuel in the h :ili: j h ildi g :: tr:1; n : h *1di g.

4 3/4 8-9 GRAND GULF-UNIT 1

2.,

(6GNS -569,322)

_ ELECTRICAL POWER SYSTEMS j

D.C. SOURCES - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.8.2.2 As a: minimum, Division 1 or Division 2, and, when the HPCS system is required to be OPERABLE, Division 3 of the D.C. electrical power sources i

shall be OPERABLE with:

a.

Division I consisting of:

l 1.

125 volt battery 1A3.

i 2.

125 volt full capacity charger IA4 or IAS.

j b.

Division 2 consisting of:

l 1.

125 volt battery 183.

l 2.

125 volt full capacity charger 184 or 185.

i

)

c.

Division 3 consisting of:

1.

125 volt battery IC3.

2.

125 volt full capacity charger 1C4 t

l APPLICABILITY:

OPERATIONAL CONDITIONS 4, 5 and *.

(son.hi e e-r "

^#Y #

"'Y ACTION-With both Division 1 battery and Division 2 battery of the above a.

i required D.C. electrical power sources inoperable, suspend CORE ALTERATIONS, handling of irradiated fuel in th. ---.i...,

...;ir.;

t sed Ercin.r: Liid n; and operations with a potential for draining j

the reactor vessel, f

b.

With Division 3 battery of the above required D.C. electrical power i

sources inoperable, declare the HPCS system inoperable and take the

~

l ACTION required by Specification 3.5.2 and 3.5.3.

l With the above required full capacity charger inoperable, demonstrate c.

the OPERABILITY of its associated battery by performing Surveillance i

j Requirement 4.8.2.1.a.1 within one hour and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter.

If any Category A limit in Table 4.8.2.1-1 is not met, declare the battery inoperable.

i t

j d.

The provisions of Specification 3.0.3 are not applicabis.

l SURVEILLANCE REQUIREMENTS 4.8.2.2 At least the above required battery and charger shall be demonstrated OPERABLE per Surveillance Requirement 4.8.2.1.

emT l

8 gr:e ary or S&ary con %'w When handling irradiated fuel in theg, ili.c, 0.iit h g-end % cia.a O.iiding.

l l

l GRAND GULF-UNIT 1 3/4 8-14

2, (6GNs-569,322)

ELECTRICAL POWER SYSTEMS D]STRIBUTION - SHUTDOWN LIMITING CON 01T10N FOR OPERATICN As a minimum, the following power distribution system divisions shall p.t.3.2 M energized:

For A.C. power distribution, Division 1 or Division 2, and when a.

the HPCS system is required to be OPERABLE, Division 3, with:

1.

Division I consisting of:

4160 volt A.C. bus 15AA.

a) b)

480 volt A.C. MCCs 15B11, 15821, 15B31, 15841, 15B51 and 15861.

c) 120 volt A.C. distribution panels in 15P11, 15P21, 15P31, 15P41, 15P51 and 15P61.

d)

LCCs 158A1, 15BA2, 15BA3, 15BA4, 15BA5 and 15BA6.

2.

Division 2 consisting of:

4160 volt A.C. bus 16AB.

a) b)

480 volt A.C. MCCs 16B11, 16821, 16B31, 16841, 16B51 anc 16B61.

c) 120 volt A.C. distribution panels in 16P11, 16P21, 16P31, 16P41, 16P51 and 16P61.

d)

LCCs 16BB1, 16BB2, 16BB3, 16BB4, 16BB5 and 16BBS.

l 3.

Division 3 consisting of:

a) 4160 volt A.C. bus 17AC.

b) 480 volt A.C. MCCs 17S01 and 17811.

c) 120 volt A.C. distribution panels 17P11.

The DPERABLE load shedding and sequencing panel associated with 4.

the division (s) required to be energized.

For D.C. power distribution, Division 1 or Division 2, and when b.

the NPCS system is required to be OPERABLE, Division 3, with:

Division I consisting of 125 volt D.C. distribution panel IDA1 1.

and IDA2.

Division 2 consisting of 125 volt D.C. dist,ribution panel IDB1 2.

and 1082.

Division 3 consisting of 125 voit D.C. distribution panel IDC1.

3.

APPLICABILITY: OPERATIONAL CONDITIONS 4, 5 and *.

O* r t end%~ t *T When handling irradiated fuel in the[ ri?i y or sea

,wa y Sef ?d! s-and 5--?:'r : 5 f?d W GRAND GULF-UNIT 1 3/4 S-17

2.- (GGNS-569 522) 3 ELECTRICAL POWER SYSTEMS

_ LIMITING CONDITION FOR OPERATION (Continued) f ACT10N:

a.

For A.C. power distribution:

With both Division 1 and Division 2 of the above required A.C.

1.

distribution system not energized and/or with the load shedding and sequencing panel associated with the division (s) required to be energized inoperable, suspend CORE ALTERATIONS, handling of irradiated fuel in the 'r"f:, S??f' ; crf :'::; :

Lildin; and operations with a potential for draining theeewaJary <*ar.% e,7 reactor vessel.

6=ry er With Divisic.n 3 of the above required A.C. distribution system

)

2.

not energized, declare the HPCS system inoperable and take the ACTION required by Specification 3.5.2 and 3.5.3.

b.

For D.C. power distribution:

With both Division 1 and Division 2 of the above required D.C.

1.

distribution system not energized, suspend CORE ALTERATIONS, handling of irradiated fuel in th '.; " ': r, S " d i n; ;.f In:?;;.r; Lilding and operation ith a potential for draining primry er seu )=ey cae N'%,,y the reactor vessel, With Division 3 of the above required D.C. distribution system 2.

not energized, declare the HPCS system inoperable and take the ACTION required by Specification 3.5.2 and 3.5.3.

The provisions of Specification 3.0.3 are not applicable.

c.

SURVEILLANCE REQUIREMENTS At least the above required power distribution system divisions shall be determined energized at least once per 7 days by verifying correct f

4.8.3.2.1 N

h ::> /**C:/;:n:12. et /kg 6ss ses/'

i

breaker ali neent :-f ::'t:;- e t':

ps.aels od veM4e eu + Le. L *a 5 LC s/MCC he above required load shedding and sequencing panel (s) shall be 4.8.3.2.2 demonstrated OPERA 8LE:

At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by determining that the auto-test system is operating and is not indicating a faulted condition.

a.

At least once per 31 days by performance of a manual test and verifying response within the design criteria to the following test b.

inputs:

a)

LOCA.

b)

Bos undervoltage.

c)

Sus undervoltage followed by LOCA.

d)

LOCA followed by bus undervoltage.

3/4 8-18 GRAND GULF-UNIT 1

l l

3. (GGNS - 52A)

SUBJECT:

Technical Specification 3/4.3.2, Table 3.3.2-1, page 3/4 3-14 DISCUSSION:

Footnote (e) of Tabic 3.3.2-1 concerns the trips trom the containment and drywell ventilation exhaust radiation monitoring system which actuate the trip system (s) to isolate the associated contsinment and drywell isolation valves. This note presently incorrectly states that "one upscale and/or two downscale actuate the trip system".

The correct logic is that "Two upscale-Hi Hi, one upscale-Hi Hi and one downscale, or two downscale signals from the same trip system actuate the trip system".

There are two trip systems and each trip system isolates its associated containment and drywell isolation valves. The footnote should be changed to indicate the correct logic.

JUSTIFICATION: Footnote (e) presently does not correctly indicate the signals which actuate each trip system. The proposed change will correct the footnote to reflect the system logic as described 4

in FSAR section 11.5.2.1.2.

SIGNIFICANT HAZARDS CONSIDERATION:

The proposed change is designated to accurately reflect Containment and Drywell Ventilation Exhaust Radiation actuation logic as described in FSAR Section 11.5.2.1.2.

This change does not involve a significant increase in the probability or consequences of an accident previously evaluated nor does it I

create the possibility of a new or different kind of accident from an accident previously evaluated. Since Table 3.3.2-1 remains unchanged in that a minimum of 2 channels per trip system are required for operability, no reduction in a margin of safety is created. Therefore, this proposed change constitutes no significant hazards consideration.

NOTE:

Technical Specification page changes marked with a PCOL number and circled are changes that were previously submitted to the NRC.

F14(3651)

8.

(GG NS -5E A )

(GG NS -569,522) l INSTRUMENTATION TABLE 3.3.2-1 (Continued)

~

150LATTON ACTUATION INSTRUMENTATION ACTION Se in at least NOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in ACTION 20 within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Close the affected system isolation valve (s) within one hour ACTLON 21

~

or:

In OPERATIONAL CONDITION 1, 2, or 3, be in at least NOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLO $NU a.

P P 'N "1 g p g.

within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

In Operational Condition *, suspend CORE ALTERATIONS

'3gg handling of irradiated fuel in the containment and b.

D"" E operations with a potential for draining the reactor vessel.

Restore the manual initiation function to 0PERABLE 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least NOT SHUTDOWN within the n ACTION 22 and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

g a

o, j

Se in at least STARTUP with the associated isolati within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or be in at least HDT SHUTDOWN within 12 y

ACTION 23

gn and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

r

.t Se in at least STARTUP within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , FACTION 24 Establish SECONDARY CONTAINNENT INTEGRITY with the standby

" ACTION 25 e> o. f >g treatment system operating within one hour.

3 y,,

Restore the manual initiation function to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or close the affected system isola

"'I F j ACTION 26 Ig dy!$,h..

Close the affected system isolation valves wit %in one hour y"3O ACTION 27 and declare the affected system inoperable.

g Lock the affected system isolation valves clos)d within one hour

}f5jACTION28 and declare the affacted system inoperable.

&n (GGNs-56322) e2

- r S* WE"y rtM 3 l

2 4 *g NOTES P e * ***

nt and during CORE When handling irradiated fuel in the3contal W "2 2 4

ALTERATIONS and operations with a potential for draining the r I

[L. f,4 During C0RE ALTERATIONS and operations with a potential far dr y

f See Specification 3.6.4, Table 3.6.4-1 for valves in each valve group.

si I h (a) reactor vessel.

A channel may be placed in an inoperable status for up te 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months fo r* I d required surveillance without placing the trip sys "g f d {

(b) oj ha M *J is monitoring that parameter.

