Text

Proposed Tech Spec Changes,Providing as-built Plant Consistency Re Min Water Level Required Above Reactor Pressure Vessel Flange During Operational Conditions 4 & 5
	ML20092B309
Person / Time
Site:	Grand Gulf
Issue date:	06/17/1984
From:	; MISSISSIPPI POWER & LIGHT CO.
To:	;
Shared Package
ML20092B304	List: AECM-84-0330, Application for Amend to License NPF-13,proposing Tech Spec Changes for as-built Plant Consistency Re Min Water Level Required Above Reactor Pressure Vessel Flange During Operational Conditions 4 & 5; ML20092B309;
References
	NUDOCS 8406200142
	Download: ML20092B309 (109)
	v • d • e

-.

, .o. t i

i L

i ATTACHMENT 1 i l-i PROPOSED CHANGES TO THE GRAND GULF NUCLEAR STATION TECHNICAL SPECIFICATIONS i

I NRC TECHNICAL REVIEW BRANCH: ACCIDENT EVALUATION l

i I

I t :

3 8406200142 840617 PDR ADOCK 05000416 l P PDR TS9sel

Attach:snt 1/ Accid:;nt Evoluntion AECM-84/0330 (6/17/84)

Page 2 Listing of item 1: umbers by Technical Specification Problem Sheet (TSPS) Number TSPS No. Item Nos.*

275 1.B.01

,r 4

V

/'

J

1 tem number format: 1.A.02 l Item number within category Catec-cy designator Attachment number TS9mm2

1 ,

l Attcchtsnt 1/Accidsnt Evolustion AECM-84/0330 (6/17/84) (

d Page 3 i

A. TYPOGRAPHICAL ERRORS, EDITORIAL CHANCES, AND CLARIFICATIONS f f

No technical specification changes in this category are included with this !

attachment. l t

l

[

i 4

t l l

> f i

..i r Y

i L f

f i

i I

i

.I f

4 ;

J i

P I

l.

I !

L

, r

) -

I L

i l 1 i

i ll- P

(

l i

4 TS9an3 i

4 O

e Attachment 1/Accidtnt Evaluation AECM-84/0330 (6/17/84)

Page 4 B. TECHNICAL SPECIFICATION /AS-BUILT PLANT CONSISTENCY The following change is proposed to render the technical specification consistent with the as-built plant. In all such cases, the as-built plant is consistent with the safety analyses and the licensing basis.

1 In that this proposed change is inherently consistent with the safety !

analyses and the licensing basis, it is concluded that the proposed change does not:

o Involve a significant increase in the probability or cons 4vences of an accident previously evaluated; or o Create the possibility of a new or different kind of accident from any accident previously evaluated; or o Involve a significant reduction in a margin of safety.

Therefore, the proposed change does not involve a significant hazards consideration.

A description of this change including justification for the change is provided below: ,

1. (TSPS 275), Reactor Vessel Water Level, Technical Specifications 3/4.8.1.2, 3/4.9.11.1, 3/4.9.11.2, 3/4.9.8, and Bases 3/4.9.11 The proposed change affects the specified minimum water level which must be maintai?2d above the reactor pressure vessel flange during OPERATIONAL CONDITIONS 4,5 and "*". Based on the as-built elevation of the reactor vessel flange, the actual depth of water above the flange is at least 22 feet 8 inches, instead of 23 feet. This change reflects the as-built elevation of the flange and does not impact overall pool volume or depth. This change has an insignificant impact on the iodine scrubbing function and heat sink capability of the large water volume. The proposed change alsc canservatively maintains at least 8.5 feet of water over the top of the active fuel for shielding purposes during irradiated fuel assembly transfer as specified in FSAR Section 12.3.2.2.2.c. This change will not adversely impact plant safety because it is consistent with the FSAR, the safety analysis and the as-built plant. (Page 3/4 8-9, 3/4 9-10, 3/4 9-16, 3/4 9-17, and Bases 3/4 9-2)

TS9mm4

AttachmInt 1/Accidsnt Evaluation AECM-84/0330 (6/17/84)

Page 5 C. ENHANCEMENTS THAT ARE CONSISTENT WITH THE SAFETY ANALYSES No technical specification changes in this category are included with this attachment.

TS9mm5

Attachmsnt 1/Accidsnt Evaluation !

AECM-84/0330 (6/17/84)

Page 6 D. REGULATORY REQUIREMENTS / REQUESTS / RECOMMENDATIONS No technical specification changes in this category are included with this i attachment.

b i

l I

l t

i t I i

I 4

i f

f I

I r

b i

I I

t TS9mm6 F

Attachmznt 1/Accidsnt Evaluation AECM-84/0330 (6/17/84)

Page 7 E. PROPOSED TECHNICAL SPECIFICATION CHARGES (AFFECTED PAGES ARE PROVIDED IN THE ORDER OF ASCENDING PAGE NUMBERS.)

i i

i-k 1

e t

I l

n

- TS9mm7

T"""" " "- U **i--9--y

- 9-,_,. ., . ,

~

ELECTRICAL POWER SYSTEMS A.C. SOURCES - SHUTDOWN . - .

LIMITING CONDITION FOR OPERATION 3.8.1.2 As a minimum, the following A.C. electrical power sources shall be OPERABLE:

a.

One Classcircuit between the 1E distribution offsiteand system, transmission network and the onsite

b. Diesel generator 11 and/or 12, and diesel generator 13 when the HPCS

~

system is required to be OPERABLE, with each diesel generator having:

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

2. A fuel storage system containing a minimum of:

a) 48,000 gallons of fuel each for diesel generators 11 and 12. i b) 39,000 gallons of fuel for diesel generator 13. ,

3. A fuel . transfer pump. ~

APPLICABILITY: OPERATIONAL CONDITIONS 4, 5 and *.

ACTION: '-

i.

With all offs.ite circuits inoperable and/or with diesel generators 11 and/or 12 of the above required A.C. ele..ctrical power sources inoperable, suspend CORE ALTERATIONS, handling of irradiated fuel in the primary or secondary containment, operations with a potential for draining the reactor vessel and crane operations over the spent fuel storage pool when fuel assemblies are stored therein. In addition, when in OPERATIONAL CONDITION 5 with the water level less g

p g g. p iate corrective action to restore the required power sources 22q ft:t 2wcWsf init ,

to OPERABLE status as soon as practical.

b. With diesel generator 13.of the above required A.C. electrical power l

t sources inoperable, restore the inoperable diesel generator 13 to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or declare the HPCS system inoperable and take the ACTION required by Specification 3.5.2 and 3.5.3.

c. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS

l 4.8.1.2 . At least the above required A.C. electrical power sources shall be ,

demonstrated OPERABLE per Surveillance Requirements 4.8.1.1.1, 4.8.1.1.2 and 4.8.1.1.3, except for the requirement of 4.8.1.1.2.a.5.

A When handling irradiated fuel in the primary or secondary containment.

GRAND GULF-UNIT 1 3/4 8-9 Amendment No. 9,

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

REFUELING OPERATIONS 3/4.9.8 WATER LEVEL - REACTOR VESSEL LIMITING CONDITION FOR OPERATION 33 FGir } tricHis 3.9.8 At least 22 f ;t of water shall be maintained over the top of )2 the reactor pressure vessel flange. 'i APPLICABILITY: During handling of fuel assemblies or control rods within the reactor pressure vessel while in OPERATIONAL CONDITION 5 when the fuel assemblies being handled are irradiated or the fuel assemblies seated within the reactor vessel are irradiated.

ACTION:

With the requirements of the above specification not satisfied, suspend all operations involving handling of fuel assemblies or control rods within the reactor pressure vessel after placing all fuel assemblies and control rods in a safe condition.

SURVEILLANCE REOUIREMENTS 4.9.8 The reactor vessel water level shall be determined to be at least its minimum required depth within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months prior to the start of and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months during handling of fuel assemblies or control rods within the reactor pressure vessel.

GRAND GULF-UNIT 1 3/4 9-10

. . - - . - --o -

O REFUELING OPERATIONS 3/4.9.11 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION HIGH WATER LEVEL LIMITING CONDITION FOR'0PERATION 3.9.11.1 At least one shutdown cooling mode train of the residual heat removal (RHR) system shall be OPERABLE and in operation

with at least:

a. One OPERABLE RHR pump, and

b. One OPERABLE RHR heat exchanger train.

APPLICABILITY: OPERATIONAL CONDITION 5, when irradiated fuel is in the reactor '

vessel ano tne water level is greater than or equal to 22 feet above the top l$2 of the reactor pressure vessel flange. **

32 Fstr- / /scuca ,

ACTION:

a. With no RHR shutdown cooling made train OPERABLE, within one hour and at !

least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter, demonstrate the operability of at least

, one alternate method capable of decay heat removal. Otherwise, suspend all operations involving an increase in the reactor decay heat load and establish SECONDARY CONTAINMENT INTEGRITY within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

b. With no RHR shutdown cooling mode train in operation, within one hour establish reactor coolant circulation by an alternate method and monitor reactor coolant temperature at least once per hour.

SURVEILLANCE REQUIREMENTS 4.9.11.1 At least one shutdown cooling mode train of the residual heat removal system or alternate method shall be verified to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

"The shutdown cooling pump may be removed from operation for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months per 8-hour period.

GRAND GULF-UNIT 1 3/4 9-16 /IM'#Do 'd

REFUELING OPERATIONS LOW WATER LEVEL LIMITING CONDITION FOR OPERATION 3.9.11.2 Two shutdown cooling mode trains of the residual heat removal (RHR) system shall be OPERABLE and at least one train shall be in operation,* with each train consisting of at least:

a. One OPERABLE RHR pump, and

b. One OPERABLE RHR heat exchanger train.

APPLICABILITY: OPERATIONAL CONDITION 5,'when irradiated fuel is in the reactor hn vessei ano tne water level is less than 22 fe t above the top of the reactor pressyre vessel flange. p;n psgr giacngs .

ACTION:

a. With less than the above required shutdown cooling made trains of the RHR system OPERABLE, within one hour and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter, demonstrate the operability of at least one alternate method capable of decay heat removal for each inoperable RHR shutdown cooling mode train.

b. With no RHR shutdown cooling mode train in operation, within one hour establish reactor coolant circulation by an alternate method and monitor reactor coolant temperature at least once per hour.

SURVEILLANCE REOUIREMENTS l

4.9.11.2 At least one shutdown cooling mode train of the residual heat removal system or alternate method shall be verified to be in operation and circulating '

i reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

l

\

I l

l =

i The shutdown cooling pump may be removed from operation for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months per 8-hour period. .

l GRAND GULF-UNIT 1 3/4 9-17 L

REFUELING OPERATIONS BASES f

3/4.9.7 CRANE TRAVEL - SPENT FUEL AND UPPER CONTAINMENT FUEL STORAGE POOLS i The restriction on movement of loads in excess of the nominal weight of a l fuel assembly over other fuel assemblies in the storage pools ensures that in i the event this' load is dropped 1) the activity release will be limited to that I contained in a single fuel assembly, and 2) any possible distortion of fuel in the storage racks will not result in a critical array. This assumption is consistent with the activity release assumed in the safety analyses.

3/4.9.8 and 3/4.9.9 WATER LEVEL - REACTOR VESSEL and WATER LEVEL -SPENT FUEL AND UPPER CONTAINMENT FUEL STORAGE P0OLS !

The restrictions on minimum water level ensure that sufficient water depth !

is available to remove 99% of the assumed 10% iodine gap activity released from l the rupture of an irradiated fuel assembly. This minimum water depth is i consistent with the assumptions of the accident analysis. ;

a 3/4.9.10 CONTROL ROD REMOVAL '

These specifications ensure that maintenance or repair of control rods or control rod drives will be performed under conditions that limit the probability of inadvertent criticality. The requirements for simultaneous removal of more ,

i than one control rod are more stringent since the SHUTDOWN MARGIN :;;scification !

provides for the core to remain subtritical with only one control rod fully withdrawn. '

3/4.9.11 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION The requirement that at least one residual heat removal loop be OPERABLE '

and in operation or that an alternate method capable of decay heat removal be' i demonstrated and that an alternate method of coolant mixing be in operation !

ensures that 1) sufficient cooling capacity is available to remove decay heat ;

and maintain the water in the reactor pressure vessel below 140*F as required :

i during REFUELING, and 2) sufficient coolant circulation would be available l through the reactor core to assure accurate temperature indication and to distribute and prevent stratification of the poison in the event it becomes necessary to actuate the standby liouid control system.

112 FES T f weMSL i The require have~two shutdown cooling loops OPERABLE when there % -

is less than 23 .,_:t of water above the reactor vessel ensures that a single failure of the operating loop wil.1 not result in a comp oss of resid-Q

.ual heat removal capability. With the reactor vessel head removed an 23 f::t h l

of water above the reactor vessel flange, a large heat sink is available for #L core cooling. Thus, in the event a failure of the operating RHR loop, adequate 1 i

time is provided to initiate alternate methods capable of decay heat removal

.or emergency procedures to cool the core. r 3/4.9.12 HORIZONTAL FUEL TRANSFER SYSTEM -

The purpose of the horizontal fuel transfer system specification is to '

control personnel access to those potentially high radiation areas immediated :

adjacent to the system and to assure safe operation of the system.

i '

GRAND GULF-UNIT 1 B 3/4 S-2 i 1

gggy

{_ _ . . _ , . ,_ _ _ _ , . _ _ . _ _ _ _ _ . . , _ . , _ _ _ _ , _ _ _ _ _ . _ _ _ _ _ _ , _ _ _ _ _ ______________ _ _ _.

ATTACHMENT 2 PROPOSED CHANGES TO THE GRAND GULF NUCLEAR STATION.

TECHNICAL SPECIFICATIONS NRC TECHNICAL REVIEW BRANCH: CONTAINMENT / SYSTEMS ZZ51aml

Attechment 2/Containmsnt Systems AECM-84/0330 (6/17/84)

Page 2 Listing of Item Numbers by Technical Specification Problem Sheet (TSPS) Number TSPS No. Item Nos.* TSPS No. Item Nos.*

012 2.C.01 172 2.A.08 019 2.A.04 233 2.C.07 020 2.B.01 235 2.D.01 031 2.D.03 240 2.B.03 067 2.D.02 266 2.A.09 069 2.D.04 269 2.A.01 127 2.C.02 276 2.A.02 144 2.A.07 283 2.A.06 >

164 2.C.04 294 2.C.03 167 2.C.05 306 2.B.02, 2.A.03 168 2.C.10 320 2.C.C3 169 2.C.01 379 2.C.09 170 2.C.06 818 2.B.04 171 2.A.05 i

[

t t

i

Item number format: 1.A.02 ltem number within category Category designator Attachment number ZZ51mm2 , - _ - - . , . - ,, . , . , _ - - . . . - - - -

Attachannt 2/Containannt Systems AECM-84/0330 (6/17/84)

Page 3 A. TYPOGRAPHICAL ERRORS, EDITORIAL CHANGES, AND CLARIFICATIONS These proposed changes correct obvious typographical errors, implement editorial changes such as correction of spelling errors, punctuation errors, and grammatical errors or provide clarification of the basic meaning or intent of the subject technical specifications.