Also actuates the standby gas treatment system.Also actu H.!q (c) oee 1J e

two huch etecte iM t-f; ce l

U (d)

- ce e (e_),,Alsolripsandisolatesthemechanicalvacuumpumps.

A channel is OPERA 8LE if 2 of 4 instruments in t (f)

Also r-tuates secondary containment ventilation isolation dampers (g) 3!

(h) Closes only RWCU s stea b?:t :20:.b*alv@i33-FC04 (1 valves per Table 3.6.6.2-1.

q qv,,

Gas Tre b 'ah h

4 iS*Ia% b A'8 Ilk'/

I 5 I

bi.

(1)

AiA wa-b %. S' g)Nb 1

1 GRA

4. (CGNS - 650)

SUBJECT:

Technical Specification Table 3.3.3-3, page 3/4 3-30.

DISCUSSION:

The above referenced table lists response times for the Emergency Core Cooling systems. However, no units are listed for the times. The word " seconds" should be added to the heading of the table.

JUSTIFICATION: The above described change adds clarity to the table and is consistent with the BWR-6 Standard Technical Specifications.

SIGNIFICANT HAZARDS CONSIDERATION:

The above change is purely administrative in that it corrects an error of omission of the units of the table values. The units are consistent with the BWR-6 Standard Technical Specifications. Based upon NRC example (i) of standards for Deternining Whether License Amendments Involve No Significant Hazards Consideration, this change constitutes no significant hazards considerations.

F15(3651)

4.(G6t45-650)

TABLE 3.3.3-3

[SEc8NAS)

_ EMERGENCY CORE COOLING SYSTEM RESPONSE TIME 1

LOW PRESSURE CORE SPRAY SYSTEM

$ 40 2.

LOW PRESSURE COOLANT INJECTION MODE OF RHR SYSTEM

$ 45 a.

Pumps A and B 1 40 b.

Pump C NA 3.

AUTOMATIC DEPRESSURIZATION SYSTEM 4.

HIGH PRESSURE CORE SPRAY SYSTEM i 27 MA 5.

LOSS OF POWER 3/4 3-30 GRANO GULT-UNIT 1

5. (GGNS - 451)

SUBJECT:

Technical Specification Table 4.3.3.1-1, page 3/4 3-32 and 3/4 3-33.

DISCUSSION:

Technical Specification 4.3.3.1 requires surveillance testing in accordance with Table 4.3.3.1-1 which requires functional testing and calibration of Division 1 and 2 Bus undervoltage protection. There are three (3) levels of undervoltage protection provided.

Each level of protection has a time delay associated with the trip point. These time delays are provided by the Load Shedding and Sequencing (LSS) Panel. The design of this panel precludes testing and calibration of the built-in time delay function.

The Technical Specification should be changed to add a note (e) to the monthly surveillance and a note (f) to the refueling surveillance as follows:

(e) Functional testing of time delay not required.

(f) Calibration of time delay not required. Time delay to be verified by channel functional test.

JUSTIFICATION: The LSS panel is a solid-state digital system utilizing printed circuit cards for logic and timing functions.

It is not possible to check individual timers with the LSS panel. Time delays can only be determined by measuring the time between a test input and panel output. This requires actual panel actuation (load shedding and sequencing); therefore, this testing cannot be performed during operation. The panel is equipped with an automatic test function which provides continuous operational surveillance. This testing verifies functioning of the time delays. Technical Specification 4.8.3.1.2.a requires verifying the operation of the automatic test system at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

The timing sequences and time delays within the LSS panel are actually programmable (rocker switches) printed circuit cards; therefore, time delays are a programmed constant and cannot be calibrated. The Channel Functional Test associated with channel calibration conducted at 18 months intervals will include verifying the time delay from the test signal input to panel output. This will verify proper setting of the time delay and proper operation of the printed circuit card.

SIGNIFICANT llAZARDS CONSIDERATION:

The proposed change establishes appropriate surveillance l

I requirements for Division 1 and 2 Bus undervoltage protection to meet the intent of the Technical Specification, i.e.,

l operability of the ECCS actuation instrumentation through verification of proper actuation and sequencing. The Load Shedding and Sequencing (LSS) Panel design precludes testing and calibration of the as-built time delay function.

Compensation for this design feature includes:

F16(3651)

f

~

(1) Panel automatic test function for operational surveillance with an associated Technical Specification surveillance 4

requirement for this feature.

(2) Solid state design of timing sequences and delays.

(3) Channel Functional Testing as 18 month intervals with associated time delay verification.

4 LSS Panel testing will therefore provide appropriate assurance 4

j of system operability for its unique design.

l Changes proposed in the surveillance requirements do not t

constitute a significant increase in the probebility or consequences of a previously evaluated accident nor do they create the possibility of a new or different type of accident 1

from any previously evaluated. LSS Panel features as described in this proposed change demonstrate that no significant i

reduction in safety margin will occur. Therefore, this change j

constitutes no significant hazards considerations, j

NOTE:

Technical Specification page changes marked with a PCOL number l

and circled are changes that were previously submitted to the NRC.

i i'

I i

i i

j 4

4 I

i f

I f

i f

I l

l l

F17(3651)

TAetE 4.3.3.1-1 (Continued)

E HERGENCY CORE COOLING SYSTEM ACIUATION INSTRUMENTATION SURVEILLANCE REQUIRGEMf5 i

l CHANNtt 0,ERAv a 5

CHANNEL FUNCTIONAL CHANNEL CO M ITIONS FOR WHICM TRIP FUNCTION CHECK IEST Call 0 RATION SURVEILLANCE REQUIRE 0 g

r-7

s. DIVISION 2 TRIP SYSTEM (Contlemed)

E 2.. AUTOMATIC DEPRES5URIZATION SYSTEM U

TRIP SYSTEM "8"#

e a.

Reactor Vessel Water Level -

R *))

1, 2, 3 I

Low Low Low, level 1 5

M IR*

1, 2, 3 b.

Drywell Pressure-High 5

M c.

AOS Timer NA M

Q 1,2,3 j

d.

Reactor Vessel Water Level -

R *I 1,2,3 I

l Low, level 3 5

M e.

LPCI Pump 5 and C Discharge

"( }

I' I' 3 Pressure-High 5

M(b) f.

Manual Initiation NA M

NA 1, 2, 3 1

}

C.

OIVISION 3 TRIP SYSTEM 1.

HPC5 5YSTEM a.

Reactor Vessel Water Level -

Y low Low, Level 2 5

M R

1,2,3,4*,5*

l M

b.

Orywell Pressure-High 5

M R(,)

1, 2, 3 c.

Reactor Vessel Water 5

M R

1, 2, 3, 4*, 5*

Level-High, Level 8 l

d.

Condensate Storage Tank R,)

1, 2, 3, 4*, 5*

g Level - Low 5

M l

e.

Suppression Pool Water R,)

1, 2, 3, 4*, 5*

4 g

l Level - High 5

M(b) l f.

Manual Initiation MA M

NA 1, 2, 3, 4*, 5' h

O.

LOSS OF POWER

,m i

1.

Division 1 and 2 Mg Rg 1, 2, 3, 4**, 5**

2 l

a.

4.16 kV pus Undervoltage NA (p

(Loss of Voltage) 6)

1, 2, 3, 4**, 5** -

i i

b.

4.16 kV Bus Undervoltage NA M

R K

1 (OOP lead Shed)

MN Rg) 1, 2, 3, 4**, 5**

m i

c.

4.16 kV Bus Undervoltage NA W

i (Oegraded Voltage)

U 1

2.

Division 3 a.

4.16 kV Bus Undervoltage NA NA R

1, 2, 3, 4a*, 5**

2 (Loss of Voltage) l i

1 i

~

Sp.f. (6GNS-451)

TABLE 4.3.3.1-1 (Continued)

EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS NOTATION Not required to be OPERABLE when reactor steam done pressure is less than p

o or equal to 135 psig, 1"

bQ,Aen the system is required to be OPERABLE, :ft:r 5:f n; :: r:11;/

t a

= :!!;m:d, :: ;;M:251:, per Specification 3.5.2g dr 3.5.3.

t Required when ESF equipment it required to be OPERABLE.

Calibrate trip unit at least once per 31 days.

(a)

Manual initiation switches shall be tested at least once per 18 months (b)

All other circuitry associated with manual initiation during shutdown.

shall receive a CHANNEL FUNCTIONAL TEST at least once per 31 days as a part of circuitry required to be tested for automatic system actuation.

Manual initiation test shall include verification of the OPERABILITY of (c) the LPCS and LPCI injection valve interlocks.

This calibration shall consist of the CHANNEL CALIBRATION of the LPCS and (d)

LPCI injection valve interlocks with the interlock setpoint verified to be

< 150 psig, h YuNc170cet TESTin6 Of" 7tur }ELAY AIC7~ $[GullEb.

$k 16LATood OF~ 7tMX b[t4 ^10T REQutub. $c wy NWNEl Yddl7tDN4L TE5T, CoAto vc Tso Tb fC /EtiftEb h) NNEL

/ts f'A R T oF c_M A C Au B RATr op.

l l

GRAND GULF-UNIT 1 3/4 3-33 I

6. (GGNS - 52B)

SUBJECT:

Technical Specification 3/4.3.7, Table 3.3.7.1-1, pages 3/4 3-56 and 3/4 3-57.

DISCUSSION:

The alarm / trip setpoint for instrument 4 (offgas post-treatment radiation monitor) should be changed as follows:

The trip setpoint which is presently indicated as a (Hi Hi) setpoint should be changed to (Hi Hi Hi). As discussed in FSAR section 11.5.2.2.2, the trip logic is based on the Hi Hi Hi and downscale trip circuit outputs, and closes the offgas system discharge and drain valves.

The Hi and Hi Hi Hi trips are annunciated in the control room.

(The Hi Hi is also annunciated in the control room).

The comma (,) between the Hi alarm and the Hi Hi Hi trip setpoint should be replaced by a slash (/) to clearly distinguish between the alarn and trip setpoint values.

In conjunction with the change in the trip setpoint nomenclature to Hi Hi Hi, the footnote (a) should be changed accordingly to indicate Hi Hi Hi.

The present footnote (a) does not adequately explain the trip logic for instruments 7, 8, and 9 (containment and drywell ventilation exhaust radiation monitor; fuel handling area I

ventilation exhaust radiation monitor; and fuel handling area pool sweep exhaust radiation monitor). Each of these have the same trip logic which is "Two upscale-Hi Hi, one upscale-Hi Hi and one downscale, or two downscale signals from the same trip system actuate the trip system." Each trip system then isolates the associated isolation valves. A new footnote (h) should be added to indicate the correct logic.

JUSTIFICATION: These changes are proposed to correctly indicate the signals which actuate each trip system. The proposed changes will reflect the system logic as described in FSAR sections t

11.5.2.1.2, 11.5.2.1.3, 11.5.2.1.4, and 11.5.2.2.2.

SIGNIFICANT HAZARDS CONSIDERATION:

This proposed change corrects the nomenclature used in the Technical Specification with respect to actuation logic signals l

to be consistent with the discussion found in the FSAR. No I

change to the actual initiation logic is involved. This change is purely administrative in nature as described by NRC example (i) of amendments not likely to involve significant hazards considerations. Therefore, this change constitutes no l

significant hazards considerations.

NOTE:

Technical Specification page changes marked with a PCOL number and circled are changes that were previously submitted to the NRC.

F18(3651)

.~

TA8tE 3.3.7.1-1

"*"" '" """ """'"' '"5'"""'""

E MINIMUM CHANNELS APPLICABLE ALARM / TRIP E ASUREM NT,,..

E INSTRLRIENTATION OPERA 8tE CON 011 IONS SETPOINT RANGE ACTION

?

1.

Component Cooling 10 E

Water Radiation 6

Q Monitor 1

At all times 11 x 10$ cpe/NA

/to10 cpm 70 2.

$taney Service Water

/d System Radiation 5

6 Monitor 1/ heat 1, 2, 3, and* 11 x 10 cpe/NA X to 10 cpm 70 exchanger train 3.

Offgac Pre-treatment 3

0 Radiation Monitor 1

1, 2 15 x 10 mR/hr/NA 1 to 10 mR/hr 70 No fo 2,I 1, 2 11 x 10 cpm (Hi)[/to106,,,

73 g

4.

Offgas Post-treatment I

Radiation Monitor 11.0 x 10 cpm (#4_NL) g 6

M: n. m 6

Rad i

tor 1

1, 2 1 2 x fu11 pouer 1 to 10 mR/hr 72

=

background /NA w

~2 2

tf ad itor 2

1,2,3.5 and** 14 mR/hr 10 to 10 mR/hr 73 15 mR/hr D

b 7.

Containment and Drywell 4

p.

-2 2

$r I

At all times 12.0 mR/

10 to 10 mR/hr 74 3

ed

$4 mR/hr z

g

-2 2

8.

Fuel Mandling Ares-3 2.3,5 W a*

< Wh/

10 to 10 mR/hr, 75.

venttiation Exhauss Radiation Monitor 14 mR/hr(d)#

9.

Fuel Mandling Area Pool h

-2 2

Sweep Exhaust Radiation N (c) i 18 mR/hr/

10 to 10 mR/hr 75 Monttor 3

NI#

135 mR/hr

b TABLE 3.3.7.1-1 (Continued)

RADIATION MONITORING INSTRUNENTATION h

MEASUREMENT 8

MININUM CNANNELS APPLICA8tE ALARM / TRIP RANGE ACTION OPERA 8LE CONDITIONS SETPOINT_

g INSTRISEMTATION q

h

10. Area Itsnitors Fuel Mand 11ng Area Q

a.

Monitors

-2 3

10 to 10 mR/hr 72 1)

New Fuel 1

(e) 52.5 d /hr/NA s

Storege Yav1t

-2 3

<2.5 mR/hr/NA 10 to 10 mR/hr 72 2)

Spent Feel 1

(f) e d % t o 4' E

3)

,p 64.,

e Arm 1

2.5 mg4,/n f

10 to 10 mR/hr 72 At all times

$0.5 mR/hr/NA 1

4 b.

Control Room Radiation Monfter se g

p,:

,,f o c-l A

with NHM heat enchangers in operation.

When irradiated fuel is being handled in thegeconddry containment. Final Setp Any required change to Y

!S Initial setpoint.

h el a-

-.n q,7

[

this setpoint shall be submitted to Commission within 90 days after tes (a) Trips system with 2 channels upscale-M p k? ? or one channel upscaleb nd one dwasc.le l

a k 4.u; H s M.4:'

2 channels '-

t ?-.dow.a n,ce3e..

(b) Isolates containment /drywell purge penetrations.

y (c) With irradiated fuel in spent fuel storage pool.(d) Alse isolates the seco 4

(e) With fuel in the new fuel storage vanit.

(f) with fuel in the spent fuel storage pool.

l}e (3) Y4 4. *.* 4e. tc9er Wr$e Are.

- l <-

\\

g hapm.le. WW.3 or:e opm\\e. %;%. U oue boseJe,. L o clooo3.ju.,1 js (h)h +ke s, ~. 4m sy s f% a< bit _ ue %.:p sy 3h.s o g ; o.,9*. g. *

o f he usee: b1, s,.t J% vst m *

(GGNS-52 b)

(GG NS-569,322)

s 1

-l s

I

7. (GGNS - 535) 3

SUBJECT:

-Technical Specificatien 3/4 3.7.12, Table 4.3.7. U-1, page 3/4 t

]

3-94.

s DISCUSSION:

Instrument 7 (the Oftgas System post-treatment monitors) of Table 4.3.7.12,1 has a ## symbol pertaining to its Channel-Calibration (at an 18 cronth frequency) which imposec > a.

requirement that the sensor be calibrated for millirem / hour (mr/hr) from th'e calibration standard and then converted to reldae rate (i.e., millicuries /sec) within one week of unit s

operatiu.. This footnote ic not' appropriate for the Offgas System post-treatmenti inonitors which readout in counts per minute (CPM). The & denoted on the Channel Calibration for thia monitor should bJ deleted.

s

~

JUSTIFICATION: The footnote ## is appropriate for Instrument 6 (Offgas System.

pre-treatment monitor),which is cal 1brated in accordance with

' ~

the methodology describert in the footnote ## and which has a s

Technical Specification limit spJcified in trillicuries/second in Technical Specification 3.11.,2.7.

Footnote ##, however, is not appropriate for Instrument 7 (post-treatment monitors).

It is not nececsary to calibrate these post-treatment monitors i

in mr/hr from the calibration standard and then convert results to release rate. Th0 post-treatment monitors are calibrated using a known microcuries/ milliliter (uCi/ml) cource to determine a CPM /vf'i/ sal calibrdion (sensitivity) factor which

~

is used to relat'e uCi/ml to the in,11cated CPM reading. The post-treatment monitors readout in CPM and have alarm and trip

'setpoints specified in CPM in Technical Specification Table s

3.3.7.1-1 (Instrumen: 4 - offgwo pbst-treatment radiaticn

' monitor).

-x SIGNIFICANT HAZARDS CONSIDERATION:

s As identified in FSAR table'11.5-1, the Offgas System post-treatment monitor reads out in counts per minute.

Calibration in millirem /heur is therefore inappropriate. This change is ourely administrative in that it delet'es the uce of the milltrem/ hour standard yet retains the channel calibration requirement at the existing 18 month frequency. Basedxupon this evalt'ation, this change is purely admiristrative, in nature s

9s' described by NRC example (i) of amendments not likely to involve.signifh ant hazards considerations. Therefore this change constit,utes no significant hazards consf derations.

NOTE:

Technical Specification page changes marked with a PCOL number and circled are changes that were previbusly submitted to the NRC.

1 l

s '

i l

F19(3651) j

r TABLE 4.3.7.12-1 (Continued)

RADICACTIVE GASEOUS EFFLUENT NOMITORING INSTRUNENTATION SURVEILLANCE

/

-ICOES IN tellCN CHANNEL CNANNEL SOURCE CHANNEL FUNCTIONAL SURVEILLANCE CHECK _

CHt;X CALIBRATION TEST REQUIRED INSTRtRENT a

M 6.

OFFGAS PRE-TREATMENT MONITOR Noble Gas Activity Monitor D

N R(3)

Q(2) a.

7.

OFFGAS POST-TREATMENT MONITOR Noble Gas Activity Monitor a

a.

Providing Alarm and Auto-l matic Temination of Release D

N R(3 Q(1) 4

.1 Y

2 Tk.s c.k o3=. w s pre: sly Pto t.1Vo r rep. sh A : p I A~bJ Apn7,itr3 8e o

Z (P

i UI be

():

v

8. (GGNS - 589)

SUBJECT:

Technical Specification 3.4.2.1, page 3/4 4-5.

DISCUSSION:

The limiting condition for operation of Technical Specification 3.4.2.1 specifies the number of safety / relief valves which shall be operable and also provides the lift settings for the safety / relief valves. The lift sectings are provided as a setpoint (in psig) plus or minus one percent (11%).

The !1% (i.e., 11.6 psi for a 1165 psig setpoint, 11.8 psi for a 1180 psig setpoint, and 11.9 for a 1190 psig setpoint) is appropriate for the safety valve function in this specification. However, the 1% is more conservative than the i

drift allowance assigned to the setpoints for the relief valve function in this specification. Technical Specification 3.4.2.1 should be amended to reflect the 15 psi tolerance (drift allowance) specified in the General Electric design specification for the setpoints for the relief valve function and the 11.6, 11.8, and 11.9 psi setpoints for the safety valve setpoints.

JUSTIFICATION: The General Electric design specification stipulates a nominal setpoint which is consistent with the relief valve function setpoint in 3.4.2.1, with a drift allowance of !15 psi versus the 11% (approximately 11.0 to 11.2 psi) in 3.4.2.1.

This proposed change will provide the correct tolerance (drift allowance) of !15 psi for the relief valve function in this specification.

SIGNIFICANT HAZARDS CONSIDERATION:

The proposed change is administrative in that it revises the drift allowance for the relief valve function of the safety /

relief valves to coincide with the design specification for these valves. This minor increase in drift allowance is well within the analytical limits described in FSAR Section 5.2.2.2.2.4 which were assured in performing the transient overpressure analyses.

Consequently, no significant reduction in safety margin is created. Therefore, this change constitutes no significant hazards considerations.