MP&L has determined that the proposed changes do not:

o Involve a significant increase in the probability or consequences of an accident previously evaluated; or o Crea te the possibility of a new or different kind of accident from any accident previously evaluated; or o Involve a significant reduction in t margin of safety.

Therefore, the proposed changes do not involve a significant hazards consideration.

A description of these changes including necessary justification for the changes is provided below:

TYPOGRAPHICAL ERRORS Typographical errors are being corrected by this submittal as listed below. Correction of these typographical errors is purely an administrative change. (See attached revised technical specification pages for exact changes proposed.)

TSPS No. TS Page No.

1. 269 3/4 6-17

2. 276 3/4 6-9

3. 306 3/4 6-29 EDITORIAL CHANGES Proposed editorial changes to the technical specifications are discussed below:

4. (TSPS 019), Drywell Purge System, Technical Specification 3.6.7.3 The proposed change to delete the word " continued" adjacent to

" Surveillance Requirements", corrects an obvious editorial error.

(Page 3/4 6-58)

ZZ51mm3

,0

- Attachm:nt 2/Containmsnt Systems AECM-84/0330 (6/17/84)

Page 4

5. (TSPS 171), E61-F056 and E61-F057 Valve Description, Technical Specification Table 3.6.4-1 The Specification currently labels valves E61-F056 and E61-F057 as

" Purge Rad. Detector" isolation valves. The correct description for these valves is " Purge Filter Train Isolation." The proposed change is purely administrative to correct this labeling error. (Page 3/4 6-32)

6. (TSPS 283), Improper Building Name, Technical Specification 5.2.3 The subject specification states that secondary containment consists of the Reactor Building and the Enclosure Building. In the Grand Gulf design the building surrounding the primary containment is designated the Auxiliary Building rather than Reactor Building. Therefore, Specification 5.2.3 should be revised to state that secondary containment consists of the Auxiliary Building and the Enclosure Building. The proposed change is editorial in that it provides the plant specific designation for the building rather than general terminology. (Page 5-1)

CLARIFICATIONS Clarifications to the technical specifications to improve understanding l and readability are discussed below:

7. (TSPS 144), Primary Containment Integrity Leak Rate Testing Require- !

ments, Technical Specification 3/4.6.1.1 Surveillance Requirement 4.6.1.1.a. as presently written, states the equipment hatch seals are required to be leak-tested following the closure of any penetration that is subject to Type B testing, (Appendix J, 10CFR50) even if the equipment hatch had not been opened.

The words " equipment hatch" should be deleted, thereby clarifying that the intent of the surveillance is to require only those penetrations that have been opened to have their seals leak-tested after the penetration is closed. The proposed change is considered administrative in that its purpose is to clarify the intent of the Surveillance Requirement. (Page 3/4 6-1)

8. (TSPS 172), Drywell Bypass Leakage, Bases 3/4.6.2.2

- This proposed change expands the bases for drywell bypass leakage to:

I

a. Provide an explanation of the A//k term and an appropriate FSAR reference.

b. Provide an equivalent flow in scfm for the design drywell leakage rate (allowable drywell leakage capability) at 3 psid.

c. Provide an equivalent flow in sefm for the integrated drywell Icakage value at 3 psid.

ZZ51mm4

Attachmsnt 2/Containmant Systems AECM-84/0330 (6/17/84)

Page 5 These additions provide clarification and are consistent with the safety analysis. (Page B 3/4 6-3)

9. (TSPS 266), Reopening of Isolation Valves - MSIVs Excluded, Technical Specification 3/4.6.4 The proposed change to the footnote of the subject technical specification will exclude the Main Steam Isolation Valves (MSIVs) from the provision that permits isolation valves to be reopened under administrative controls, thus rendering this specification consistent with the requirements of Technical Specification 3/4.4.7. This proposed change involves no safety significance as it represents a clarification of the intent of the Technical Specification. (Page 3/4 6-27) l ZZ51mm5

Attacharnt 2/Containmsnt Systems AECM-84/0330 (6/17/84)

Page 6 B. TECHNICAL SPECIFICATION /AS-BUILT PLANT CONSISTENCY The following changes are proposed to render the technical specifications consistent with th,e as-built plant. In all such cases, the as-built plant is consistent with the safety analyses and the licensing basis.

In that these proposed changes are inherently consistent with the safety analyses and the licensing basis, it is concluded that the proposed changes do not:

o Involve a significant increase in the probability or consequences of an accident previously evaluated; or o Create the possibility of a new or different kind of accident from any accident previously evaluated; or o Involve a significant reduction in a margin of safety.

Therefore, the proposed changes do not involve a significant hazards consideration.

A description of these changes including justification for the changes is provided below:

1. (TSPS 020), Containment Isolation Valve Testing, Technical Specification Table 3.6.4-1 Revisions to Technical Speelfication Table 3.6.4-1 are proposed to require pneumatic testing of several valves that previously required hydrostatic testing. MP&L committed to adopt this more stringent method and to apply for these changes in a lettet from L. F. Dale to H. R. Denton dated September 12, 1983 (AECM-83/0540). The proposed changes are listed below, by page number:

a. Page 3/4 6 Delete footnote (c) notation for valves E12-F008-A, E12-F009-B, and E12-F023-A.

b. Page 3/4 6 Delete footnote (c) notation for valves E12-F042A-A, E12-F028A-A, E12-F037A-A, E12-F042B-B, E12-F028B-B, E12-F037B-B. Delete footnote (d) notation for valves E12-F021-B and E21-F012-A.

c. Page 3/4 6 Delete footnote (c) notation for valves ,

E12-F027A-A, E12-F027B, E12-F042C-B, E22-F004-C, and E21-F005-A.

Delete footnote (d) notation for valves E12-F064C-B and r E21-F011-A.

d. Page 3/4 6 Delete footnote (c) notation for valves E12-F308, E51-F066, E12-F344. E12-F044A, E12-F025A, E12-F107A, E12-F025B, E12-F044B, and E12-F107B.

l ZZ51mm6

Attechmant 2/Containmsnt Systems AECM-84/0330 (6/17/84)

Page 7

e. Page 3/4 6 Delete footnote (c) notation for valves E12-F234, E12-F041C-B, E22-F005, E22-F218, E22-F201, E21-F006, E21-F200, and E21-F207. Delete footnote (e) notation for valves E12-F280 E12-F281, E21-F217, and E21-F218.

f. Page 3/4 6 Change footnote (e) notation to (c) for valves E51-F251 and E51-F252.

g. Page 3/4 6 Delete footnote (d) notation for valves E12-F036 and E12-F005.

h. Page 3/4 6 Delete footnote (c) notation for valves E12-F002, E12-F342, E12-F061, E12-1056C, E12-F311, E12-F304, E22-F021, E21-F013, E21-F222, and E21-F221.

An evaluation has been performed which determined that eleven penetra-tions, which are currently tested hydrostatically, are located above the minimum suppression pool drawdown level and could therefore be exposed to containment atmosphere if a worst case single failure is considered. Standard Review Plan 6.2.6, " Containment Leakage Testing", states that hydrostatic testing of isolation valves is permissible if the line is not a potential containment atmosphere leak path; otherwise, pneumatic testing should be performed. Since the evaluation determined that it would be possible for the penetrations to become uncovered due to suppression pool drawdown, this change is proposed to require pneumatic testing of the isolation valves associated with these penetrations by deleting the hydrostatic testing footnotes. The valves associated with four of the af fected penetrations have not yet been pneumatically tested; these valves will be pneumatically tested on their next scheduled test date. A list of the penetrations and subject valves is presented below:

Penetration No. Valve No.

18 E12-F023 E12-F344 E12-F342 E12-F061 E51-F066 24 E12-F280 E12-F281 32 E21-F217 E21-F218 76 E12-F005 It should be noted that a design change is required in order to be able to perform pneumatic testing on four of the affected valves, namely valves E12-F280, E12-F281 E21-F217, and E21-F218. It should also be noted that although penetration 27 had previously been identified as requiring pneumatic testing of its isolation ZZ51mm7

Attachm nt 2/Contain xnt Systems AECM-84/0330 (6/17/84)

Page 8 valves subsequent investigation has determined that this pene-tration is below the minimum suppression pool drawdown level; therefore, exposure to containment atmosphere will not occur and pneumatic testing for its valves is not required.

In addition, a recent design change has added a valve and capped tee downstream of containment isolation valves E51-F251 and E51-F252 to enable direct hydrostatic testing of these valves rather than testing during system functional tests. The system configuration was such that direct observation of leakage from these valves could not be accomplished during system functional tests. Table 3.6.4-1, page 3/4 6-39, should be revised to reflect this design change by deleting footnote (e) and adding footnote (c).

The proposed changes, which require pneumatic testing of valves previously tested hydrostatically, represent additional limita-tions not presently included in the Technical Specifications.

Pneumatic testing provides additional assurance that the affected isolation valves will provide an adequate seal and is a more stringent surveillance requirement than the previously required hydrostatic testing, as documented in a letter from A. Schwencer (NRC) to J. P. McGaughy (MP&L) dated September 23, 1983.

The proposed changes represent r.o adverse impact to the safe operation of GGNS as they represent more stringent requirements and render the subject technical specification consistent with the as-built design. (Pages 3/4 6-29, 3/4 6-30, 3/4 6-34, 3/4 6-37, 3/4 6-38, 3/4 6-39, 3/4 6-40, and 3/4 6-42)

2. (TSPS 306), Containment and Drywell Isolation Valves, Technical Specification Tables 3.6.4-1 and 3.6.6.2-1 Changes are proposed to revise the maximum isolation times of 61 valves in Section 1 of Table 3.6.4-1. In a previous Technical Specification change submittal (letter No. AECM-83/0492 from L. F.

Dale (MP&L) to H. R. Denton (NRC), dated August 29, 1983), MP&L identified valves which have analytical closure time requirements.

Subsequently, MP&L has identified that the reactor water cleanup (RWCU) valves in Section 1 of the table should also be included in the list of valves that have analytical closure time requirements.

The revisions to the RWCU (G33) valves' closure times reflect

- General Electric's analyses of radiological dose and containment overpressurization resulting from a break in the RWCU system.

These analyses have shown that a maximum isolation time of 35 seconds for these valves will ensure an acceptable response for a RWCU pipe break either inside or outside of containment.

Additionally, it should be noted that although containment spray valves E12-F028A and B do have an analytical closure limit, the limit specified is based upon the maximum analytical opening time, not the maximum analytical closure time as stated in previous submittals.

ZZ51mm8

Attachmtnt 2/Containmsnt Systems AECM-84/0330 (6/17/84)

Page 9 In order to assure identification of all valves that have analyti-cal closure times, MP&L initiated a review to be conducted by both the NSSS vendor and the Architect / Engineer. This review identified all valve closure times that are based on analytical values. The valves in Table 3.6.4-1 that have been determined to have analytical closure time are listed below along with the analytical isolation times.

Maximum Isolation Times Valve Number (Seconds)

B21-F028A,B,C,D 5 B21-F022A,B,C,D 5 E12-F008 40 E12-F009 40 E12-F024A,B 90 M41-F011 4 M41-F012 4 M41-F034 4 M41-F035 4 M41-F015 4 M41-F013 4 M41-F016 4 M41-F017 4 E51-F063 20 E51-F064 20 G33-F028 35 G33-F034 35 G33-F039 35 G33-F040 35 G33-F001 35 G33-F004 35 G33-F252 35 G33-F053 35 G33-F054 35 G33-F250 35 G33-F251 35 G33-F253 35 The remaining revisions to the maximum isolation times are for valves that have no analytical closure times. For valves without analytical closure times, the margin of safety and the consequences of an accident will not be af fected as long as valve closure is assured. MP&L has determined that the maximum isolation time for these valves should be based on the results of ASME Section XI testing. This is consistent with ASME Section XI, which states that the owner shall set the maximum stroke time for valves.

An additional proposed change would move valves E12-F042A and B from Section 1.a to Section 2.a of the table. These valves are LPCI injection valves which open on the signals listed as group 5 isolation signals in Table 3.3.2-1. While these valves do receive a closure signal from a containment spray initiation signal, their inclusion in section 2.a of the table is appropriate since they do ZZ51mm9

Attachm:nt 2/Containmsnt Systems AECM-84/0330 (6/17/84)

Page 10 not receive an automatic containment isolation signal. Note that a discussion of the four sections of Table 3.6.4-1 has been added to Bases Section 3/4.6.4 as resolution to TSPS 170.

The final proposed change would revise an incorrect penetration number and add information designating the divisional power supply associated with the valves in Table 3.6.4-1 and Table 3.6.6.2-1.

The designators A, B, and C, and (A), (B), and (C), correspond to electrical divisions 1, 2, and 3 for motor-operated valves and the solenoids for air-operated valves, respectively. No designators were added to the Main Steam 1 solation Valves; these valves receive power from both Division 1 and 2. These changes do not affect any specification requirement but are enhancements that will provide additional information.

The proposed changes do not affect safety, even though the closure times are increased for most of the affected valves, because all of the proposed times are less than or equal to the analytical closing values, if applicable. A time history of the closure times for these valves (observed during surveillance testing) will be maintained and reviewed to detect degradation of valve performance, thereby decreasing the probability of an unexpected valve failure.