(

F20(3651)

8. (GGNS-587)

REACTOR COOLANT SYSTEM 3/4.4.2 SAFETY VALVES SAFETY / RELIEF VALVES LIMITING CONDITION FOR OPERATION j

3.E.2.1 Of the following safety / relief valves, the safety valve function of et'least 7 valves and the relief valve function of at least 6 valves oth those satisfying the safety valve function requirement shall be OPERABLE with the specified lift settings:

Number of Valves Function Setpoint* (psio) 8 Safety 1165

.r//.6fmA l

6 Safety 1180 r //.S p64 6

Safety 1190 s //.9 psi Relief 1103 x is psi 1

1113 x 15 Asl Relief 10 1123 2 /6 psi Relief 9

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 3.

ACTION:

With the safety and/or relief vaTve function of one or more of the above required safety / relief valves inoperable, be in at least HOT SHUTDOWN l

a.

within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , With one or more safety / relief valves stuck open, provided that suppression pool average water temperature is less than 105'F, close the stuck open b.

relief valve (s); if unable to close the open valve (s) within 2 minutes or if suppression pool average water temperature is 105'F or greater, place the reactor mode switch in the Shutdown position.

With one or more safety / relief tail-pipe pressure switches inoperable, restore the inoperable switch (es) to 0PERABLE status within 7 days or be c.

in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COL within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE REQUIREMENTS The tail-pipe pressure switch for each safety / relief valve shall be demonstrated OPERABLE with the setpoint verified to be 30 1 5 psig by 4.4.2.1.1 performance of a:

CHANNEL FUNCTIONAL TEST at least once per 31 days, and a CHANNEL CALIBRATION at least once per 18 months.**

a.

b.

The relief valve function pressure actuation instrumentation shall 4.4.2.1.2 be demonstrated OPERA 8LE by performance of a:

CHANNEL FUNCTIONAL TEST, including calibration of the trip unit, at 4

least once per 31 days.

CHANNEL CAL 18 RAT 10N, LOGIC SYSTEN F b.

'The lift setting pressure shall correspond to ambient conditions of the valves at nominal operating temperatures and pressures.

- The provisions of Specification 4.0.4 are not applica adequate to perform the test.

3/4 4-5 GRAND GULF-UNIT 1

9. (GGNS - 502)

SUBJECT:

Technical Specification 3.4.2.2, page 3/4 4-6.

DISCUSSION:

The limiting condition for operation of Technical Specification i

3.4.2.2 specifies the relief valves with the low-low set function which shall be operable and also provides the low-low set function lift settings for these relief valves. The low-low set function lift settings are provided as a setpoint (in psig) plus or minus one percent (11%).

The !1% is appropriate for the safety valve function for these valves as presented in Technical Specification 3.4.2.1.

However, the 11% is more conservative than the drift allowance assigned to the setpoints for the relief valve low-low set function in this specification. Technical Specification 3.4.2.2 should be amended to reflect the 15 psi tolerance (drift allowance) specified in the General Electric design specification for the setpoints for the relief valve low-low set function.

JUSTIFICATION: The General Electric design specification stipulates a nominal setpoint which is consistent with the relief valve low-low set function setpoint in 3.4.2.2, with a drift allowance of !15 psi versus the 11% (approximately 10.3 to 11.1 psi) in 3.4.2.2.

This proposed change will provide the correct tolerance (drift allowance) of 15 psi for the relief valve low-low set function in this specification.

SIGNIFICANT HAZARDS CONSIDERATION:

This proposed change establishes consistency with the requested change to Technical Specification 3.4.2.1, page 3/4 4-5 (GGNS-589) in that it incorporates the drift allowance for the relief valve function of the safety / relief valves to coincide with their design specification. This change is administrative in nature and for the reasons discussed in the above mentioned proposed change does not constitute significant hazards consideration.

F21(3651)

-- ~ -. - _

~

l l.

E(GG NS -302)

's REACTOR COOLANT SYSTEM i

i SAFETY / RELIEF VALVES LOW-LOW SET FUNCTION LIMITING CONDITION FOR OPERATION 3.4.2.2 The: relief valve function and the low-low set function of the following reactor coolant system safety / relief valves shall be OPERABLE with the following low-low set function lift settings:

Set>oint* (psia)M t-tipsi.'

Valve No.

) pen uose i

F051D 1033 926 F051B 1073 936 F047D 1113 946 F047G 1113 946 F051A 1113 946 F051F 1113 946 APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 3.

ACTION:

With the relief valve function and/or the low-low set function of one of the above required reactor coolant system safety / relief valves inoperable, a.

restore the inoperable relief valve function and the low-low set function to OPERABLE status within 14 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

With the relief valve function and/or the low-low set function of more than one of the above required reactor coolant system safety / relief valves b.

inoperable, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE REQUIREMENTS The relief valve function and the low-low set function pressure 4.4.2.2.1 actuation instrumentation shall be demonstrated OPERABLE by performance of a:

CHANNEL FUNCTIONAL TEST, including calibration of the trip unit, at least a.

once per 31 days.

LOGIC SYSTEM FUNCTIONAL TEST and simulated aut 4

CHANNEL CALIBRATION, ire system at least once per 18 months.

b.

operation of the ent 1

"Ine lin setting pressure shall correspond to ambient conditions of the valves at nominal operating temperatures and pressures.

GRAND GULF-UNIT 1 3/4 4-6

10. (CCNS - 605, 606)

SUBJECT:

Technical Specifications 3.6.1.3 and 3.6.2.3, pages 3/4 6-5 and 3/4 6-6, and pages 3/4 6-15 and 3/4 6-16.

DISCUSSION:

Technical Specifications 3.6.1.3 and 3.6.2.3 refer to the containment and drywell air lock door inflatable seal system

" air flask" pressure instrumentation channels. The GGNS design for the containment and drywell air lock door seal systems does not provide instrumentation channels for the air lock door seal system air tanks (flasks). Instrumentation is provided instead for each air lock door inflatable seal (i.e., for each of the two seals on both doors of the air lock). The Technical Specifications should be amended to reflect the correct instrumentation for the air lock door seal systems.

Surveillance Requirements 4.6.1.3.d.3 and 4.6.2.3.d.3 specify surveillance leak tests of the air lock door seal pneumatic systems from an initial pressure of 104.7 psig. This initial pressure should be changed to 90.0 psiji corresponding to the air supply pressure to the seal system of 90 psig to 100 psig.

JUSTIFICATION: The proposed change to the Technical Specifications correctly specifies the instrumentation channel provided for each air lock door inflatable seal in accordance with the GGNS design.

The change also indicates the initial pressure for conduct of the surveillance leak tests of the air lock door seal pneumatic systems as 90.0 psig. The pressure of the air supply to the air lock door seals will be between 90 psig and 100 psig. The corresponding starting pressure for the leak test should be within the pressure range for the air supply to the door seals.

SIGNIFICANT HAZARDS CONSIDERATION:

The proposed change corrects an error in the units for the initial pressure during the seal pneumatic system leak rate r

test. This is purely an administrative change. The change also modifies the Technical Specifications to be consistent with existing plant design. No change in the margin of safety is created since the intent of the surveillance requirement, i.e., assuring air lock operability and integrity, is still served. Since this change does not involve a significant increase in the probability or consequences of an accident previously evaluated nor create the possibility of a new or different kind of accident from any accident previously evaluated, this change does not constitute a significant hazards consideration.

F22(3651)

Igf.(GGN5-6o5, Go6) s.

CONTAINMENT SYSTEMS CONTAINMENT AIR LOCKS LII(1TINGCONDITIONFOROPERATION 3.6.1.3 Each containment air lock shall be OPERABLE with:

Both doors closed except when the air lock is being used for normal s.

transit entry and exit through the containment, then at least one air lock door shall be closed, and An overall air lock leakage rate of less than or equal to 2 scf per b.

hour at P,. 11.5 psig.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2* and 3.

ACTION:

With one containment air lock door inoperable:

a.

Maintain at least the OPERABLE air lock door closed and either restore 1.

the inoperable air lock door to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or lock the OPERABLE air lock door closed.

Operation may then continue until performance of the next required 2.

overall air lock leakage test provided that the OPERABLE air lock door is verified to be locked closed at least once per 31 days.

Otherwise, be in at least NOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and 3.

in COLO SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

The provisions of Specification 3.0.4 are not applicable.

l 4.

With the containment air lock inoperable, except as a result of an inoperable air lock door, maintain at least one air lock door closed; b.

restore the inoperable air lock to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least NOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SH within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SEAL With one containment air lock door inflatable seal system Of

h:'-

pressure instrumentation channel inoperable, restore the c.

t t's be 3,60 psig at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

f I Tnt AssetteTO IsJFs41s&LE' sitAL

' See Special Test Exception 3.10.1.

3/4 6-5 GRAND GULF-UNIT 1

/0,f.2[GGNS -605 Go6) 3 CONTAlle4ENT SYSTEMS SURVEILLANCE REQUIRENENTS Eac'h containment air lock shall be demonstrated OPERABLE:

4.

1.3 uthin 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after each closing, except when the air lock i,s being used for multiple entries, then at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , by a.

verifying seal leakage rate less than or equal to 2 scf per hour when the gap between the door seals is pressurized to Pa, 11.5 psig.

Fy conducting an overall air lock leakage test at P,w,11.5 psig, and ithin its limit:

b.

verifying that the overall air lock leakage rate is At least once oer 6 months #,and 1.

Prior to attablishing PRIMARY CONTAINMENT INTEGRITY when 2.

maintenance has been performed on the air lock that could affect the air lock sealing capability."

At least once per 6 months by verifying that only one door in each c.

air lock can be onened at a time.

EUH 4telotx)le seal system OPERABLE by:

un iatsb d.

By verifying we anorGxH orTuDArarnstg]

Demonstratingftwo geil W L 3 pressure instrumentation 1.

channel PERABLE by performance of a:

CHANNEL FUNCTIONAL TEST at least once per 31 days, and fCL RIhDCZ a) gg b)

CHANNEL CALIBRATION at least once per 18 months, with a low pressure setpoint of > 60 psig.

At least once per 7 days, verifying seal air flask pressure to 2.

be greater than or equal to 60 psig.

At least once per 18 months, conducting a seal pneumatic system leak test and verifying that system pressure does not decay 3.

more than 2 psig from 104.7 es4g within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

S'A The provisions of Specification 4.0.2 are not applicable.

I Exemption to Appendix J of 10 CFR 50.

m l

3/4 6-6 GRAND GULF-UNIT 1

l t

1a 3@*6-' 5,6%)

r CONTAINMENT SYSTEMS

(

DRYWELL AIR LOCKS LIMITING CONDITION FOR OPERATION

?

T.

3.6.2.3 Each drywell air lock shall be OPERABLE with:

Both doors closed except when the air lock is being used for normal transit a.

entry and exit through the drywell, then at least one air lock door shall be closed, and An overall air lock leakage rate of less than or equal to 2 scf per hour b.

at P,, 11.5 psig.

APPLICABIllTY: OPERATIONAL CONDITIONS 1, 2* and 3.

ACTION:

With one drywell air lock door inoperable:

l a.

Maintain at least the OPERABLE air lock door closed and either restore 1.

the inoperable air lock door to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or lock the OPERABLE air lock door closed.

Operation may then continue provided that the OPERABLE air lock door 2.

is verified to be locked closed at least once per 31 days.

3.

Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

4.

The provisions of Specification 3.0.4 are not applicable.

With the drywell air lock inoperable, except as a result of an inoperable b.

air lock door, maintain at least one air lock door cDsed; restore the inoperable air lock to OPERABLE status within 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . or be in at least NOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within th following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

f

M-' pressure With one drywell air lock door inflatable seal syste instrumentation channel inoperable, restore the inoperable channel to c.

& 'L.2 pressure to be > 60 psig OPERABLE status within 7 days or verif pt least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

7 I Tdt 45GottAT&

IMhATA6Lt' t (4c.

j 5

"See special Test Exception 3.10.1.

GRAND GULF-UNIT 1 3/4 6-15

l@4(G6NS-Go5,6o6)

CONTAINNENT SYSTEMS r

SURVEILLANCE REQUIRENENTS l

4.(.2.3 Each drywell air lock shall be demonstrated OPERABLE-I a.

Within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months after each closing, except when the air lock is being i

used for multiple entries, then at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , by-verifying seal leakage rate less than or equal to 2 scf per hour when the gap between the door seals is pressurized to P,, 11.5 psig.

At least once per 6 months by conducting an overall air lock b.

leakage test at P,is within its limit, 11.5 psig and by vgrifying tha lock leakage rate At least once per 6 months by verifying that only one door in each c.

air lock can be opened at a time.

Est ArdLDC &

d.

By verifyin %e door intlatable seal svsten. OPERABLE by:

@H erTMc JdalrLAYn&LE)

Demonstraung4twopeal & P..a pressure instrumentation 1.

channel OPERABLE by performance of a:

Alb a)

CHANNEL FUNCTIONAL TEST at least once per 31 days, and I

D002 b)

CHANNEL CALIBRATION at least once per 18 months, with a low pressure setpoint of > 60 psig.

At least once per 7 days verifying seal air flask pressure to 2.

be greater than or equal to 60 psig.

At least once per 18 months, conducting a seal pneumatic system 3.

leak test and verifying that system pressure does not decay more than 2 psig from 104. ps.ig within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

Slh The provisions of Specification 4.0.2 are not applicable.

GRAND GULF-UNIT 1 3/4 6-16 l

11. (GGNS - 674)

SUBJECT:

Technical Specification 4.6.6.3.d.3, page 3/4 6-54 DISCUSSION:

The proposed change to Technical Specification 4.6.6.3.d.3 would add " manual initiation" to the list of SGTS actuation test signals. The proposed change would also indicate that, by LOGIC SYSTEM FUNCTIONAL TEST, the SGTS filter train would be verified to automatically start and that the system isolation dampers open.

JUSTIFICATION: The addition of the manual initiation to the list of system actuation test signals will provide for testing of the SGTS functions with a manual initiation, in addition to the present actuation test signals, to be consistent with section 6.5.3.3 of the FSAR.

The addition of the phrase "by LOGIC SYSTEM FUNCTIONAL TEST" will allow testing of the SGTS such that, in accordance with the definition of a legic system functional test, each of the actuation test signals can be tested through overlapping systems tests. This will enable verification of each entire logic system from the sensor through and including the actuation of the SGTS. In this manner, only one actuation of the filter train and isolation dampers will be necessary to meet the test requirements of 4.6.6.3.d.3.

SIGNIFICANT HAZARDS CONSIDERATION:

The proposed change clarifies the testing to be performed to verify the operability of the SGTS filter train. Logic system functional testing will allow overlapping systems tests to be performed to satisfy the surveillance requirement, however safety margin will be maintained since all actuation test signals will be tested. This change does not involve a significant increase in the probability or consequences of an accident previously evaluated nor does it create the possibility of a new or different kind of accident from any accident previously evaluated. The addition of manual initiation to the list of test signals which will be subject to 18 month surveillance is a change that constitutes an additional control not presently included in the Technical Specifications. For these reasons, this proposed change does not constitute a significant hazards consideration.

i F23(3651)

il. (GG NS - 67+)

I CONTAlletENT SYSTEMS

)

j

$URVEILLANCE REQUIRENENTS (Continued)

At least once per 18 months or (1) after any structural maintenance p.

on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or cheatcal release in any ventilation zone communicating with the subsystem by:

Verifying that the subsystem satisfies the in-place testing 1.

acceptance criteria and uses the test procedures of Regulatory Positions C.5.a, C.S.c and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978, and the system flow rate is 4000 cfm i 105.

Verifying within 31 dvs after removal that a laboratory analysis 2.

of a representative carbon sample obtained in accordance with t

Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978.

Verifying a subsystem flow rate of 4000 cfm i 10% during system 3.

operation when tested in accordance with ANSI N510-1975.

After every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal adsorber operation by verifying c.

within 31 days after removal that a laboratory analysis of a repre-sentative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52 Revision 2 March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52. Revision 2, March 1978.

d.

At least once per 18 months by:

1.

Performing a system functional test which includes simulated automatic actuation of the system throughout its emergency operating sequence for the:

a)

LOCA, and b)

Fuel handling accident.

Verifying that the pressure drop across the combined HEPA filters 2.

and charcoal adsorber banks is less than 10.75 inches Water Gauge while operatino the filter train at a flow rate of 4000 cfm 1 105.

l L

61tT&M FvMCTTOML TECT Verifying;v 662Cthat the filter train starts and isolation dampers I

3.

open on each of the following test signals:

Drywell pressure - high, a.

b.

Reactor vessel water level - low low, level 2, Fuel handling area ventilation exhaust radiation - bigh, and c.

d.

Fuel ngqiryg grea pool sweep exhaust radiation - high.

l 4

A..n

...r..v..

e 4.

8 Verifying that the fan can be manually started.

5.

Verifying that the heaters dissipate 5015.0 kW when tested in. accordance with ANSI N510-1975.

GRAND GULF-4MIT 1 3/4 6-54

12. (GGNS - 100)

SUBJECT:

Technical Specification 3.7.1.3, page 3/4 7-4 DISCUSSION:

Technical Specification 3.7.1.3 presently requires operability of two independent SSW cooling tower basins in operational conditions 1, 2, 3, 4, 5, and *. The action statements, however, do not include an action if one SSW cooling tower basin is inoperable in operational condition 4, 5, or *.

It is possible, therefore, to be unable to meet the Limiting Condition for Operation and have no action statement to enter.

The Action Statements presently do not refer to the action required by Specifications 3.7.1.1 and 3.7.1.2 when both SSW Basins are inoperable in operational conditions 1, 2, and 3.

The Technical Specification should be revised to clarify the limiting condition for operation to indicate those conditions when it is acceptable to have only one SSW basin operable in operational conditions 4, 5, and *.

A change is also preposed to Action a and Action b to indicate the respective operational conditions in which these statements apply. The limiting conditions for operation for the SSW basin in order for the basin to be operable in conjunction with the HPCS service water operability requirements should be revised.

These requirements should indicate that only the associated basin water level is required for basin operability associated with the operability of the HPCS service water.

JUSTIFICATION: Technical Specification 3.7.1.3 presently requires both SSW cooling tower basins be operable in operational conditions 4, 5, and *. This represents an unnecessary restriction to activitics in those operational conditions where only a single SSW basin is required by Technical Specification 3.7.1.1 and 3.7.1.2 to be operable. The proposed Technical Specification revision clarifies the limiting condition for operation to indicate conditions when it is appropriate for only one SSW basin to be operable in operational conditions 4, 5, and *

(i.e., when only one SSW basin is required to be operational by Technical Specification 3.7.1.1 or by 3.7.1.2).

The proposed change to Action a will refer to the action required by Specifications 3.7.1.1 and 3.7.1.2 in operational conditions 4, 5, and *. This will provide consistency in the action required by Specifications 3.7.1.1, 3.7.1.2, and 3.7.1.3.

The proposed change to Action b will refer to the action required by Specification 3.7.1.1. and 3.7.1.2 when both SSW basins are inoperable in operational conditions 1, 2, and 3, and will provide consistency in the action required by these specifications. Without this change, Specification 3.0.3 would be applicable when both SSW basins were inoperable in 1

conditions 1, 2, and 3.

F24(3651)

d i

Technical Specification Surveillance Requirement 4.7.1.3 should L

be revised to indicate that the surveillance is required for the one or the two SSW basins required to be operable per Specification 3.7.1.3.

A footnote "#" should be added to the limiting condition for operation to indicate that the basin e

cooling tower fans are not required to be operable in order for the basin to be considered operable in conjunction with HPCS service water system operability requirements. The operability of the HPCS service water system requires the' associated basin water level but does not require operability of either of the associated cooling tower fans.

SIGNIFICANT HAZARDS CONSIDERATION:

Technical Specification 3.7.1.1, 3.7.1.2, and 3.7.1.3 are closely related in that they identify operability restrictions

-and associated surveillance requirements for the plant Standby i

Service Water System. Currently, Technical' Specification 3.7.1.3 and Surveillance Requirement 4.7.1.3 are inconsistent with these other related items. This change establishes consistency between these specifications in the areas of SSW basin operability, action required in the case of inoperability, surveillance required in determining operability, and operability requirements for the HPCS Service Water System. These changes are purely administrative in nature for achieving consistency throughout the plant Technical Specifications. They constitute no significant hazards considerations as indicated by NRC examplef(i) of amendments not likely to involve significant hazards considerations.