(Pages 3/4 6-29, 3/4 6-30, 3/4 6-31, 3/4 6-32, 3/4 6-33, 3/4 6-34, 3/4 6-35, 3/4 6-36, 3/4 6-37, 3/4 6-38, 3/4 6-44, 3/4 6-48, 3/4 6-49, 3/4 6-50, 3/4 6-51 and 3/4 6-52)

3. (TSPS 240), Containment and Drywell Hydrogen Recombiners, Technical Specification 3/4.6.7.1 The proposed changes to the subject Specification will delete all references to drywell hydrogen recombiner systems because the GGNS plant design incorporates hydrogen recombiners only in the contain-ment, not in the drywell. The proposed changes involve no safety significance as they represent a clarification of plant design and do not affect the technical content of the specifications. (Page 3/4 6-56)

4. (TSPS 818), Secondary Containment Isolation, Technical Specifica-tions 1.37, 3/4.6.6.1, and Bases 3/4.6.6 The proposed changes to Definition 1.37, surveillance Requirement 4.6.6.1, and Bases 3/4.6.6 add rupture discs and blind flanges, as appropriate, to the list of equipment necessary for SECONDARY CONTAINMENT INTEGRITY. These revisions are required because the GGNS design utilizes these components as one method of ensuring secondary containment integrity. The proposed changes involve no safety significance as they represent clarifications that more accurately reflect plant design and the intent of the existing Specifications. (Pages 1-7, 3/4 6-46, and B 3/4 6-6)

ZZ51mm10

Attachmint 2/Containmsnt Systcm3 AECM-84/0330 (6/17/84)

Page 11 C. ENHANCEMENTS THAT ARE CONSISTENT WITH THE SAFETY ANALYSES The following proposed changes are enhancements which are consistent with the safety analyses and the licensing basis and which provide clarifica-tion, render areas consistent with the philosophy and intent of the technical specifications, or provide additional plant operational margin.

Since these proposed changes are included in the current licensing bases and are bounded by existing safety analyses, the proposed changes do not:

o Involve a significant increase in the probability or consequences of an accident previously evaluated; or o Create the possibility of a new or different kind of accident from any accident previously evaluated; or o Involve a Lignificant reduction in a margin of safety.

Therefore, the proposed changes do not involve a significant hazards consideration.

A description of these changes including justification for the changes is provided below:

1. (TSPS 012 and 169), Containment Spray and Suppression Pool Cooling, Technical Specifications 3.6.3.2, 3.6.3.3 and Bases 3/4.6.3 The following technical specification changes are proposed:

a) Change "SSW heat exchanger" to "RHR heat exchanger," in Technical Specification 3.6.3.2.b and 3.6.3.3.b.

b) Add the containment spray spargers to Technical Specification 3.6.3.2.b, i

1 i

c) Revise " ACTION b" of Technical Specification 3.6.3.2 to be consistent with " ACTION b" of Technical Specification 3.6.3.3, d) Change the time in " ACTION a" of Technical Specification 3.6.3.3 i

from 7 days to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , e) Add new Surveillance Requirement 4.6.3.2.d, and

. f) Revise Bases 3/4.6.3 to reflect the cided Surveillance Requirement, f The change from SSW to RHR heat exchanger corrects an error in I terminology and is made to reflect correct nomenclature. The addition of the containment spray spargers to Specification 3.6.3.2.b reflects the system design and is proposed to ensure system OPERABILITY. The l changes to the ACTION statements are enhancements to achieve consist-ency between the containment spray and suppression pool cooling modes j

of RHR operation. The addition of the requirement to perform an air l

ZZ51mm11

- - - . - . - _ = . - - . - -. . -. . -- _ - - _ . .

Attcchmsnt 2/Containmtnt Systems i AECM-84/0330 (6/17/84) '

Page 12 or smoke flow test to Surveillance Requirement 4.6.3.2 and the revision to Bases 3/4.6.3 constitutes an additional requirement not presently included in the technical specifications. The design of the containment spray system is such that nozzle obstruction should not occur unless caused by maintenance activities; therefore, the surveillance i frequency should not be time dependent but should instead be coordinated with the completion of applicable maintenance activi-

' ties. The containment spray nozzles were initially air-tested during ;

the preoperational test phase and no maintenance has been performed on l' the system since that time which could cause nozzle blockage. (Pages 3/4 6-24, 3/4 6-25 and B 3/4 6-4)

2. (TSPS 127), Drywell to Containment Differential Pressure, Technical ,

Specification Bases 3/4.6.2.5 l L

The change in the minimum differential pressure from -0.1 to -0.26 in Specification 3.6.2.5 and the corresponding correction to Bases 3/4.6.2.5 were previously submitted as item 22 of AECM-83/0565.

The 0.26 psid value resulted from an analysis performed to resolve one of the Humphrey concerns, (overflow of the weir wall). An additional change to the Bases, which states that the 2.0 psid limit for positive drywell to containment pressure will "not allow 4

clearing of the top vent" is proposed to reflect the results of the drywell/ containment analyses. The proposed change is considered an i enhancement that will render the specifications consistent with safety analyses. (Page B 3/4 6-3)

3. (TSPS 294), Containment Leakage Rates. Technical Specification i 3/4.6.1.2 The proposed changes to the subject technical specification clarify [

' the leakage requirements for containment isolation valves and i i penetrations and their applicable testing methods so as to ensure ;

compliance with 10 CFR 50. As presently written, the specification requires only ECCS and RCIC containment isolation valves (which are j

(

in hydrostatically tested lines that penetrate the primary containment) to be included in the group of valves that must have a combined leakage rate of less than or aqual to 1 gpm times the total number of valves in the group. Other containment isolation .

valves should also be included in the group of valves that are subject to containment leak rate requirements; therefore the terms "ECCS and RCIC" should be deleted from the appropriate sections so as to broaden the scope of this specification. In addition, an -

editorial change to Specification 3.6.1.2.b and Action b removed a 4

portion of the text and placed it in a footnote, thus rendering

' these sentences more readable. The proposed changes involve no safety significance as they represent a clarification of the technical specifications addressing containment leakage. (Pages 3/4 6-2, 3/4 i 6-3, 3/4 6-4) i i

ZZ51mm12

Atttchnint 2/Conteinmtnt Syctems AECM-84/0330 (6/17/84)

Page 13

4. (TSPS 164), Primary Containment Integrity Requirements, Definition 1.30 and Technical Specification 3/4.6.1.1 The proposed changes to Definition 1.30 and Surveillance Requirement 4.6.1.1 will render the definition of containment air lock and suppression pool OPERABILITY consistent with the intent of Specifications 3.6.1.3 and 3.6.3.1. Specifications 3.6.1.3 and 3.6.3.1 contain exceptions to the basic OPERABILITY requirements that permit continued operation under the conditions specified in the ACTIONS.

Since PRIMARY CONTAINMENT INTEGRITY is maintained if the conditions of these ACTIONS are satisfied, the term OPERABLE in Definition 1.30 and Surveillance Requirement 4.6.1.1 should be deleted and replaced with a statement that requires the containment air locks and suppression pool to be in compliance with requirements of Technical Specifications 3.6.1.3 and 3.6.3.1, respectively.

These changes are considered enhancements in that they clarify the intent of those specifications related to primary containment integrity. (Pages 1-6 and 3/4 6-1)

5. (TSPS 167), Drywell Integrity, Definition 1.10 and Technical Specification 3/4.6.2.1 The proposed changes to Definition 1.10 and Surveillance Requirement 4.6.2.1 will render the definition of drywell air lock and suppression pool OPERABILITY consistent with the intent of Specifications 3.6.2.3 and 3.6.3.1. Specifications 3.6.2.3 and 3.6.3.1 contain exceptions to the basic OPERABILITY requirements that permit continued operation under the conditions specified in the ACTIONS. Since DRYWELL INTEGRITY is maintained if the conditions of these ACTIONS are satisfied, the term OPERABLE in Definition 1.10 and Surveillance Requirement 4.6.2.1 should be deleted and replaced with a statement that requires the drywell air locks and suppression pool to be in compliance with the requirements of Technical Specifications 3.6.2.3 and 3.6.3.1, respectively. These changes are considered enhancements in that they clarify the intent of those specifications related to DRYWELL INTEGRITY. (Pages 1-2 and 3/4 6-13)

6. (TSPS 170), Additional Bases Information Concerning Table 3.6.4-1, Technical Specification Bases 3/4.6.4 Technical Specification 3.6.4 addresses the containment and drywell isolation valves which are listed by section in Table 3.6.4-1. The proposed technical specification change to add a description of the different valve categories listed in Table 3.6.4-1 is made to the Bases to clarify the arrangement of the table. This proposed clarification is an enhancement and involves no safety significance because it is consistent with the information presently contained in Table 3.6.4-1. (Page B 3/4 6-5)

ZZ51mm13

Attachm nt 2/Containm:nt Systems AECM-84/0330 (6/17/84)

Page 14 7 (TSPS 233), RHR Flows for Containment Spray Mode, Technical Specification Bases 3/4.6.3 The proposed change to Bases Section 3/4.6.3 is to add a statement confirming that Surveillance Requirements 4.5.1.b and 4.6.3.2.b provide adequate assurance that the Containment Spray System will be OPERABLE when required. Suf ficient flow through the RHR heat exchangers will ensure sufficient flow to the containment spray nozzles, since the minimum acceptable flow and total developed head values, stated in the surveillance requirement, account for inherent system losses. This change is an enhancement that clarifies the Bases for the Containment Spray System specification. (Page B 3/4 6-4)

8. (TSPS 320), Depressurization Systems Bases, Technical Specification Bases 3/4.6.3 This change clarifies that the pressure used in the reactor blowdown analysis was 1060 psia instead of 1089 psig. The 1060 psia is consistent with the numerical value listed in FSAR Table 15.0-3 " Input Parameters and Initial Conditions for Transients Used in ODYN Code."

This change is an enhancement which is consistent with the safety analysis that is referenced above. (Page B 3/4 6-4)

9. (TSPS 379), Air Lock Seal Decay Test, Technical Specification Bases 3/4.6.1.3 and 3/4.6.2.3 This proposed change to the Bases provides clarification that a shorter time period for the leak-detection test may be used if the test is conducted in accordance with ANSI N45.4-1972.

The methodology used for assuring that the leakage rate can be accurately determined with a test period less than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months is described below. This methodology was developed and used in previous tests. Further refinements may be made to the methodology as experience dictates.

Three test instruments are used, a pressure gauge, a barometer, and a the rmometer. Considering accuracy, readability and repeatability, for the pressure gauge there is an uncertainty of !0.07 psi; for the barometer, the uncertainty is !0.02 psi; and for temperature varia-tions, the uncertainty is !0.005 psi. The algebraic sum of the uncertainties is 0.095 psi. A conservative !0.1 psi uncertainty is used.

Although the pressure decay is exponential, linear interpolation will be used, which is more conservative; e.g. , for a 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months test, the allowable pressure loss for a 30 day exponential decay is 0.588 psi; for a linear interpolation, the pressure loss cannot exceed 0.500 psi.

r l

ZZ51mm14

AttachmInt 2/Containrant Systems AECM-84/0330 (6/17/84)

Page 15 Using the linear interpolation of the allowable 30 day pressure loss, a plot of the actual pressure decay minus 0.1 psi is plotted on the same chart. When the plot of the actual pressure decay crosses the plot of the 30 day linear interpolation, the test will be considered to have been successfully completed.

This method is conservative and even with zero pressure loss, a minimum of 2.4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is required before it may be concluded that the results are acceptable based on the 10.1 psi uncertainty that is used.

The proposed change to the Bases of Technical Specification 3/4.6.1.3 and 3/4.6.2.3 provides an operational enhanceuent that is consistent with the philosophy and intent of the technical specifications.

Approval of this change by the NRC will constitute prior acceptance and an acknowledgement that leakage rates can be accurately determined during test periods that are less than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . (Page B 3/4 6-1 and B 3/4 6-3)

10. (TSPS 168), Emergency Core Cooling Systems - Suppression Pool, Depressurization Systems - Suppression Pool, and Depressurization Systems Bases, Technical Specifications 3/4.5.3, 3/4.6.3 and Bases 3/4.6.3 The proposed changes to the subject technical specifications are as follows:

a. Revise the suppression pool low and high water levels (depths) to 18' 4-1/12" and 18' 9-3/4", respectively, to make them consistent with FSAR TABLE 6.2-50, " SUPPRESSION POOL GE0 METRY -

251 PLANT."

b. Delete the reference to OPERATIONAL CONDITION 1 or 2 in LCO 3.6.3.1.b, ACTION 3.6.3.1.b, and Surveillance Requirement 4.6.3.1.b.

c. Expand Bases 3/4.6.3, "DEPRESSURIZATION SYSTEMS," to provide bases for:

1. Suppression pool volumes

2. Suppression pool levels (depths)

3. Suppression pool temperatures

. The changes to the suppression pool water levels are considered enhancements made to be consistent with the volumes used for the safety analyses. The deletions of the reference to OPERATIONAL CONDITION 1 or 2 are considered enhancements which make the Limiting Conditions for Operation, ACTION statements, and Surveillance

Requirements consistent with the Applicability Statement. The changes to Bases 3/4.6.3 are enhancements which provide substantial clarification and are consistent with the safety analyses. (Pages 3/4 5-8, 3/4 5-9, 3/4 6-20, 3/4 6-21. B 3/4 6-4)

ZZ51mm15

Attachm:nt 2/ Containment Systems AECM-84/0330 (6/17/84)

Page 16 D. REGULATORY REQUIREMENTS / REQUESTS / RECOMMENDATIONS The following changes are proposed to render the technical specifications consistent with recent changes in NRC policy and the Code of Federal Regulations, as we'll as to implement changes or enhancements recently requested or recommended by NRC reviewers.

These proposed changes are required to render the technical specifications consistent with recent NRC guidance, and it has been concluded based on a review of each item that the proposed changes do not; o Involve a significant increase in the probability or consequences of an accident previously evaluated; or o Create the possibility of a new or different kind of accident from any accident previously evaluated; or o Involve a significant reduction in a margin of safety.

Therefore, the proposed changes do not involve a significant hazards consideration.