]

h I

f I

F25(3651)

.ff 12.(GGNS-100) an j

PLANT SYSTEMS 3

- 4 ULT!*TE HEAT SINK

. $ t2' 11f!NG CONDITION FOR OPERATION Tip

-4f.

AYIme.d 4 Le f.floww3 l

3.7.1.3 4weA ndependent SSW cooling tower basins W11 h CCLE, each l

i y w-with:

1g A minimum basin water level at or above elevation 130'3" Mean Sea n p {<,U ) ** Level, USGS datum, equivalent to an indicated level of > 87".

a.

g I$ g~d b.

Two OPERABLE cooling tower fan 3 V O

.ke be OPERA BLE 7, jV RPPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3, 4, 5 and *.

d '3s'T

'4 JCTION:

o2 jhSAaa*

a p 9 %.e In OPERATIONAL CONDITION 1, 2,ee 3 pith one 55W cooling tower basin l

4 4W a.

U

,f d inoperable, declare the associated SSW subsystem inoperable and, if 9C applicable, declare the HPCS service water system inoperable, and gC take the ACTION required by Specifications 3.7.1.1 and 3.7.1.2, as 1

2EE applicable.

HHd*

I,2, 3, a

fj b.

In OPERATIONAL CONDITIO k4 or 5 with both SSW cooling tower basins l

inoperable, declare the SSW system and the HPCS service water system i

inoperable and take the ACTION required by Specifications 3.7.1.1 and 3.7.1.2.

In Operational Condition

with both 55W cooling tower basins c.

inoperable, declare the 55W system inoperable and take the ACTION required by Specification 3.7.1.1.

The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS At le.s4- + Le a.bwe reept, red

% 55W cooling tower basins shall be determined OPERABLE at least l

4.7.1.3 once per:

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying basin water level to be greater than of equal a.

to 87".

b.

31 days by starting each SSW cooling tower fan from the control roos and operating the fan for,at least 15 minutes.

~

18 months by verifying that each 55W cooling tower fan shrts c.

automatically when the associated SSW subsystem is started.

When handling tiradiated fuel in the Auxiliary Building or Enclosure Building.

r4e b ca c hg her b, are u.f res r J+. be opstAstt Ar '

l e

// pe s sc ev.co. vsnAec sy S4em cPE.RAB Wy, GRAND GULF-UNIT 1 3/4 7-4

13. (GGNS - 699)

SUBJECT:

Technical Specification 4.7.2.d.2, page 3/4 7-6.

DISCUSSION:

The proposed change to Technical Specification'4.7.2.d.2 would add " manual initiation" to the list of Control Room emergency filtration system actuation test signals, The proposed change would also indicate that, by LOGIC SYSTEM FUNCTIONAL TEST, the Control Room system would be verified to automatically switch to the isolation mode and that the isolation valves close within 4 seconds.

JUSTIFICATION: The addition of the manual initiation to the list of system actuation test signals will provide for testing of the system functions with a manual initiation, in addition to the present actuation test signals, to be consistent with section 9.4.1.3 of the FSAR.

The cddition of the phrase "by LOGIC SYSTEM FUNCTIONAL TEST" will allow testing of the Control Room emergency filtration system such that, in accordance with the definition of a logic system functional test, each of the actuation test signals can be tested through overlapping system steps. This will enable verification of each entire logic system from the sensor through and including the actuation of the Control Room system.

In this manner, only one actuation of the Control Room emergency filtration system and isolation valves will be required to meet the test requirements of 4.7.2.d.2.

SIGNIFICANT HAZARDS CONSIDERATION:

The proposed change clarifies the testing to be performed to verify the operability of the Control Room Emergency Filtration System. Logic system functional testing will allow overlapping systems tests to be performed to satisfy the surveillance requirement, however, sefety margins will be maintained since all actuation test signals will be tested. This change does not involve a significant increase in the probability or consequences of an accident previously evaluated nor does it create the possibility of a new or different kind of accident from any accident previously evaluated. The addition of manual initiation to the list of test signals which will be subject to 18 month surveillance is a change that constitutes an additional control not presently included in the Technicel Specifications. For these reasons, this proposed change does not constitute a significant hazards consideration.

I

~

NOTE:

Technical Specification page changes marked with a PCOL number and circled are changes that were previously submitted to the

NRC, F26(3651) i

13. (sms-scr9)

PLANT SYSTEMS SURVEILtANCE REQUIREMENTS (Continued) i g.

Verifying that the subsystee satisfies the in place testing h

, acceptance criteria and uses the test procedures of Regulatory

,t

' Positions C.5.a C.5.c and C.5.d of Regulatory Guide 1.52, Revision 2, h rch 1978, and the system flow rate is 4000 cfm 4n 1 105.

2.%

Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with s

Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2,

'1 March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, h rch 1978.

3,1(

Verifying a subsystem flow rate of 4000 cfm i 105 during h'

subsystem operation when tested in accordance with ANSI N510-1975.

After every'720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal adsorber operation by verifying c.

within 31 days after removal that a laboratory analysis of a repre-sentative carbon sample obtained in accordance with Regulatory Positon C.6.b of Regulatory Guide 1.52, Revision 2, hrch 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, h rch 1978.

d.

At least once per 18 months by:

1.

Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 7.2 inches Water Gauge while operating the subsystem at a flow rate of 4000 cfm i 101.

LO&TC SystfM FWCTEDHAL Te5T l

2.

Verifying 4 hat on each of the below isolation mode actuation l

test signals, the subsystem automatically switches to the isola-tion mode of operation and the isolation valves close within 4 seconds:

a)

High radiation in the outside air intake duct, b)

High chlorine concentration in the outside air intake duct, c)

High drywell pressure, and d)

Low reactor water level.

4.)

A.,al :.J%t.'m l

i 3.

Verifying that the heaters dissipate 20.7 1 2.1 W when tested l

in accordance with ANSI N510-1975.

After each complete or partial replacement of a HEPA filter bank by e.

verifying that the HEPA filter banks remove greater than or equal to 99.95% of the DOP when they are tested in-place in accordance with ANSI N510-1975 while operating the system at a flow rate of 4000 cfm t 105.

i f.

After each complete or partial replacement of a charcoal adsorber bank by verifying that the chucoal adsorbers remove 99.95% of a halogenated hydrocarbon refrigertnt test gas when they ere tested in-plac;e in accordance with Ah51 N510-1975 while operating the system

(

at a flow rate of 4000 cfm i 105.

CRAND GULF-UNIT 1 3/4 7-6

l

14. (GGNS - 651)

SUBJECT:

Technical Specification Table 3.7.4-2, page 3/4 7-16.

DISCUSSION:

Technical Specification Table 3.7.4-2 provides a listing of safety related mechanical snubbers. Due to additional snubbers being added to the recirculation system, this table is no longer accurate. Two (2) additional snubbers (AQ1B33G35501 and AQ1B33G318R01) were added to 3/4" DCB-24 sensing lines of the Recirculation System (B33). Design Change Request (DCR) 82/784 identified these lines as experiencing excessive vibration during nonnuclear startup.

JUSTIFICATION: The note (*) on Technical Specification Table 3.7.4-1 indicates that snubbers may be added to safety related systems without prior Licensee Amendment to Table 3.7.4-1 provided that a revision to Table 3.7.4-1 is included with the next License

. Amendment request. This note is interpreted to also apply to Table 3.7.4-2 for safety related mechanical snubbers and should for clarification be added to page 3/4 7-16.

SIGNIFICANT HAZARDS CONSIDERATION:

This proposed change is purely administrative in nature as it adds a footnote inadvertently omitted from the bottom of the table. The footnote is is in accordance with the Standard Technical Specifications. This change conforms to NRC example (i) of amendments not considered to involve significant hazards considerations. Therefore, this change does not constitute a significant hazards consideration.

I I

F27(3651) l

T A8i t 3.1.4-2 j

5 SAFETY RFLATED MECHANICAL SNUB 8ERS*

SNUB 8ER

$Nyggg4 NO.

AREA ELEVATION NO.

AREA ELEVATION e

RECIRCULATION SYSTEM (Continued) g a.

RECIRCULATION SYSTEM Q1833G023R01 11 117 Q1833G112R02 11 101 Q1833G023R01 11

!!7 QlR33G124R01 11 122 l

Q1833G024R01 11 102 QlR33G128C01 11 121 Q1833G024R02 11 102 Ql833G128C01 11 121 Q1833G024R02 11 102 Q1833G129C01 11 121 Q1833G024R05 11 101 Ql833G262R02 11 103 Ql833G105C01 11 101 Q1833G265001 11 102 g

i Q1833G105R01 11 101 Ol833G265R04 11 101 g

Q1833G105R02 11 101 Q1833G265R05 11 112

)

Q1833G105R02 11 101 QlB33G322R01 11 112 V

Y Q1833G108C01 11 101 Ol833G322R01 11 112 Q1833G108R01 11 101 Q1833G311R02 11 111 Q1833G10BR01 11 101 QlR33G331R02 11 109 g

Ql833G108R01 11 101 Q1H33G339R01 11 111 QIB13G109R02 Ii 101 Q1833G346R01 11 105 i

UlB33GlpMR02 11 101 913176 355Roi ga oo QlB3361E S M*C ll 800 ba.

aJJa) to s h ty relata) syste m s s.n % er prio n-Liam Q SmLbos my A*** S M ta' %b!*

3 7. E - 1.

yee ;Je) rf,ar

,.a.eisle '

To T Ua 2. 1. 4

~2-t e

is

,.e.l.a)*)

lTf, 71,a ua,r Lw

A_g, l

i

15. (GCNS - 242)

SUBJECT:

Technical Specification 4.7.7.2; page 3/4 7-41 DISCUSSION:

The Surveillance Requirements for fire doors contained in Technical Specification Section 4.7.7.2 should be revised to agree with the requirements contained in Subsection N,Section III, of Appendix R to 10CFR Part 50 The Surveillance Requirements are being revised to clarify that only one of the verifications of operability per 4.7.7.2.a thru d is required for each fire door and to clarify the requirement for 1

semiannual surveillance of each fire door automatic hold-open, release, and closing mechanism and latch in Specification 4.7.7.3.