A description of these changes including justification for the changes is provided below:

1. (TSPS 235), Containment Air Lock, Technical Specification 4.6.1.3 Technical Specification Surveillance Requirement 4.6.1.3.a should be revised to add a footnote to both "72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months " time limits, which will specify that the provisions of Technical Specification 4.0.2 are not applicable. This change clarifies that no extension of the 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months time limits to demonstrate OPERABILITY of the containment air lock (s) is allowed and ensures that the Technical Specifications are in compliance with 10CFR50, Appendix J. (Page 3/4 6-6)

2. (TSPS 067), Type A Test Accuracy Determination. Technical Specification 3/4.6.1.2 Surveillance Requirement 4.6.1.2.c.1 should be revised so as to l provide clarification concerning the method to be used for verifi-l cation of the accuracy of Type A containment leakage rate testing (Appendix J, 10CFR50). In addition, Surveillance Requirement l

l 4.6.1.2.c.3 should be revised to indicate that the required I quar.tity of gas injected into, or bled from, the containment during the supplemental test must be between 0.75 L and 1.25 L . These changes are considered to be enhancements and are in compliance with 10 CFR 50 Appendix J. (Page 3/4 6-3) l i

l i

l l

l ZZ51mm16

Attachment 2/Containmsnt Systems AECM-84/0330 (6/17/84)

Page 17

3. (TSPS 031), Drywell Air Lock, Technical Specification 3/4.6.2.3 Surveillance Requirement 4.6.2.3.a is changed to require that air lock OPERABILITY be demonstrated within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , instead of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , after closing. Surveillance Requirement 4.6.2.3.b is changed to add a requirement that airlock OPERABILITY be demonstrated following any maintenance that could affect the sealirg capability. The change to Surveillance Requirement 4.6.2.3.a makes the Technical Specification consistent with 10 CFR 50 Appendix J, paragraph D.2.iii and industry standards. The change to Surveillance Requirement 4.6.2.3.b represents an additional requirement not included in 10CFR50 Appendix J that will ensure that the airlock is OPERABLE prior to establishing DRYWELL INTCCRITY and following maintenance that could affect air lock sealing capability. (Page 3/4 6-16)

4. (TSPS 069), Containment and Drywell Hydrogen Ignition System. Technical r Specification 3/4.6.7.2 This proposed change replaces present Specification 3/4.6.7.2 with an expanded Specification that adds requirements which ensure OPERABILITY of the H,, ignition system. Changes to the Limiting Condition for Operation (LCO) require at least two igniter assemblies in each enclosed area in the Containment to be OPERABLE as well as all igniter assemblies adjacent to any inoperable igniter assembly in each open area in the Containment and Drywell. Proposed changes to the ACTION Statements are provided to coincide with changes to the LCO's. New ACTION "a" requires that with less than two igniter assemblies OPERABLE in any enclosed area in the Containment, at least two must be restored 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 . New ACTION "b" requires that with any adjacent igniter assemblies within open areas of the Containment or Drywell inoperable, the igniter assemblies must be restored so that all igniter assemblies adjacent to an inoperable igniter assembly are OPERABLE 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 . The current ACTION Statement is retained as new ACTION Statement "c." Proposed changes to Surveillance Requirement 4.6.7.2 are made to demonstrate OPERABILITY of the H igniters required OPERABLE by the LCO. New 2

Surveillance 4.6.7.2.a.1 and 4.6.7.2.a.2 require energizing the supply breakers at least once per 92 days and verifying a visible glow from each normally accessible igniter assembly in the Containment and verifying that each circuit of each Containment and Drywell H, igniter subsystem is conducting sufficient current to energize the minimum number of igniter assemblies required as specified on new Table 4.6.7.2-1. New Surveillance Requirement 4.6.7.2.b requires, at every COLD SHUTDOWN but no more frequently than once per 92 days, that each normally inaccessible igniter assembly is verified OPERABLE by energizing the supply breakers and verifying a visible glow from the glow plugs. New Table 3.6.7.2-1 lists the H2igniters by electrical division and by circuits within l

j each division. New Table 3.6.7.2-2 lists the H2 igniters by t

l ZZ51mm17

Attacharnt 2/Containm2nt Systs.ms AECM-84/0330 (6/17/84)

Page 18 electrical division / circuit, elevation, azimuth, and distance from the center line of the reactor. New Table 3.6.7.2-2 also lists those igniters in normally accessible, inaccessible, open or enclosed areas within the containment and/or drywell. New Table 4.6.7.2-1 lists the minimum required igniters per circuit.

The proposed changes to the9H igniter specification follow NRC guidance and ensure OPERABILIN of the system. The proposed changes to the LCO, ACTION and Surveillance Requirements address igniter assemblies in both enclosed and open areas to ensure that all required areas have OPERABLE igniters. These changes assureyH igniter system OPERABILITY and address system design by requiring tne normally inaccessible assemblies in the drywell and containmer.t to have a visible glow verification during COLD SHUTDOWN. These igniters are inaccessible due to high levels of radiation and/or high temperatures in these areas during plant operation; however, OPERABILITY is verified for the minimum number required per circuit by electrical current checks at least once per 92 days. The new tables enhance the Specification by providing tabulations for igniter location, electricaldivisionandcircuit,andminimumnumberrequiredfo$each circuit. The proposed changes provide an increase in safety through more stringent requirements than those currently in the technical specifications and are consistent with the licensing basis. (Pages 3/4 6-57, 3/4 6-57a, 3/4 6-57b, 3/4 6-57c 3/4 6-57d, 3/4 6-57e, 3/4' 6-57f) f 4

ZZ51mm18 - - - - -_ - - _ . _, _ _. - __

Attcchmsnt 2/Containunt Systems AECM-84/0330 (6/17/84)

Page 19 E. PROPOSED TECHNICAL SPECIFICATION CHANGES (AFFECTED PAGES ARE PROVIDED IN THE ORDER OF ASCENDING PAGE NUMBERS.)

I i

E I

I ZZ51mm19

l DEFINTTIONS CORE ALTERATION 1.7 CORE ALTERATION shall be the addition, removal, relocation or movement of fuel, sources, incere instruments or reactivity controls within the reactor pressure vessel with the vessel head removed and fusi in the vs::el.

Suspension of CORE ALTERATIONS shall not preclude completion of t.1e movement of a component to a safe conservative position. l CRITICAL POWER RATIO i

1. 8 The CRITICAL POWER RATIO (CPR) shall be the ratio of that power in the ;

assembly which is calculated by application of the GEXL correlation to cause !

some point in the assembly to experience boiling transition, divided by the '

actual assembly operating power.

i DOSE EOUIVALENT I-131 3 i

1.9 DOSE EQUIVALENT I-131 shall be that concentration of I-131, microcuries t per gram, which alone would produce the same thyroid dose as the quantity and !

isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present. '

The thyroid dose conversion factors used for this calculation shall be those t listed in Table III of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites." ,

t DRWELL INTEGRITY l 1.10 DRWELL INTEGRITY shall exist when: l

a. All drywell penetrations required to be closed during accident $

conditions re either:

1. Capable of being closed by an OPERABLE drywell automatic -

isolation system, or

2. Closed by at least one manual valve, blind flange, or deactivated automatic valve secured in its closed position, !

except as provided in Table 3.6.4-1 of Specification 3.6.4. i

b. The drywell equipment hatch is closed and sealed

c. The drywell airlock is4 in com 0rOA::L.pliece.

pursua t te WM b ukhNds3.6.'2.3.

SpecTfication d $l

d. The drywell leakage rates are within the limits of Specification 3.6.2.2.

. 16 compli,1cc. we%k /eparer' sed

d y ,

e. Thesuppressionpoolis40PERASLE purwar.t t: Specification 3.6.3.1. -

f. The sealing mechanism associated with each drywell penetration; e.g., welds, bellows or 0-rings, is OPERABLE.

P GRAND GULF-UNIT 1 1-2 Andet M*- '

L

. .~ -. _ . _ _ . - . _ _ _ _ _ _ _ _ _ . . _ _ _ _ _ _ . _ _ _ _ _ _ . . ___ _ _ . _ -

DEFINITIONS

'b REACTOR PROTECTION SYSTEM RESPONSE TIME 1.34 REACTOR PROTECTkON SYSTEM RESPONSE TIMb shall be the time inte when the monitored parameter exceeds its trip set; int at the channel sensor until de energization of the scram pilot valve solencies. The response time may be measured by any series of secuen-ial, overlapping or total steps sucn that the entire response time is measured.

REPORTABLE OCCURRENCE I 1.25 A REPORTABLE OCCURRENCE shall be any of those cencitions specified in Specifications 6.9.1.12 and 6.9.1.13.

ROD DENSITY

~1.36 RCD DENSITY shall be the numeer of centrol red netches inserted as a frac-ion of the total number of con rol rod notc. as. All reds fully inser:cc is eq ivalent to 100% ROD DENSITY.

i SECONDARY CONTAINMENT INTEGRITY 1.37 SECONDARY CONTAINMENT INTEGRITY shall exist when: i

a. All Auxiliary Building and Enclosure Building penetrations

__...__ _ required to be closePth:rtif acetcent csnditions are 4Tther: ---~ ~

1. Cacable of being closed by an OPERABLE secondary containment autc=atic isolation system, or r roptere dise ,,,

2. Closed by at least one manual valve, blind flange,Aor i :

deactivated automatic valve or damper, as acplicable, securec !

in its closed position, except as proviced in Table 3.6.5.2-1 of Specification 3.5.6.2.

b. All Auxiliary Building and Enclosure Building equipment hatches and blowout panels are closac anc sealed.

c. The stancby gas treatment system is OPERABLE pursuant to Specification 3.6.6.3.

d. The door in each access to the Auxiliary Building and Enclosure Building is closed, except for normal entry and exit.

e. The sealing mechanism asscciated with each Auxiliary Building anc Enclosure Building penetration, e.g., welcs, bellows or 0-rings, is OPERABLE.

k_.

.,e<< wo. l GRAND GULF-UNIT 1 1-7

DEFfNTTIONS PRIMARY CONTAINMENT INTEGRITY 1.30 PRIMARY CONTAINMENT INTEGRITY shall exist when:

a. All containment penetrations required to be closed during accident ;

conditions are either:

1. Capable of being closed by an OPERABLE containment aut:matic isolation system, or

2. Closed by at least one manual valve, blind flange, or deactivated automatic valve secured in its closed position, except as provided in Table 3.6.4-1 of Specification, 3.6.4.

b. The containment equipment hatch is closed and sealed.

c. Each containment air lock is.0rERAELE pur: cent tc Specification $

3.6.1.3. 88 (* N" WM

N M *"*3

1' -

d. The containment leakage rates are within the limits of Specification i 3.6.1.2. -

e. The suppression pool is OPERASLE pur;ucnt t: Specification 3.6.3.1. 3 w co wt oew won su pg-os.- m~u -+ .

f. The sealing mechanism associated with each primary containment ,

penetration; e.g., welds, bellows or 0-rings, is OPERABLE. -

PROCESS CONTROL PROGRAM (PCP) 1.31 The PROCESS CONTROL PROGRAM shall contain the sampling, analysis, and formulation determination by which SOLIDIFICATION of radioactive wastes from l liquid systems is assured. i PURGE - PURGING i 1.32 PURGE or PURGING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required !

to purify the confinement. T RATED THERMAL POWER 1.33 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 3833 MWT.

GRANO GULF-UNIT 1 1-6 -

l l . _

. _ - - - . __ _ M"!'!_ _ ' * * -. -

EMERGENCY CORE COOLING SYSTE'45 3/4.5.3 SUPPRES5 ION PCOL LIMITING C0tIDITION FOR OPERATICN 3.5.3 The suppression pool shall be OPERABLE:

a. In OPERATIONAL C0tlDITION 1, 2 or 3 with a contained water volume of at least 135,291 ft3 , equivalent to a level of 18'4- W "- 16 8 t/l2

b. In OPERATIONAL C0tIDITION 4 or 5" with a contained water volume of at least 93,600 ft3 , equivalent to a level of 12'8", excep- tnat the suppression pool level may be less than the limit or may be drained provided that:

1. No operations are performed that have a potential for draining the reactor vessel,

2. The reactor mode switch is locked in the Shutdcwn or Refuel position,

3. The condensate storage tank contains at least 170,000 available gallons of water, equivalent to a level of 18', and

4. The HPCS system is OPERABLE per Specification 3.5.2 with an OPERABLE flow path cacable of taking suction from the condensate storage tank and transferring the water thrpugh the spray sparger to the reactor vessel.

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

ACTION:

a. In GPERATIONAL CONDITION 1, 2 or 3 with the suppression pool water level less than the above limit, restore the water level to within the limit within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT SHUT 00WN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTD0'4N within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b. In OPERATIONAL CONDITION 4 or 5* with the suppression pool water level less than the above limit or drained and the above required

. conditions not satisfied, suspend CORE ALTERATIONS and all opera-tions that have a potential for draining the reactor vessel and lock the reactor mode switch in the Shutdown position. Establish SECONDARY CONTAINMENT INTEGRITY within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

See Specification 3.6.3.1 for pressure suppression requirements.

The suppression pool is not required to be OPERABLE provided that the reactor vessel head is removed, the cavity is flooded or being flooded from the suppression pool, the upper containment' fuel pool gates are removed when the cavity is flooded, and the water level is maintained within the limits of Specification 3.9.8 and 3.9.9.

GRAND GULF-UNIT 1 3/4 5-8 Amendment No. _

EMERGENCY CORE COOLING SYSTEMS !

LIMITING CONDITION FOR OPERATION (Continued) ,

ACTION: (Continued)

c. With one suppression pool water level instrumentation division '

inoperable, restore the inoperable division to OPERABLE status !

within 7 days or verify the suppression pool water level to be -

greater than or equal to 18'4-3/4" or 12'8", as applicable, at 1 ternate indicator. ll68 least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by apl/l2

d. With both suppression pool water level instrumentation divisions l inoperable, restore at least one inoperable division to OPERA 8tE i status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or be in at least HOT SHUTDOWN within the next l 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 and i verify the suppression pool water level to be greater than or equal 1

. to 18'4-3/4" or 12'8", as applicable, at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by at l[68 least one0 alternate indicator. .

1/12 :

SURVEILLANCE REQUIREMENTS >

4.5.3.1 The suppression pool shall be detemined OPERABLE by verifying-l

a. The water level to be greater than or equal to, as applicable: !

i

1. 18'4-4/4" at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . l 168 'f I/12

2. 12'5" at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . ;

b. Two suppression pool water level instrumentation divisions, with I channel per division, OPERABLE with the low water level alarm i setpoint > 18'5 " or 12'8", as applicable, by performance of a

1. CHANNEL CHECK at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , f

2. CHANNEL FUNCTIONAL TEST at least once per 31 days, and f

3. CHANNEL CALIBRATION at least once per 18 months. I 4.5.3.2 With the suppression pool level less than the above limit or drained !

in OPERATIONAL CONDITION 4 or 5*, at least once per 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months s: l l

a. Verify the required conditions of Specification 3.5.3.b to be

' satisfied, or j i

b. Verify footnote conditions

to be satisfied. l GRAND GULF-UNIT 1 3/4 5-9 Amenchent No.3 8 -

r

3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT PRIMARY CONTAINMENT INTEGRITY ,

LIMITING CONDITION FOR OPERATION 3.6.1.1 PRIMARY CONTAINMENT INTEGRITY shall be maintained.