JUSTIFICATION: The proposed changes to the Technical Specification Surveillance Requirements assure that the requirements in Appendix R to 10CFR Part 50 relating to surveillance of fire doors are correctly implemented through one of the surveillance requirements of 4.7.7.2 and through 4.7.7.3.

The requirement in 4.7.'i.3 has been separated to assure that the fire door hold-open, release, and closing mechanisms and latches are verified to be operable semiannually for each fire door. This requirement is also in accordance with Subsection N,Section III of Appendix R to 10 CFR Part 50.

SIGNIFICANT HAZARDS CONSIDERATION:

This propcsed change documents GGNS compliance with 10CFR50 Appendix R, Section III, Subsection N with respect to surveillance of fire doors. GGNS compliance is documented in FSAR Table 9A-4.

This change does not involve a significant increase in the probability or consequences of an accident previously evaluated nor does it create the possibility of a new or different kind of accident from an accident previously evaluated. Since the revised Technical Specification is in complete agreement with the appropriate requirements of 10CFR50 Appendix R, no significant reduction in the margin of safety is involved. For these reasons, this proposed change does not l

constitute a significant hazards consideration.

l 4

F28(3651)

15.g; css-24z)

)

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

E'pch of the ab e required fire doors shall be verified OPERABLE byl l

M7.i.2 e ue *I f lse [e//ow;w 3 ;

g Verifying the position of each closed fire door at least once per a.

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b.

Verifying that doors with automatic hold-open and release mechanisms are free of obstructions at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Verifying the position of each locked closed fire door at least once c.

per 7 days.

d.

Verifying OPERABILITY of the fire door supervision system by performing a CHANNEL FUNCTIONAL TEST at least once per 31 days.

l I

I h I

.... E..I..$.3__ E1??_ l!3.N5, Fi92li '"' "I ; " ^ "2

^

E7 Es.c.L of J Lc s ke. re9 0:rebhoors si \\\\ k la9ebeb w

4. 7. 7. 5 Per (a soak 'fe Ver'dy } k + co lo mod.c hold -oges, rele s.e., o ccl cles q me cLu:sm3 uel 1.f cLes re l

OPE R. AB L. E,

i GRAND GULF-UNIT 1 3/4 7-41

l.

l

16. (CGNS - 479, 480, 486)

SUBJECT:

Technical Specification 4.8.4.3.b. page 3/4 8-46.

DISCUSSION:

Technical Specification 4.8.4.3.b establishes the setpoints and tolerances for the Reactor Protection System (RPS) electrical protection assembly (EPA) over-voltage, under-voltage, and under-frequency protective instrumentation. The basis of these setpoints is 110% of nominal for over-voltage, 97.5% of nominal for under-voltage, and 95% of nominal for under-frequency. Nominal is defined as 120 VAC at 60 hertz.

With the under-voltage setpoint of 117 VAC and a 12 VAC voltage drop from the RPS bus to the scram solenoids, a minimum voltage of 105 VAC is provided at the scram solenoids for proper operation of these devices. Although the setpoints stated in the Technical Specification are correct, the associated tolerances are incorrect. The setpoint tolerances should be changed to +0, -3.3 VAC for over-voltage;

+2.9, -0 VAC for under-voltage, and +1,1, -0 Hz for under-frequency to be consistent with the equipment design specification.

JUSTIFICATION: FSAR subsection 8.3.1.1.5.2 presents the bases of the trip setpoints. An FSAR change will be submitted to address the effect of the line voltage drop to the scram discharge coils and to reflect the EPA protective circuitry setpoint tolerances.

The General Electric design specification provides trip setpoint tolerances as percentages of each specific setpoint for the over-voltage, under-voltage, and under-frequency protective instrumentation. These percentages were included in the present Technical Specifications as values (i.e., VAC or Hz) rather than percentages. The proposed change incorporates the correct tolerances in VAC or Hz based on the specification percentages.

SIGNIFICANT HAZARDS CONSIDERATION:

This proposed change is a correction to the setpoint tolerances for the RPS EPA protective circuitry to agree with the equipment design specification. This change does not involve a significant increase in the probability or consequences of an accident previously evaluated nor does it create the possibility of a new or different kind of an accident from any accident previously evaluated.

Since the revised tolerances are changed in the conservative direction, no reduction in the margin for safety is involved.

For these reasons, this proposed changes does not constitute a significant hazards consideration.

4 F29(3651)

16.(GGNb-479,4Bo 4%)

o ELECTRICAL POWER SYSTEMS r

REACTOR PROTECTION SYSTEM ELECTRIC POWER MONITORING LIMITING CONDITION FOR OPERATION i

Two RPS electric power monitoring assemblies for each inservice,RPS 3.8 4.3 IE set or alternate power supply shall be OPERABLE.

)

APPLICABILITY: At all times.

ACTION:

With one RPS electric power monitoring assembly for an inservice RPS MG set or alternate power supply inoperable, restore the inoperable power a.

monitoring system to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or remove the associated RP5 MG set or alternate power supply from service.

With both RPS electric power monitoring assembifes for an inservice RPS MG set or alternate power supply inoperable, restore at least one electric b.

power monitoring assembly to OPERABLE status within 30 minutes or remove the associated RPS MG set or alternate power supply from service.

SURVEILLANCE REQUIREMENTS The above specified RPS electric power monitoring assemblies shall be 4.8.4.3 determined OPERABLE:

At least once per six months by performance of a CHANNEL FUNCTIONAL a.

TEST, and At least once per 18 months by demonstrating the OPERASILITY of I

over-voltage, under-voltage and under-frequency protective b.

instrumentation by performance of a CHANNEL CALIBRATION including simulated automatic actuation of the protective relays, tripping logic and output circuit breakers and verifying the following setpoints:

S3 Over-voltage 132 + 0, -gVAC, 1.

l 2.

Under-voltage 117 +

, - 0 VAC, and 3.

Under-frequency 57 +

, - 0 Hz.

e 3/4 8-46 GRAND GULF-UNIT 1

O O

17. (GGNS - 418)

SUBJECT:

Technical Specification 4.9.6, page 3/4 9-8 DISCUSSION:

The primary fuel handling equipment at GGNS consists of a Refueling Platform and an Auxiliary Platform in the primary containment and a Fuel Handling Platform in the fuel handling area of the Auxiliary Building which are utilized in handling fuel assemblies and control rods. Technical Specification 3/4.9.6 provides specifications for only the Refueling Platform and does not address other equipment that may be used during fuel assembly and control rod handling operations.

The present GGNS Technical Specifications are based on the guidance presented in NUREG-0123, Standard Technical Specifications for General Electric Boiling Water Reactors, which is not directly applicable to the GGNS design.

Therefere, the present Technical Specification 3/4.9.6 should be deleted and a new section 3/4.9.6 added which conforms to the intent of the regulatory guidance and accurately reflects the GGNS design.

JUSTIFICATION: The CGNS design includes handling equipment in addition to the Refueling Platform that is used for handling fuel assemblies and control rods. The proposed change includes Technical Specifications for this additional equipment, i.e.,

for the Auxiliary Platform and the Fuel Handling Platform.

The proposed change maintains the applicable technical requirements of the existing specification for the Refueling Platform. However, the applicability has been expanded to include fuel handling operations within both the primary containment and the secondary containment. This assures the proper operation of handling equipment in each fuel handling area. The specification title has been changed to " Refueling Eqcipment" with subsections entitled " Refueling Platform",

" Auxiliary Platform", and " Fuel Handling Platform" in order to identify those requirements pertaining to the specific equipment in use.

The proposed change includes the Limiting Conditions for Operation and associated Action requirements as specified in NUREG-0123. However, only the main hoist on the Refueling Platform and on the Fuel Handling Platform may be used to handle irradiated fuel assemblies as stated in FSAR subsection 9.1.4.1.

The Surveillance Requirements included in the proposed change met the intent of the guidance specified in NUREG-0123.

However, changes have been incorporated in order to (1) conform to the GGNS design (2) eliminate confusion resulting from nomenclature differences and (3) organize the requirements according to the operation being performed and the equipment in use.

FSAR aubsection 9.1.4 provides a description of the fuel F30(3651) i

handling equipment and capabilities. FSAR subsection 7.6.1.1 provides a description of the refueling interlocks. However, due to several design changes which have been implemented, this section of the FSAR will be updated and revised at a later date. These changes included (1) eliminating the refueling interlocks associated with the Refueling Platform auxiliary hoists since these haists are not used for handling irradiated fuel assemblies, and (2) providing separate and redundant refueling interlock circuits for the Refueling Platform main hoist. Specific changes to the Surveillance Requirements are i

proposed as follows:

1.

The words " frame mounted and monorail" referring to the Refueling Platform hoists have been changed to

" auxiliary".

(The Fuel Handling Platform has one auxiliary hoist.)

FSAR section 9.1.4 and GGNS specifications refer to these hoists as auxiliary hoists.

2.

Technical Specification limits, with the exception of the slack cable cutoff, have been changed to allowable values rather than setpoints plus tolerance.

This is consistent with the philosophy used in other Technical Specifications providing limits. Where unacceptable consequences could result from exceeding only an upper or only a lower limit, allowable values are specified; however, where unacceptable consequences could result from exceeding either an upper or a lower limit, setpoints with tolerances are specified.

3.

Demonstration of the uptravel mechanical stop function on the Refueling Platform auxiliary hoists has been deleted.

The intent of this specification was to ensure _ irradiated fuel would be adequately shielded during handling operations; however, as described in FSAR subsection 9.1.4.1, the auxiliary hoists are not used in the handling of irradiated fuel assemblies. The proposed change specifically prohibits the use of auxiliary hoists for this purpose; therefore, this specification is not required.

4.

The word " load" in relation'to the main hoist on the Refueling Platform (and on the Fuel Handling Platform) has been changed to " total cable load".

As discussed in_FSAR subsection 9.1.4.2.7.1, the main hoists utilize a telescoping mast. As the' mast is extended the load is transferred from the cable to the platform; therefore, the load on the hoist varies depending on extension length. The design load setpoints are based on cable load and not the load on the grapple.

This change will' ensure there will be no confusion as to.

the definition of load.

F31(3651)

5.

The designation " loaded interlock" in the present GGNS Technical Specification 4.9.6.f has been corrected to identify it as the " grapple engaged loaded interlock".