APPLICABILITY:

OPERATIONAL CONDITIO'NS 1, 2* and 3.

ACTION:

Without PRIMARY CONTAINMENT INTEGRITY, restore PRIMARY CONTAINMENT INTEGRITY within I hour 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 SHUTD0hW within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

i SURVEILLANCE REQUIREMENTS 4.6.1.1 PRIMARY CONTAINMENT INTEGRITY shall be demonstrated:

a. After each closing of each penetration subject to Type B testing, '

except the containment air locks, if opened following Type A or B test, by leak rate testing the ; wig,;r.t hets seals with gas at Pa, l$

11.5 psig, and verifying that when the measured leakage rate for -

these seals?.is added to the leakage rates determined pursuant to Surveillance Requirement 4.6.1.2.d for all other Type B and C penetrations, the combined leakage rate is less than or equal to 0.60 La.

b. At least once per 31 days by verifying that all containment penetrations ** not capable of being closed by OPERABLE containment ;

automatic isolation valves and required to be closed during accident conditions are closed by valves, blind flanges, or deactivated

! automatic valves secured in position, except as provided in Table 3.6.4-1 of Specifica" tion 3.6.4.

w W cmpbco w%h ryrweh J r c. By verifying each containment air lock"0PERACLE g r $

Specification 3.6.1.3. %g, g i

.g g ,.

d. By verifying the suppression poo DOPERA"LE g r Specification 3.6.3.1.

"See Special Test Exception 3.10.1

- Except valves , blind flanges, and deactivated automatic valves which are located inside the containment, steam tunnel or drywell and are locked, sealed or otherwise secured in the closed position. These penetrations shall be ,

verified closed during each COLD SHUTD0hH except such verification need not be performed more often than once per 92 days.

GRAND GULF-UNIT 1 3/4 6-1 Amendment No. 9j l ,

.9- .,-,.-_$__--,,,,.-,-_.__,-,_,,_%y,. . _ -., .,,,_,.,,_.,__,,._.,,--_,--,,wm..._v. -

.,,.-c.-.,,..

CONTAINMENT SYSTEMS

, CO?!TAINMENT LEAKAGE LIMITING CONDITION FOR OPERATION 3.6.1.2 Containment leakage rates shall be limited to:

a. An ov'erall integrated leakage rate of less than or equal to L ,

0.437 percent by weight' of the containment air per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months al P,,

11.5 psig.

b. A combined leakage rate of less than or eoual to 0.50 L f penetrations and all valvesN i:::d 'r T:bi: 2.5.*-1,::le:crall

'cr >

0

1se: rieb cre hydr :::. tic lly 1:n te:::d ;;r T:bl: 2. 5. E 1, $

subject to Type B and C tests wnen pressurizeo to P a

,11.5 psig.

c. Less than or equal to 100 scf per hour for all four main steam lines through the isolation valves when tested at P,, 11.5 psig.

d. A combined leakage rate of less than or equal to 1 gpm times the total, number of ECCS en: C::: containment isolation valves in hydrostatically $

tested lines which penetrate the primary containment, when tested at N 1.10 P,, 12.65 psig. -

APPLICABILITY: When PRIMARY CONTAINMENT INTEGRITY is required per Specification 3.6.1.1. '

ACTION:

With:

a. The measured overall integrated containment leakage rate exceeding 0.75 L,, or

b. The measured combined leakage rate for all penetrations and all valves *licted in Tel: 2.5.'-1, ex::pt for ;;1ve: which are hydr:- * '

staticall;. le:k tested per Tebl. 3.G.4--1x subject to Type B and C h tests exceeding 0.60 La, r

c. The measured leakage rate exceeding 100 scf per hour for all four main steam lines through the isolation valves, or

d. The measured combined leakage rate for all ECC end RCIC containment IsC isolation valves in hydrostatically tested lines which penetrate the primary containment exceeding 1 gpm times the total number of such valves, restore:

. a. The overall integrated leakage rate (s) to less than or equal to 0.75 L,, and 46= Lewdes aff vahes I;d e d in Tame S.I 4-1, e xcep t- Ec 16,se g b 3dios tabenlly lea k 4eded, g Wt are GRAND GULF-UNIT 1 3/4 6-2 4,Ac* Ah l l -

CONTAINMENT SYSTEMS LIMITING CONDITION FOR OPE MTION (Centinued)

ACTION (Continued) #

b. The combined ieskage rate for all penetrations and all valves 'i t:0 3 2- T;ble 2.5."-1, except f r ;;1;c: *ich Orc hydr ;t: tic:' , i::,

g t :::d p ' bi; 2.5.' 1, subject to Type B and C tests to less tnan or equal to 0.60 L , and a

c. The leakage rate to less than 100 scf per hour for all fcur main steam lines through the isolation valves, and

d. The combined leakage rate for all ECOS cr.d 9:IC containment isolation Ie>

W valves in hydrostatically tested lines which penetrate tne primary containment to less than or equal to 1 gpm times the total number of such valves, prior to increasing reactor coolant system temperature above 200 F.

SURVEILLANCE RE0VIREMENTS 4.6.1.2 The containment leakage rates shall be demonstrated at the following test schedule and shall be determined in conformance with the criteria speci-fied in Appendix J of 10 CFR 50 using the methods and provisions of ANSI N45.4 -

1972:

a. Three Type A Overall Integrated Containment Leakage Rate tests shall be conducted at 40 + 10 month intervals during shutdown at Pa , 11.5 psig, during each 10 year service period. The third test of each set shall be conducted during the shutdown for the 10 year plant inservice inspection.

l b. If any periodic Type A test fails to meet 0.75 L the test schedule for subsequent Type A tests shall be reviewed an8, approved by the Commission. If two consecutive Type A tests fail to meet 0.75 L Type A test shall be performed at least every 18 months until tw$, a consecutive Type A tests meet 0.75 L,, at which time the above test schedule may be resumed.

i

! c. The accuracy of each Type A test shall be verified by a supplemental test which:

1. CerN m the accuracy of the t00t by verifyir.g that the di'fereate I nsert,- 7 bctw;cn th; ;uppl;;;ntal data and the Type A test data i; within $

o See nex* 798 0.25 L,.

2. Has duration sufficient to establish accurately the change in l ,

leakage rate between the Type A test and the supplemental test.

or bled 6fm o35 L. 3. Requires the quantity of gas injected into the containa from the containment during the supplemental test to be ,qu;valeet-M M5La,

-tc at leest 25 pcreent of th; total m;asured icekes: at P,, h al. 5 p:i;;.

s

znetuor os att vatus lis ted I" 7~*bl* 3 '" V-Is except for- ssa g that este hy&o r/adiea!!y hat tested.

GRAND GULF-UNIT 1 3/4 6-3 Amendment No. l L

Insert to Technical Specification 3/4.6.1.2, Page 3/4 6-3 Confirms the accuracy of the test by verifying that the containment leakage i rate, L' calculated in accordance with ANSI N45.4-1972, Appendix C, is within 23, percent of the containment leakage rate, L , measured prior to the introduction of the superimposed leak.

ZZ51Aabm1

CONTAINMEtiT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

d. Type B and C tests shall be conducted with gas at P , 11.5 psig,* at i,ntervals no greater than 24 months except for test! involving:

1. Air locks,

2. Main steam line isolation valves,

3. Penetrations using continuous leakage monitoring systems,

4. Valves pressurized with fluid from a seal system, 3

C >

5. [C 0 u; r :C rentainment isolation valves in hycrostatically tested lines which penetrate the primary containment, and N

6. Purge supply and exhaust isolation valves with resilient material seals.

e. Air locks shall be tested and demonstrated OPERABLE per Surveillance Requirement 4.6.1.3.

f. Main steam line isolation valves shall be leak tested at least once per 18 months.

g. Type B tests for penetrations employing a continuous leakage monitoring system shall be conducted at P,, 11.5 psig, at intervals no greater than once per 3 years.

h. Leakage from isolation valves that are sealed with fluid from a seal system may be excluded, subject to the provisions of Appendix J,Section III.C.3, when determining the combined leakage rate provided the seal system and valves are pressurized to at least 1.10 P ,

12.65psig,andthesealsystemcapacityisadequatetomaint$in system pressure for at least 30 days.

C

1. ECCS and RCic containment isolation valves in hydrostatically 1 tested lines which penetrate the primary containment shall be leak N tested at least once per 18 months.

j. Purge supply and exhaust isolation valves with resilient material seals shall be testfd and demonstrated OPERABLE per Surveillance Requirement 4.6.1.9.2.

,k . The provisions of Specification 4.0.2 are not applicable to 24 month or 40 + 10 month surveillance intervals.

"Unless a hydrostatic test is required per Table 3.6.4-1.

l

/%enum nr No l GRAND GULF-UNIT 1 3/4 6-4

CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS 4.6.1.3 Each containment air lock shall be demonstrated OPERABLE:

a. Within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months #a'fter each closing, except when the" air lo k is being used for multiple entries, then at least once per 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , by verifying 235 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.

b. By conducting an overall air lock leakage test at P,,11.5 psig, and verifying that the overall air lock leakage rate is within its limit:

1. At least once per 6 months #, and

2. Prior to establishing PRIMARY CONTAINMENT INTEGRITY when maintenance has been performed on the air lock that could affect the air lock sealing capability.* .

c. At least once per 6 months by verifying that only one door in each air lock cati be opened at a time.

d. By verifying each airlock door inflatable seal system OPERABLE by:

1. Demonstrating each of the two inflatable seal pressure instrumentation channels per airlock door OPERABLE by performance of a:

a) CHANNEL FUNCTIONAL TEST at least once per 31 days, and b) CHANNEL CALIBRATION at least once per 18 months, with a low pressure setpoint of > 60 psig.

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

be greater than or equal to 90 psig.

3. At least once per 18 months, conducting a seal pneumatic system leak test and verifying that system pressure does not decay more than 2 psig from 90 psig within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

The provisions of Specification 4.0.2 are not applicable.

Exemption to Appendix J of 10 CFR 50.

Amendment No. l GRAND GULF-UNIT 1 3/4 6-6 Order

'APR 1 p 19.:

CONTAINMENT SYSTEMS CONTAINMENT STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.6 The structural integrity of the containment shall be maintained at a level consistent with the acceptance criteria in Specification t.C.1.C. lA

4. G . t . l. . I , "

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 3.

ACTION: '

With the structural integrity of the containment not conforming to the above requirements, restore the structural integrity to within the limits within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 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 ir. COLD SHUT 00WN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE REOUIREMENTS ,

4.6.1.6.1 The structural integrity of the exposed accessible interior and exterior surfaces of the containment, including the liner plate, shall be determined during the shutdown for each Type A containment leakage rate test by a visual inspection of those surfaces. This inspection shall be performed prior to the Type A containment leakage rate test to verify no apparent changes '

in appearance or other abnormal degradation. -

4.6.1.6.2 Reoorts Any abnormal degradation of the primary containment structure detected during the above required inspections shall be reported to the Commis-sion pursuant to Specification 6.9.1. This report shall include a description of the condition of the concrete, the inspection procedure, the tolerances on cracking, and the corrective actions taken.

GRAND GULF-UNIT 1 3/4 6-9 pggopckT' N0 l

CONTAINMENT SYSTEMS 3/A.6.2 DRYWELL DRYWELL INTEGRITY LIl4ITING CONDITION FCR OPERATION 3.6.2.1 DRYWELL INTEGRITY shall be maintained.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2* and 3.

ACTICN:

Without DRYWELL INTEGRITY, restore DRYWELL INTEGRITY within 1 hcur 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 .

SURVEILLANCE RECUIREMENTS 4.6.2.1 DRYWELL INTEGRITY shall be demonstrated;

a. At least once per 31 days by verifying that all drywell penetrations **

not capable of being closed by OPERABLE drywell autcmatic isolation valves and required to be closed during accident conditions are closed by valves, blind flanges, or deactivated automatic valves secured in position, except as provided in Table 3.6.4-1 of Specification 3.6.4.

is in certpl/uu. wh $c <ghe<da 4 3

b. By verifying each drywell air lock A0?: ASLE p r Specification 3.6.2.3. wp is in **^f il ues wi& #c !*yI!w'*T* *f By verifying the suppression pooln0I ASL. p;r Specification 3.6.3.1.

c. ,

A See Special Test Exception 3.10.1.

an Except valves, blind flanges, and deactivated automatic valves which are located inside the drywell or containment and are locked, sealed or otherwise secured in the closed position. These penetrations shall be verified closed during each COLD SHUTDOWN except such verification need not be performed more often than once per 92 days.

GRAND GULF-UNIT 1 3/4 6-13 Amcalme^t A/8.

I CONTAINMENT SYSTEMS )

1

$URVEILLANCE REQUIREMENTS !

l 4.6.2.3 Each drywell air lock shall be demonstrated OPERABLE: '

l 7A 4

a. Within-e hours after each closing, except when the air lock is being , ,

used for multiple entries, then at least once per 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months si 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. l

b. ^t ?:::t :::: p:r 5 centh: by ::rt: tin; en :::r:11 :f- 1::t 1::h:;:

n 5:t

?::t:;;: et *,t r_ f:,11.5;ig:dby::rfpfe;th:tth:

rfth'- it: 'f=it. :::r:11 :tr 1 :t *

c. At least once per 6 months by verifying that only one door in each air lock can be opened at a time.

d. By verifying each airlock door inflatable seal system OPERABLE by:

, 1. Demonstrating each of the two inflatable seal pressure

instrumentation cha'nnels per airlock door OPERABLE by performance of a:

a) CHANNEL FUNCTIONAL TEST at least once per 31 days, and b) CHANNEL CALIBRATION at least once per 18 months, with a low pressure setpoint of > 60 psig.

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

1 be greater than or equal to 90 psig.

f 3. At least once per 18 months, conducting a seal pneumatic system leak test and verifying that system pressure does not decay more than 2 psig from 90 psig within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

The provisions of Specification 4.0.2 are not applicable.

i -

e o

l Ry conheting an owsil oir lack kskye te,sf st Ps , ll. 3

, psis and vudyin3 L+ h ovusil air loek leakn< ra i is J&in Hs, limi+ :

i s. AtImtones fu 6 m . ~% ",

l 2, t<ior 4.establishins D MWu Tarasmry when ninfana ha been fuf"med *ktke *ir /*ck Lt cedd dir d he-i de lock sesli% caf*bility . ,

GRAND GULF-UNIT 1 3/4 6-16 A,nendment Na.

i

_.,__._.,__-.m. _ . . . , . . _ - _ _ .

O CONTAINMENT SYSTEMS DRYWELL STRUCTURAL INTEGRITY LIMITING CONDITICN FOR OPERATION 3.6.2.4 The structural integrity of the crywell shall be maintainee at a level consistent with the acceptance criteria in 5pe:ificatica 2.:.:. . 8; 4.4. a . v. / <*

ApoLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 3.

ACTION:

With the structural integrity of the drywell net conforming to the above racuire-ments, restore the structural integrity to within tne limits within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 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 EMUTOC'..M within the fellowing 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE RE0V!2EMENT5 4.6.2.4.1 The structural integrity of the exposec accessible interior and exterior surfaces of the dryweil shall be determinec curing the snu c:wn f:r each Type A containment leakage rate test by a visual inspectica cf : nose surfaces. This inspection shall be performed prior to the Type A c:n ainnent leakage rate test to verify no apparent enanges in appearance or otner abncreal degradation.

4.6.2.4.2 Recorts Any abnormal degradation of the drywell structure cete:ted durir.g the aoove required inspections shall be reported to the Co nission pursuant to Specification 6.9.1. This report shall include a cescription of the condition of the concrete, the inspection procedure, the tolerances on cracking, and the corrective actions taken.

Y GRAND GULF-UNII 1 3/4 6-17 Am44 m f Mo. -

, _ _ , , _ _ . , . ___..,_-,__._____.,-_,,.._,m. ..m_, _ _.,._- _ _

CONTA!N"ENT SYSTEMS 3/J.6.3 DEPRESSUR!?ATION SYSTEMS SUPPoESSICN PCOL LD1ITING CONDITION FOR ODE:ATICN 3.6.3.1 The suppression pool shall be OPERABLE with the pool water:

a. Volume between 135.291 ft3 and 138,851 ft3, ecuivalent to a level between 13'4-97&' and 18'M2, and a l 9-3

b. Maximum averag/12.e te.Tperature o/4r 95 F :;.rH; :PI- ~::"'L ::',:: ::: _'

lN

a-4, except that the maximum avera;e tem;:erature may be per?.itted to increase to:

1. 105*F during. testing which adds heat to the suppression pool.

2. 110*F with THERMAL POWER less than or equal to 1% of RATED THERMAL POWER.

3. 120*F with the main steam line isolation valves closed following a scram.

APPLICABILITY: OPERATIONAL CONDITICNS 1, 2 and 3.

ACTION:

a. With the suppression pool water level outside the above limits, restore the water level to within the limits within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 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 SHUTDCWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

With

b. ^:r. :P: PAT::: L CCt:::T: t' 1 ;r 2 ' P the suppression pool average ll68 water temperature greater than 95 F, restore the average temoerature to less than or equal to 95 F within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 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 , except, as permitted above:

1. With the suppression pool average water temperature greater than 105 F during testing which adcs heat to the suppression pool, stop all testing which adds heat to the suppression pool and restore the average temperature to less than 95 F within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 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 .

2. With the suppression pool average water temoerature greater than 110*F, place the reactor mode switch in the Shutdown position and operate at least one residual heat removal loop in

. the suppression pool cooling mode.

3. With the suopression pool average water temperature greater than 120*F, depressurize the reactor pressure vessel to less than 200 psig within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

5ee Specification 3.5.3 for ECCS requirements.

GRAND GULF-UNIT 1 3/4 6-20 Ame O ed b ~

l CONTAINMENT SYSTEMS I LIMITING CONDITION FOR OPERATION (Continued) f l

ACTION: (Continued)

c. With one suppression pool water level instrumentation division inoper-

)

able and/or with one suppression pool water temperature instrumentation i channel in any pair (s) of temperature instrumentation channels in the '

same sector inoperable, restore the inoperable channel (s) to OPERABLE .

status within 7 days or verify suppression pool water level and/or temperature to be within the limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

d. With both suppression pool water level instrumentation divisions [

inoperable and/or with both suppression pool water temperature i instrumentation channels in any pair (s) of temperature instru- [

, mentation channels in the same sector inoperable, restore at least !

1

. one inoperable water level division and at least one inoperable j water temperature instrumentation channel in each pair of temperature ['

instrumentation channels in the same sector to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or be in at least HOT SHUTDOWN within the next i 4

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 . !

l l

\ l

!,* SURVEILLANCE REQUIREMENTS i 4.6.3.1 The suppression pool shall be demonstrated OPERABLE:

l

a. By verifying the suppression pool water volume to be within the I

! limits at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by ll68 ,

verifying the suppression pool average water temperature to be less ;

than or equal to 95'F, except: j

1. At least once per 5 minutes during testing which adds heat to {

the suppression pool, by verifying the suppression pool average i water temperature le*;s than or equal to 105'F. ;

l

2. At least once per hour when suppression pool average water i temperature is greater than or equal to 95'F, by verifying ;

suppression pool average water temperature to be less than or i equal to 110*F and THERMAL POWER less than or equal to 1% of RATED THERMAL POWER. l

3. 'At least once per 30 minutes following a scram with suppression ;

pool average water temperature greater than or equal to 95'F. i by verifying suppression pool average water temperature less i than or equal to 120*F.

1 GRAND GULF-UNIT 1 3/4 6-21 Amendment No. 8 3- ! ,

CONTAINMENT SYSTEMS CONTAINMENT SPRAY LIMITING CONDITIO'N FOR OPERATION 3.6.3.2 The containment spray mode of the residual. heat removal (RHR) system shall be OPERABLE with two independent loops, each loop :ensisting of:

a. One OPERABLE RHR pump, and

b. An OPERABLE flow path capable of recirculating water from the suppression pool through a -56W- heat exchanger, and t/m containment 012 RHR 16 9 OPERATIONAL CONDITIONS 1, 2 and 3.

es.

APPLICABILITY:

ACTION:

a. With one containment spray loop inoperable, restore the inoperable loop 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 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 withir. the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b. With both containment spra oops inopera be in at least HOT 012 SHUTOOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months nd in COLD SHUTDOWN" within the eeset followm.g 24 hourr. the. next restore at least one locP to OPERABLE SURVEILLANCE REOUIREMENTS 4.6.3.2 The containment spray mode of the RHR system shall be demonstrated OPERABLE:

a. At least once per 31 days by verifying that each valve, manual, power operated or automatic, in the flow path that is not locked, sealed or othemise secured in position, is in its correct position.

b. By verifying that each of the required RHR pumps develops a flow of at least 5650 gpm en recirculation flow through the RHR heat exchange to the suppression pool when tested pursuant to Specification 4.0.5.

c.. At least once per 18 months by performance of a system functional test which includes simulated automatic actuation of the system throughout its emergency operating sequence and verifying that each automatic valve in the flow path actuates to its correct position.

Actual spraying of coolant into the primary containment may be excluded from this test,

d. Aoo tNssRT l169 7

"Whenever botn RHR subsystems are ineperable, if unable to attain COLD SHUTDOWN as required by this ACTION, maintain reactor coolant temperature as low as practical by use of alternate heat removal methods.

Admendment yo. l GRAND GULF-UNIT 1 3/4 6-24

) 1 Insert to Technical Specification 3/4.6.3.2, Page 3/4 6-24 By performance of an air or smoke flow test of the containment spray nozzles and verifying that each spray nozzle is unobstructed following maintenance which could result in nozzle blockage.

1 I

i i

i

[

l T

f

( i-I t 1 i

i ZZ51Aaba6 l - - - . . - . . , - - _ _ _ . . _ _ . _ ..,___ _.__..__._ ___,,_ ., __., __ _ ,_ . _ ,__

CONTAINMENT SYSTEMS SuoPRESSION POOL C::C'.!NG LDtITING CONDITION FOR OPERATION 3.6.3.3 The suppression pool cooling mode of the residual heat removal (RHR) system shall be OPERABLE with two independent loops, each loop consisting of:

a. One OPERAELE RHR pump; anc

b. An OPERAELE flow path capable of recirculating water from tne suppression poci tnrough a ST. heat excnanger. 1012 RHR APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 3.

ACTICN:

72 houvS

a. With one suppression pool cooling icco inocerable, restore the inoperable loop to GPERABLE status withi 7 n.,; or be in at least l 169 HOT SHUTDOWN within the nr.xt 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hears,

b. With both suppression pool cooling locos inoperable, restore at least one loop to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 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 .

SURVEILLANCE REOUIREMENTS 4.6.3.3 The suppression pool cooling mode of the RHR system shall be demonstrated OPERABLE:

a. At least once per 31 days by verifying that each valve, manual, power operated or automatic, in the flow path that is not locked, sealed or otherwise secured in position, is in its correct position. ,

b. By verifying that each of the required RHR pumps develops a flow of at least 7450 gpm on recirculation flow through the RHR heat exchangers to the suppression pool when tested pursuant to Specification 4.0.5.

"Whenever both RHR subsystems are inoperable, if unable to attain COLD SHUTDOWN as required by this ACTION, maintain reactor coolant temperature as low as ,

practical by use of alternate heat removal methods. l GRAND GULF-UNIT 1 3/4 6-25 Admednfed No- l

CONTAINMENT SYSTEMS 3/4.6.4 CONTAINMENT AND DRYWELL ISOLATION VhlVES -

LIMITING CONDITION FOR OPERATION 3.6.4 The containment:and drywell isolation valves shown in Table 3.6.4-1 shall be OPERABLE with isolation times less than or equal to those shown in

~

Table 3.6.4-1. '

APPLICA9ILITY: OPERATIONAL CONDITIONS 1, 2, 3, and #.

ACTION:

g With one or more of the containment or drywell isolation valves shown in Table 3.6.4-1 inoperable, maintain at least one isolation valve OPERABLE in each affected penetration that is open and within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months either:

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

b. Isolate each affected penetration by use of at least one deactivated aut:matic valve secured in the isolated position," or

c. Isolate each affected penetration by use of at least.o,ne closed .

manual valve or blind flange." ..

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

i

,emcepf FASIVs 4

" Isolation valvesAclosed to satisfy these requirements may be reopened on an intermittent basis under administrative controls. 4

Isolation valves shown in Table 3.6.4-1 are also required to be OPERABLE when their associated actuation instrumentation is required to be OPERAELE per ."

Table 3.3.2-1.

GRAND GULF-UNIT 1 3/4 6-27 Amendment No. 9, -

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

. . _ . . . _ . . . _ ~.

~ - -

. TABLE 3.6.4-1 CONTAINMENT AND DRYWELL ISOLATION VALVES MhXIMUM SYSTEM AND . PENETRATION. ' ISOLATION TIME i i

VALVE NUMBER NUMBER VALVE GROUP (*) (Seconds)

1. Automatic Isolation Valves f

a. Containment .

Main Steam Lines '

B21-F028A 5(0)* 1 5 '

Main Steam Lines 821-F022A 5(I)* 1 5 4 Main Steam Lines B21-F067A-A 5(0)* 1 #'1 Q Main Steam Lines B21-F0288 6(0)* 1 5 -

Main Steam Lines 821-F0228 6(I)* 1 5 .e l Main Steam Lines B21-F0678-A 6(0)* 1 #1 14 Main Steam Lines B21-F028C 7(0)* 1 5 Main Steam Lines B21-F022C 7(I)* 1 5 a >

Main Steam Lines B21-F067C-A 7(0)* 1 E9 lX Main Steam Lines B21-F028D 8(0)* 1 5 I Main Steam Lines 821-F022D 8(I)* 1 5 4 Main Steam Lines 824-F067D-A 8(0)* ~

1 E1 $

RHR Reactor E12-F008-A 14(0) N 3 40 [. o Shutdown Cooling N !

0 :

Suction RHR Reactor E12-F009-8 14(I) N 3 40 Shutdown Cooling i

.- Suction ,

Steam Supply to E51-F063-B 17(I) 4 20 RHR and RCIC Turbine ;

Steam Supply to E51-F064-A 17(0) 4 20 RHR and RCIC Turbine ;

Steam Supply to E51-F076-B 17(I) 4 20 -

RHR and RCIC Turbine J0 l RHR to Head Spray E12-F023-A 18(0) N 3 96 e4 l$ ;

Main Steam Line B21-F019-A 19(0) 1 pf L' y*

Drains l (a) See Specification 3.3.2, Table 3.3.2-1, for isolation signal (s) that J operates each valve group.

(b) Hydrostatically tested to ASME Section XI criteria. .

1 (c) Hydrostatically tested with water to 1.10 P , 12.65 psig.

(d) Hydrostaticallytestedbypressurizingsyst$mto1.10P,,12.65psig.

(e) Hydrostatically tested during system functional tests.

(f) Hydrostatically sealed by feedwater leakage control system. Type C '

test not required.

(g) Normally closed os locked closed manual valves may be opened on an lf l intermittent basis under administrative control.

The provisions of Specification 4.0.4 are not applicable for entry into OPERATIONAL CONDITIONS 2 or 3 provided the surveillance is performed within L 12 hours after reaching a reactor steam pressure of 600 psig and prior to entry into OPERATIONAL CONDITION 1.

l

The "- A - 5,- C ' tr/ cal-(h);-(B),-(C) "deripators on the vahe swaher.r tha%fe $ 1 associa$d efec Swsias.