The " grapple engaged loaded interlock" applies to both the Refueling Platform main hoist (allowable value of 535 pounds) and the Fuel Handling Platform main hoist (allowable value of 400 pounds).

The " grapple engaged loaded interlock" circuitry ensures a fuel assembly can not be raised unless the grapple hooks are full closed.

j

6.

The designation " redundant loaded interlock" in the present GGNS Technical Specification 4.9.6.g has been corrected to identify both the primary and its redundant interlock as the " primary and redundant fuel load interlocks" on the Refueling Platform main hoist. This change also clarifies that this interlock is different l

from the interlock referred to in the present GGNS Technicel Specification 4.9.6.f.

The Refueling Platform " primary and redundant fuel load interlocks" (allowable value of 600 pounds) provide input to the refueling interlock circuitry.

7.

The words " overload cutoff" in relation to the main hoist on the Refueling Platform (and on the Fuel Handling Platform) has been changed to "j am cutof f".

The installed annunciator system and equipment specifications refer to this function as the jam cutoff.

8.

The specification for the downtravel cutoff on the Refueling Platform main hoist has been changed to specifically identify the reference point as the bottom of the grapple and the limit has been changed to 3.5 0.5 inches.

The downtravel cutoff for the Refueling Platform main hoist is based upon the bottom of'the grapple since the purpose of'this' limit is to prevent striking the fuel bundle with the grapple. The-limit is specified as a setpoint with tolerance since exceeding the lower value -

could result in striking the top of the fuel bundle and exceeding the upper value could result in improper engagement. The value of 3.5 1 0.5 inches has been verified by reviewing "as built" drawings.

The Surveillance Requirements prepared for the Auxiliary Platform and the Fuel Handling Platform are based on the guidance of NUREG-0123, where applicable, and are

These requirements pertain to Refueling Platform operations in or over the-reactor pressure vessel only.

F32(3651)

structured in accordance with the changes discussed above.

Allowable values in relation to the Fuel Handling Platform main hoist are slightly lower than the corresponding values for the Refueling Platform based on the lower weight of the Fuel Handling Platform telescoping mast.

Additional specifications pertaining to the Fuel Handling Platform auxiliary hoist are necessary since this hoist is designed to be capable of handling new fuel assemblies.

These specifications assure the Fuel Handling Platform auxiliary hoist will not be utilized for handling irradiated fuel assemblies.

Finally, the Technical Specification basis has been corrected and revised to incorporate the additional specifications that are proposed.

SIGNIFICANT HAZARDS CONSIDERATION:

1 Changes which will result from this request fall into the 2

following categories:

s 1.

Administrative changes to establish consistency with plant terminology and nomenclature.

2.

Modifications to make the Technical Specifications consistent with actual plant refueling equipment design as described in FSAR Section 9.1.4 and 7.6.1.1.

3.

Changes to establish consistency throughout the Technical Specifications in the manner. single limit parameters are presented (i.e., allowable values as opposed to setpoints and tolerances).

4 Changes which incorporate additional plant features which should be covered by appropriate Technical Specifications.

These changes constitute additional restrictions, limitations and controls not' presently included in the Technical Specifications.

The changes do not involve a significant increase in the probability or consequences of an accident previously evaluated

~

nor do they create the possibility of a new or different kind of accident from any accident previously evaluated. No significant reduction in safety margin results from these changes. As they conform to NRC examples (i) and (ii) of amendments considered not likely to involve significant hazards considerations, they do not constitute a significant hazards consideration..

r s

V F33(3651) d

.a

N~

REFUELING OPERATION 3 i

3/4.9.6 REFUELING PLATFORM

\\MITINGCONDITIONFOROPERATION The re' fueling platform shall be OPERA 8LE and used for handling fuel its or control rods within the reactor pressure vessel.

3.9, ass During handling of fuel assemblies or control rods within the APPLICA8 ITY:

reactor pr sure vessel.

ACTION:

nts for refueling platform OPERABILITY not satisfied, suspend use of any 'noperable refueling platform equipment from With the require involving the hand 11 pressure vessel after lacing the load in a safe condition.

5URVEILLANCE REQUIREMENTS1 rane or hoist used for handling of control 4.9.6 Each refueling platform e reactor pressure vessel shall be demon-rods or fuel assemblies within r to the start of such operations with that strated OPERABLE within 7 days pr crane or hoist by:

e overload cutoff on the main hoist Demonstrating operation of a.

when the load exceeds 1200 t pounds.

ricad cutoff on the frame mounted Demonstrating operation of the o ceeds 500 1 50 pounds.

b.

and monorail hoists when the load Demonstrating operation of the uptrav 1 mechanical stop on the frame 1 brings the top of a fuel c.

mounted and monorail hoists when uptra assembly to 8 feet below the normal fuel torage pool water level.

hanical cutoff on the Demonstrating operation of the downtravel s 4 inches below d.

main hoist when grapple hook down travel renc fuel assembly handle, n the main hoist Demonstrating operali. ion of the slack cable cutoff when the load is less than 50110 pounds.

e.

sin hoist Demonstrating operation of the loaded interlock on the f.

when the load exceeds 485 1 50 pounds.

the Demonstrating operation of' the redundant loaded interlock o main hoist when the Toad exceeds 550 1 50 pounds, p.

i

)

3/4 9-8 GRAND GULF-UNIT 1

17.cacus-42.8)

REFUELING OPERATIONS 3/4.9.6 REFUELING EQUIPMENT REFUELING PLATFORM LIMITING CONDITION FOR OPERATION 3.9.6.1 The refueling platforn shall be OPERABLE and only the main hoist shall be used for handling fuel assemblies.

APPLICABILITY: During handling of fuel assemblics or control rods in the primary containment with the refueling platform.

ACTION:

With the requirements for refueling platform OPERABILITY not satisfied, suspend use of any inoperabic refueling platform equipment from operations involving the handling of fuel assemblics or control rods after placing the load in a safe condition.

SURVEILLANCE REQUIREMENTS 4.9.6.1 Each refueling platform hoist to be used for handling fuel assemblies or control rods shall be demonstrated OPERABLE within 7 days prior to the handling of fuel assemblies or control rods:

a.

In the containment fuel pool, reactor cavity or reactor pressure vessel by:

1.

Demonstrating operation of the slack cable cutoff on the main hoist when the total cable load is 50 10 pounds.

2.

Demonstrating operation of the grapple engaged loaded interlock on the main hoist before the total cable load exceeds 535 pounds.

3.

Demonstrating operation of the jam cutoff on the main hoist before the total cable load exceeds 1250 pounds.

4.

Demonstrating operation of the primary and redundant overload cutoff on the auxiliary hoists before the load exceeds 550 pounds.

s b

4 '

\\

s F34(3651) 3/4 9-8

17, (66Ns-418) b.

In or over the recctor pressure vessel by:

1.

Demonstrating operation of the downtravel cutoff on the main hoist when the bottom of the grapple is 3.5 1 0.5 inches below the top of the fuel assembly handles in the reactor core.

2.

Demonstrating operation of the primary and redundant fuel load interlocks on the main hoist before the totn1 cable load exceeds 600 pounds.

P 4

F35(3651) 3/4 9-8a 3

n. (cans-us) o REFUELING OPERATIONS 3/4.9.6 REFUELING EQUIPMENT AUXILIARY PLATFORM LIMITING CONDITION FOR OPERATION 3.6.9.2 The auxiliary platform shall be OPERABLE.

APPLICABILITY: During handling of control rods with the auxiliary platforn.

ACTION:

With the requirements for auxiliary platform OPERABILITY not satisfied, suspend use of the auxiliary platform after placing the load in a safe condition.

SURVEILLANCE REQUIREMENTS 4.9.6.2 The auxiliary platform hoist shall be demonstrated OPERABLE within 7 days prior to the handling of control rods by demonstrating operation of the overload cutoff before the load exceeds 550 pounds.

F36(3651) 3/4 9-8b

c L7. (C4NS -418)

REFUELING OPERATIONS 3/4.9.6 REFUELING EQUIPMENT FUEL HANDLING PLATFORM LIMITING CONDITION FOR OPERATION 3.9.6.3 The fuel handling platform shall be OPERABLE and only the main hoist shall be used to move irradiated fuel.

APPLICABILITY: During handling of fuel assemblies or control rods in the auxiliary building with the fuel handling platform.

ACTION:

With the requirements for fuel handling platform OPERABILITY not satisfied, suspend use of any inoperable fuel handling platform equipment from operations involving the handling of fuel assemblies or control rods after placing the load in a safe condition.

SURVEILLANCE REQUIREMENTS 4.9.6.3.1 Each fuel handling platform hoist to be used for handling fuel assemblies or control rods shall be demonstrated OPERABLE within 7 days prior to the handling of fuel assemblies or control rods by:

a.

Demonstrating operation on the slack cable cutoff on the main hoist when the total cable load is 50!10 pounds, b.

Demonstrating operation of the grapple engaged loaded interlock on the main hoist before the total cable load exceeds 400 pounds.

c.

Demonstrating operation of the jam cutoff on the main hoist before the total cable load exceeds 1150 pounds.

d.

Demonstrating operation of the primary and redundant overload cutoff on the auxiliary hoist before the load exceeds 550 pounds with the load override switch at the 500 pound position.

Demonstrating operation of the primary and redundant overload cutoff e.

on the auxiliary hoist before the load exceeds 1050 pounds with the load override switch at the 1000 pound position.

4.9.6.3.2 The auxiliary hoist load override switch shall be verified to be in the 500 pound position within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during hoist operation, except when engaged in new fuel movement in which case the switch may be in the 1000 pound position.

l F37(3651) 3/4 9-8c t

17. (GGNs-4te)

BASES 3/4.9.6 REFUELING EQUIPMENT The OPERABILITY requirements ensure (1) only the refueling platform main hoist will be used for handling fuel assemblies; (2) only the fuel handling platform main hoist will be used for handling irradiated fuel assemblies; (3) each platform hoist has sufficient load capacity for handling fuel assemblies and/or control rods as applicable; (4) the reactor internals are protected from excessive lif ting or impact force in the event they are inadvertently engaged or impacted during lifting or lowering operations; (5) the load sensing devices will provide the hoist fuel loaded interlock signals; and, (6) the probability of a fuel handling accident is minimized.

b i

i F38(3651) 3/4 9-8d

-, -,

ML20076G568

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