! GRAND GULF-UNIT 1 3/4 6-29 Amendment No. 9y l

TABLE 3.6.4-1 (C:ntinued)

CONTAINMENT AND DRYWELL ISOLATION VALVES MAXIMUM PENETRATION ISOLATION TIME SYSTEM AND NUMBER VALVE GROUP (8) (Seconds)

VALVE NUMBER Containment (Continued) 19(I) 1 -H 2.*

Main Steam Line B21-F016-B Drains 22

- ::" M::t E :5.: ;:r E22-F0$2." " 20(!) E ,

MAH am i B f* T

" ' .28. 9o RHRHeai5xchanger E12-F028A-A 20(I) N 5 O "A" to LPCI 63- 7 4 ,o E

~

RHR Heat Exchanger E12-F037A-A 20(I)N 3 "A" to LPCI 22

" M::t Ex:5.: ;:r E12-F0'2" " 21(!)N 5 ,

mon .. i o ,, ,

38- to 21(I) N 5

RHRHeatExEhanger E12-F028B-B "B" to LPCI +S- 7 Y i

RHR Heat Exchanger E12-F0378-B 21(I)N 3 "B" to LPCI 5 90 RHR "A" Test Line E12-F024A-A 23(0)(d) to Supp. Pool 5 36 RHR "A" Test Line E12-F011A-A 23(0)(d) 2 1 to Supp. Pool 5 -Mi- fv4 RHR "C" Test Line E12-F021-B 24(0) N i to SJpp. Pool n 6B T* 75 HPCS Test Line E22-F023-C 27(0)(d) 4 56 RCIC P mp Suction E51-F031-A 28(0)(d) 9 26 RCIC Turbine E51-F077-A 29(0)(C)

Exhaust 5 144 h LPCS Test Line E21-F012-A 32(0)N 7 4 Cont. Purge and M41-F011-(A) 34(0) l Vent Air Supply 4 i

7 ,

Cont. Purge and M41-F012<B) 34(I) o Vent Air Supply n 7 4 Cont. Purge and M41-F034-(a) 35(I) and Vent Air Exh. 7 4 Cont. Purge and M41-F035-U0 35(0) and Vent Air Exh. 33 36(I) 6A Plant Service P44-F070-B l

Water Return 6A -24 33 Plant Service P44-F065-A 36(0)

Water Return -24 33 9 37(0) 6A Plant Service P44-F053-A S Water Supply 6A 40- is Chilled Water P71-F150-M) 38(0)

Supply l 5

3/4 6-30 Amendment No. 9, 10, - l GRAND GULF-UNIT 1

- TABLE 3.6.4-1 (Centinued)

CONTAINMENT AND DRYWELL ISOLATION VALVES MAXIMUM SYSTEM AND PENETRATION ISOLATION TIME VALVE NUMBER NUMBER VALVE GROUP (a) (Seconds)

~

Containment (Continued)

Chilled Water ' P71-F148-CA') 39(0) 6A 12.

Return .

Chilled Water P71-F149-C8) 39(I) 6A -BG-#2.

Return Service Air P52-F105 -CO 41(0) 6A -t 4 Supply I'nst. Air Supply P53-F001-CO 42(0) 6A -4 . b RWCU to Main G33-F034-A 43(0) 8 - 35 Condenser RWCU te Main G33-F028-B 43(I) 8 -24 35

. Condenser RWCU Backwash to G36-F106-(20 49(I) '6A - i t '

C/U Phase Sep. Tank RWCU Backwash to G36-F101-(R) 49(0) 6A Il C/U Phase Sep. Tank Drywell & Cont. P45-F067-(0) 50(I) 6A -4 h Equip. Drain Sump Disch. 4 Drywell & Cont. P45-F068-(^) 50(0) 6A 7 o g

Equip. Drain ,

Sump Disch.

Drywell & Cont. P45-F061-(0) 51(I) 6A 7 Floor Drain Sump Disch.

Drywell & Cont. P45-F062-N 51(0) 6A -4 7 Floor Drain Sump Disch.

Condensate Supply P11-F075-(4) 56(0) 6A Jo FPC & CU to Upper G41-F028-A 57(0) 6A W Si i Cont. Pool Upper Cont. Pool G41-F029-A 58(0) 6A Si to Fuel Pool Drain Tank Upper Cont. Pool G41-F044-B 58(I) 6A 40 to Fuel Pool Drain Tank .

Aux. Bldg. Fir. P45-F273-A 60(0) 6A -e3- h and Equip. Drn.

Tks. to Supp. Pool $

Aux. B1dg. Fir. P45-F274-B 60(0) 6A -34. 12.

and Equip. Drn.

Tks. to Supp. Pool GRAND GULF-UNIT 1 3/4 6-31 Amendment No. 9, -l

-.. her~ -- ,

TABLE 3.6.4-1 (Cantinued)

CONTAINMENT AND ORYWELL ISOLATION VALVES MAXIMUM SYSTEM AND PENETRATION ISOLATION TIME VALVE NUMBER NUMBER VALVE GROUP (a) (Seconds)

Containment (Continued)

E61-F0094A) 65(0) 7 4 s Comb. Gas Control Cont. Purge (Outside Air Supply)

E61-F01048) 65(I) 7 4 Comb. Gas Control Cont. Purge 3 (Outside Air y Supply)

Purge Red- Filter E61-F056-0) 66(1) 7 4 Tra*a Oete;t;r Isot.W s Purge 4ed- Filter E61-F057(A) 66(0) 7 4 k Train Cet;;t:r Isola b E12-F024B-B 67(0)(d) 5 90 RHR "B" Test Line To Suppr. Pool %

RHR "B" Test Line E12-F0118-B 67(0)(d) 5 f7- a To Suppr. Pool g Refueling Water P11-F130 4A) 69(0)(C) 6A +e

,Transf. Pump Suction Refueling Water P11-F131-CB.) (9(0)(c) 6A 8 Transf. Pump Suction 70(0) 6A 4 Instr. Air to ADS P53-F003-A RCIC Turbine Exh. E51-F078-B '/5(0) 9 4-Vacuum Breaker i RWCU to Feedwater G33-F040-B 83(I) 8 8

&'8 49-J 1

RWCU to Feedwater G33-F039-A 83(0) l Chemical Waste P45-F098 -(0) 84(I) 6A +a l

Sump Discharge Chemical Waste P45-F099 -(M 84(0) 6A +8 Sump Discharge P60-F009-A 85(0) 6A 8 Supp. Pool Clean-

. up Return +e Supp. Pool Clean- P60-F010-8 85(0) 6A up Return Demin. Water P21-F017-A 86(0) 6A le- a O

Supply to Cont. le '9

P21-F018-B 86(I) 6A Demin. Water Supply to Cont.

RWCU Pump Suction G33-F001-B 87(I) 8 & #

3/4 6-32 Amendment No. 9, 10 3 l GRAND GULF-UNIT 1

TABLE 3.6.4-1 (Continued)

CONTAINMENT AND ORYWELL ISOLATION VALVES -

. $XIMUN SYSTEM AND PENETRATION ISOLATION TIME VALVE NUMBER NUMBER VALVE GROUP (*) (Seconds) l Containment (Continued) !

RWCU Pump Suction G33-F252-A 87(I) 8 40 35 RWCU Pump Suction G33-F004-A 87(0) 8 4G 35

RWCU Pump Disch. G33-F053-B. 88(I) 8 43 35 1 :

RWCU Pump Disch. G33-F054-A 88(0) 8 --84 35 l

b. Drywell 335 !

Instrument Air P53-F007-B -3M(0) 6A + 7-Plant Service P44-F076-A 331(I) 6A 32 Water Return Plant Service P44-F077-B 331(0) 6A 32 l Water Return t Plant Service P44-F074-B 332(0) 6A 32 Water Return RWCU Pump Suction G33-F250-A 337(I) 8 40 35 i RWCU Pump Suction G33-F251-B 337(0) 8 49 35 J Combustible Gas E61-F003B-B 338(0) 5 4G M 2 Con.

Combustible Gas E61-F003A-A 339(0) 5 6& M Con. :

Combustible Gas E61-F005A-A 340(0) 5 84 -

Con.

Combustible Gas . E61-F0058-B 340(0) 5 84 Con. .-

Combustible Gas E61-F007-(A) 341(0) 5 9 ,

1 Con. ;

Combustible Gas E61-F020 -(a) 341(0) 5 18 Con.

Drywell Air Purge M41-F015-(o) 345(I) 7 4 ,

Supply !

Drywell Air Purge M41-F013 -Cs) 345(0) 7 4 Supply Drywell Air Purge M41-F016-CO 347(I) 7 4 '

! Exhaust

Drywell Air Purge M41-F017-(0) 347(0) 7 4 Exhaust Equipment Drains P45-F009-C e 348(I) 6A -+ b 3 !

Equipment Drains P45-F01Q-ta) 348(0) 6A +b . l Floor Drains -

P45-EQO3-l A ) 349(I) 6A -46 6 !

Floor Drains P45-F004 -(. 6 ) 349(0) 6A 6

Service Air PS2-F195-B 363(0) 6A 16 l Chemical Sump P45-F096-A 364(I) 6A 4- 9 Disch. .

Chemical Sump P45-F097-B 364(0) 6A 9

! Disch. 0 RWCU to Heat G33-F253 -8 366(0) 8 -35 35 ,

Exch. 36

! Reactor Water B33-F019 -5 465(1) 10 -24r4- ;

Sample Line 36 Reactor Water B33-F020 - A 465(0) 10 -28r4-Sample Line GRAND GULF-UNIT 1 3/4 6-33 Amendment No. 9, l l

. - . - , . . , , - _ _ , _ . , _ _ , _ - - _ _.._._,_.,_,_,.,y, ,m,,---.,_.--_y,.-.c m- ,, ,,._m__,.,,-p_,y -

,, , . , , .,y,_y_

TABLE 3.6.4-1 (Continu:d)

CONTAINMENT AND DRYWELL 350LATION VALVES SYSTEM AND PENETRATION i VALVE NUMBER NUMBER

2. Manual Isolation Valves I9)#

a. Containment Main Steam Lines E32-F001A-A' 5(0)

Main Steam Lines E32-F001E-A 6(0)

Main Steam Lines E32-F001J-A 7(0)

Main Steam Lines E32-F001N-A 8(0) ,

Feedwater Inlet . B21-F065A-A 9(0)(b)

Feedwater Inlet B21-F065B-A 10(0)(D)

RHR Pump "A" E12-F004A-A 11(0)(d) '

Suction RHR Pump "B" E12-F004B-B 12(0)(d)

. Suction RHR Pump "C" E12-F004C-B 13(0)(d) ,

l,3. !

Suction A N HR Heat Ex. "A" E12-F327A-A 20(0) N :

7H H Ex. "B" E12-F027B- 8 21(0) N $

to LPCI o RHR Pump "C" to .' E12-F042C-B 22(O') N LPCI t RHR "A" Test Line E12-F064A-A 23(0)(d) '

To Suppr. Pool O RHR "C" Test Line E12-F064C-B 24(0) N @

To Suppr. Pool HPCS Suction E22-F015-C (d)

HPCS Discharge E22-F004-C 25(0) 26(0) N g HPCS Test Line E22-F012-C (d) i RCIC Turbine Exh.

LPCS Pump Suction E51-F068-A

E21-F001-A 27(0)fC 29(0) 30(0)y LPCS Pump E21-F005-A 31(0) o N

, Discharge O i

LPCS Min. Flow E21-F011-A 32(0) N CRD Pump C11-EQ83-A 33(0) :

Discharge '

CCW Supply P42-F066-A 44(0)

CCW Return P42-F067-A 45(0) '

CCW Return P42-F068-B 4

RCIC Pump E51-F019-A 45(I)(d) 46(0) ,

Discharge Min. Flow -

, Reactor Recire. B33-F128-B 47(I)

Post Accident

Sampling GRAND GULF-UNIT 1 3/4 6-34 Amendment No. 9, I l

i Inserts to Table 3.6.4-1, Page 3/4 6-34 Insert A RHR Heat Exchanger E12-F042A-A 20(I)

"A" to LPCI Insert B RHR Heat Exchanger E12-F042B-B 21(I)

"B" to LPCI l

4 e

i 1

t i

l I

i i

l ZZ51Aabe4

.

TABLE 3.6.4-1 (Centinued)

CONTAINMENT AND DRYWELL ISOLATION VALVES SYSTEM AND PENETRATION-VALVE NUMBER NUMBER Containment (Continued)

Reactor Recirc. B33-F127-A 47(0)

Post Accident Sampling -

Vent Header to E12-F073B-B- 4B(0)(d)

Supp. Pool RHR Pump "B" E12-F064B-B 67(0)(d)

Test Line RHR "C" Relief E12-F346-B 71B(0)(C)

'Viv. Vent Hdr.

to Suppr. Pool

& Post-Acc.

Sample Ret.

RHR Heat Ex. "A" E12-F073A-A 77(0)(d)

Relief Reactor Recire. B33-F126-B 81(I)

Accident Sampling Reactor Recire. B33-F125-A 81(0)

Accident Sampling SSW Supply "A" P41-F159A-A 89(0)(c)

SSW Return "A" P41-F16BA-A 90(I)(c)

S54 Return "A" P41-F160A-A SSW Return "B" P41-F168B-B 90(0)((C)

S!W Return "B" P41-F160B-B 91(I)(C)

SSW Supply "B" P41-F159B-B 91(0)(c) 92(0) c)

Drywell Press. M71-F593-A 101C(0)

Inst.

Drywell Press. M71-F591A-A 101F(0)

Inst.

Dr.ywell Press. M71-F591B-B 102D(0)

Inst.

I Ctmt. Press. Inst. M71-F592A-A 103D(0)

Ctmt. Press. Inst. M71-F592B'B 104D(0)

Drywell H2 E61-F595C-(4) 106A(0)

Analyzer Sample l Drywell H2 E61-F595D-(s) 106A(I)

Analyzer Sample Drywell H2 Ana- E61-F557C-(A) 106B(0) lyzer Sample Ret.

Drywell H2 Ana- E61-F597D-(.8 ) 106B(I) lyzer Sample Ret. .n Ctat. H2 E61-F596C-CA) 105A(0) o Analyzer Sample n Ctmt. H E61-F596D-CB) 105A(I) >

Analyzer Sample Ctmt. H2 Analyzer E61-F598C-(O 106E(0)

Sample Ret.

Ctmt. H2 Analyz.r E61-F5980-(8 ) 306E(I)

Sample Ret.

GRAND GULF-UNIT 1 3/4 6-35 Amendment No.3 9 - 1

TABLE 3.6.4-1 (Continued)

CONTAINMENT AND DRYWELL ISOLATION VALVES SYSTEM AND PENETRATION VALVE NUMBER NUMBER Containment (Continued)

Ctmt. H 2 A,nalyzer E61-F596A-68) 108A(0)

Sample Ctmt. H 2 Analyzer E61-F596B-(a) 108A(I)

Sample Ctmt. H 2 Analyzer E61-F598A-(d) 107B(0)

Sample Ret.

Ctmt. H 2 Analyzer E61-F598B -(8) 107B(I)

Sample Ret.

Drywell H2 E61-F595A-W 107D(0) 4 Analyzer Sample j Drywell H2 E61-F595B-fo) 107D(I) :

Analyzer Sample Drywell H2 Ana- E61-F597A -C4) 107E(0) lyzer Sample Ret.

Drywell H2 Ana- E61-F5978-(8) 107E(I) lyzer Sample Ret.

Drywell Fiss. D23-F592-A 109A(0)

Prod. Monitor -

Sample Drywell Fiss. D23-F591-B 109A(I)

Prod, Monitor Sample Drywell Fiss. 023-F594-A 109B(0)

Prod. Mon. '

Smp1. Ret.

Drywell Fiss. D23-F593-B 109B(I)

Prod. Mon.

Smp1. Ret.

Ctat. Press. Inst. M71-F594-B 109D(0)

(Post Acc. Smp1.)

Ctmt. Press. Inst. M71-F595-A 109D(I)

(Post Acc. Smp1.)

Suppr. Pool Level E30-F593A-A 113(0)(c)

Inst.

Suppr. Pool Level E30-F592A-A 114(0)

Inst.

Suppr. Pool Level E30-F594A-A - 115(0)(c)

Inst.

Suppr. Pool Level E30-F591A-A 116(0)

Inst.

Suppr. Pool Level E30-F5938-B 117(0)(c)

Inst. .

Suppr., Pool Level E30-F592B-B 118(0)

Inst.

Suppr. Pool Level E30-F594B-B 119(0)(C)

Inst.

Suppr. Pool Level E30-F5918-B 120(0)

Inst. .

GRAND GULF-UNIT 1 3/4 6-36 Amendment No. l

9 TABLE 3.6.4-1 (Continu2d)

CONTAINMENT AND DRYWELL ISOLATION. VALVES SYSTEM AND PENETRATION VALVE NUMBER NUMBER

b. Drywell Cont. Cooling :P42-F114-B 329(0)

Water Inlet Cont. Cooling P42-F116-A ' 330(I)

Water Outlet Cont. Cooling P42-F117-B 330(0)

Water Outlet ,

3. Other Isciation Valves (9) #

a. Containment Fuel Transfer F11-E015 4(I)

Tube Feedwater Inlet B21-F010A Feedwater Inlet B21-F032A 9(I)((I) 9(0) I)

Feedwater Inlet B21-F010B Feedwater Inlet B21-F032B 10(I)(CI)

I)

RHR "A" Suction E12-F017A 10(0)(d)

RHR "B" Suction E12-F0178 11(0)(d)

RHR "C" Su: tion E12-F017C 12(0)(d) 13(0)

RHR Shutdown E12-F308 14(I) N .

Cooling Suction f RHR Head Spray E51-F066$) 18(I) m RHR Head Spray E12-F344 18(I) g RHR Heat Ex. "A" E12-F044A 20(I) to LPCI RHR Heat Ex. "A" E12-F025A 20(I) N -

to LPCI RHR Heat Ex. "A" E12-F107A 20(I) N o N

to LPCI RHR Heat Ex. "B" E12-F025B 21(I) N to LPCI RHR Heat Ex. "B" E12-F044B 21(I) N to LPCI RHR' Heat Ex. "B" E12-F107B 21(I) N to LPCI GRAND GULF-UNIT 1 3/4 6-37 Amendment No. 4, 7, 9,_ l

I TABLE 3.6.4-1 (Continuid)

CONTAINMENT AND DRWELL ISOLATION VALVES SYSTEM AND PENETRATION VALVE NUMBER NUMBER Containment (Continued)

RHR Heat Ex. "C" .E12-F234 22(0) N k to LPCI RHR Pump "C" to E12-F041C-B 22(I) N LPCI 23(0)I')

~

RHR Pump "A" Test E12-F259 Line to Suppr.

. Pool

- RHR Pump "A" Test E12-F261 23(0)I')

Line to Suppr.

Pool RHR Pump "A" Test E12-F227 23(0)I')

Line to Suppr.

Pool RHR Pump "A" Test E12-F262 23(0)(')

Line to Suppr.

Pool RHR Pump "A" Test E12-F228 23(0)(')

Line to Suppr.

Pool RHR "A" Test Line E12-F290A-A 23(0)(d) to Supp. Pool RHR Pump "A" Test E12-F338 23(0)(g)

Line to Suppr.

Pool RHR Pump "A" Test E12-F339 23(0)I#)

Line to Suppr.

Pool RHR Pump "A" Test E12-F260 23(0)(*)

Line to Suppr.

Pool RHR Pump "C" Test E12-F280 24(0) N o Line to Suppr. g Pool RHR Pump "C" Test E12-F281 24(0) N l

Line to Suppr.

I Pool (d) J E22-F014 o

( HPCS Suction 25(0) N l% N HPCS Discharge E22-F005-(c) 26(I)

26(I) N l

HPCS Discharge E22-F218 HPCS Discharge E22-F201 26(I)

HPCS Test Line E22-F035 27(0)(*)

HPCS Test Line E22-F302 27(0)(*)

HPCS Test Line E22-F301 27(0)(d)

LPCS Pump Suction E21-F031 30(0) 31(I) lT 31(I) N LPCS Discharge E21-F006 -33

4 LPCS Discharge E21-F200 LPCS Discharge E21-F207 31(I) $

LPCS Test Line E21-F217 32(0) g LPCS Test Line E21-F218 32(0) 3/4 6-38 Amendment No. 9, 10, l GRAND GULF-UNIT 1

TABLE 3.6.4-1 (Continued)

CONTAINMENT AND DRYWELL ISOLATION VALVES SYSTdM AND PENETRATION VALVE NUMBER NUMBEP.

Containment (Continued)

CRD Pump C11-F122 33(I)

Discharge PSW Supply P44-F043 37(I)

Plant Chilled P71-F1El 38(I)

Water Supply Service Air P52-F122 41(I)

Supply Instr. Air Supply P53-F002 CCW Supply P42-F035 42(I)j) 44(I)(

i RCIC Disch. E51-F251 46(0) $

Min. Flow o RCIC Disch. E51-F252 46(0)($)

Min. Flow RHR Heat Ex. "B" E12-F055B 48(0)(d)

Relief Vent Header RHR Heat Ex. "B" E12-F103B 48(0)(d)

Relief Vent ,

Header RHR Heat Ex. "B" E12-F104B 48(0)(d)

Relief Vent Header Refueling Wtr. G41-F053 54(0)

Stg. Tk. to Upper Ctmt. Pool Refueling Wtr. G41-F201 54(I)

Stg. Tk. to Upper Ctat. Pool Condensate Supply P11-F004 56(I)

FPC & CU to Upper G41-F040 57(I)

Cont. Pool Stby. Liquid C41-F151 61(I)

Control Sys.

Mix. Tk.

(future use)

Stby. Liquid C41-F150 61(0)

Control Sys.

Mix. Tk.

(future use)

RHR Pump "B" Test E12-F276 67(0)(')

Line RHR Pump "B" Test E12-F277 67(0)(')

.- Line RHR Pump "B" Test E12-F212 67(0)(')

Line 3/4 6-39 Amendment No. 9, l GRAND GULF-UNIT 1

TABLE 3.6.4-1 (Continued)

CONTAINMENT AND DRYWELL ISOLATION VALVES SYSTEM AND PENETRATION VALVE NUMBER NUMEER Containment (Continued) 67(0)I')

RHR Pump "B" Test E12-F213 Line RHR Pump "B" Test E12-F249 67(0)I')

Line RHRPump"B"T$st E12-F250 67(0)I')

Line RHR Pump "B" Test E12-F334 67(0)(C)

Line RHR Pump "B" Test E12-E335 67(0)(c)

Line RHR "B" Test Line E12-F290B-B 67(0)Id)

To Suppr. Pool Inst. Air to ADS P53-F006 70(I) (d)

LPCS Relief Valve E21-F018 71A(0)

Vent Header RHR Pump "C" E12-F025C 71B(0)Id)

Relief Valve 1 Vent Header RHR Shutdown E12-F036 73(0) N o !

N Vent Header O RHR Shutdown E12-F005 76B(0) N Suction Relief Valve Disch.

RHR Heat Ex. "A" E12-F055A 77(0)Id)

Relief Vent Header RHR Heat Ex. "A" E12-F103A 77(0)Id)

Relief Vent Header l RHR Heat Ex. "A" E12-F104A 77(0)(d)

Relief Vent Header SSW "A" Supply P41-F169A 89(I)

SSW "B" Supply P41-F169B 92(I)

J Ctat. Leak Rate M61-F015 110A(I) 1 Test Inst.

l Ctat. Leak Rate M61-F014 110A(0)

Test Inst.

Ctat. Leak Rate M61-F019 110C(I)

Test Inst. '

Ctat. Leak Rate M61-F016 110C(0)

Test Inst.

Ctat. Leak Rate M61-F017 110F(I) -

Test Inst.

Ctat. Leak Rate M61-F016 110F(0)

Test Inst.

3/4 6-40 Amendment No. 9, 10,. l GRAND GULF-UNIT 1

TABLE 3.6.4-1 (Continued)

CONTAINMENT AND DRYWELL ISOLATION. VALVES .

SYSTEM AND PENETRATION VALVE NUMBER NUMBER

4. Test Connections (9)

a. Containment .

Main Steam T/C B21-F025A 5(0)

Main Steam T/C B21-F025B 6(0)

Main Steam T/C B21-F025C 7(C)

Main Steam T/C B21-F025D Feedwater T/C B21-F030A 8(0)(f)

Feedwater T/C B21-F063A 9(0)(I) 9(0)

Feedwater T/C B21-F063B 10(0)(I) CI)

Feedwater T/C B21-F030B e RHR Shutdown Cool. E12-F002 10(0) 14(0) N @

Suction T/C RCIC Steam Line E51-F072 17(0)

T/C RHR to Head E12-F342 18(0) N Spray T/C O RHR to Head E12-F061 16(0)N o Spray T/C LPr.I "C" T/C E12-F056C 22(0)

RHR "A" Pump E12 322 23(0)

Test Line T/C RHR "A" Pump E12 2336 23(0)(C)

Test Line T/C RHR "A" Pump E12-F349 23(0)(C)

Test Line T/C RHR "A" Pump E12 303 23(0)(c)

Test Line T/C RHR "A" Pump E12-F310 23(0)(c)

Test Line T/C '

RHR "A" Pump E12-F348 23(0)(g)

Test Line T/C RHR"C" Pump E12-F311 24(0) N o Test Line T/C N O

RHR"C" Pump E12-F304 24(0) N Test Line T/C .

l HPCS Discharge T/C E22-F021 26(0) '

HPCS Test Line T/C E22-F303 HPCS Test Line T/C E22-F304 27(0)(c)

RCIC Turbine E51-F258 27(0)(c) 29(0)

Exhaust T/C i RCIC Turbine E51-F257 29(0)(C)

Exhaust T/C LPCS T/C E21-F013 31(0) M o

' LPCS Test Line E21-F222 32(0)4,9, N O

~

T/C LPCS Test Line E21-F221 32(0) M T/C GRAND GULF-UNIT 1 3/4 6-42 Amendment No. 4, 7, 9, [ ,

TABLE 3.6.4-1 (Continued)

CONTAINMENT AND DRYWELL ISOLATION. VALVES SYSTEM AND PENETRATION VALVE NUMBER NUMBER Containment (Continued)

RHR "B" Test Line E12-F350 . 67(0)(c)

T/C RHR "B" Test Line E12-F312 67(0)(c)

T/C RHR "B" Test Line E12-F305 67(0)(c)

T/C Refueling Water P11-F425 69(0)(c)

Transf. Pump .

Suction T/C !

Refueling Water P11-F132 69(0)(C)

Transf. Pump !

Suction T/C Inst. Air to ADS P53-F043 70(0)

T/C i Cont. Leak Rate M61-F010 82(I) !

T/C RWCU To Feedwater 333-F055 83(0)

T/C Suppr. Pool P60-F011 85(0)

Cleanup T/C Suppr. Pool P60-F034 85(0)

Cleanup T/C ,

RWCU Pump Suction G33-F002 87(0) [

T/C :

RWCU Pump G33-F061 88(0) ,

Discharge T/C P41-F163A SSW T/C SSW T/C P41-F163B 89(0)((c) 92(0) c)

b. Drywell .

.LPCI "A" T/C E12-F056A 313(0)

LPCI "B" T/C E12-F056B 314(0)

l Instrument Air T/C P53-F493 n5 MJ(0) , ,

t SLCS T/C C41-F026 328(0)

! Service Air T/C P52-F476 363(0) ~

RWCU T/C G33-F120 366(I)

Reactor Sample B33-F021 465(0)

T/C GRAND GULF-UNIT 1 3/4 6-44 Amendment No. 9, l

i CONTAINMENT SYSTEMS 3/4.6.6 SECONDARY CONTAINMENT .

SECONDARY CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.6.1 SECONDARY CONTAINMENT INTEGRITY shall be maintained.

APPLICABILITY: OPERATIONALCONDITIdNS1,2,3and".

ACTION:

Without SECONDARY CONTAINMENT INTEGRITY:

a. 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 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 *

, suspend handling of irradiated fuel in the primary or secondary containment. CORE ALTERATIONS and operations- ,

with a potential for draining the reactor vessel. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REOUIREMENTS 4.6.6.1 SECONDARY CONIAINMENT INTEGRITY shall be demonstrated by:

a. Verifying at least once per 31 days that: -

1. All Auxiliary Building and Enclosure Building equipmen*

hatches and blowout panels are closed and sealed. '

2. The door in each access to the Auxiliary Building and Enclosure i Building is closed, except for routine entry and exit.

. 3. All Auxiliary Building and Enclosure Building penetrations not -

capable of being closed by CPERABLE secondary containment automatic isolation dampers / valves and required to be closed during accident '

or deactivated

conditionsareclos6d'byvalves,blindflanges,f automatic dampers / valves secured in position. l5

b. At least once per 18 months: ruptore discs, i

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

t t the secondary cgntainment to greater than or equal to 0.25 inches of vacuum water gauge in less than or equal to 120 seconds, and

.2. Operating one standby gas treatment subsystem for one hour and maintaining greater than or equal to 0.266 inches of vacuum water gauge in the secondary containment at a flow rate not exceeding.

4000 CFM.

"When irraciated fuel is being Pandled in the primary or secondary containment and during CORE ALTERATIONS (sc operations with a potential for draining the reae. tor vessel.

l GRAND GULF-UNIT 1 3/4 6-46 Amendment No. 9, --- I 1

TABLE 3.6.6.2-1 SECONDARYCONTAINMENTVENTILATIONSYSTEMAUTOMATICISOLATIONDAMPERS/V(LVES MAXIMUM DAMPER / VALVE FUNCTION ( $ 25;-)

& ISOLATION TIME (Seconds)

a. Dampers Auxiliary Building Ventilation Supply Damper (Q1T41F006)-C8) 4 Auxiliary Building Ventilation Supply Damper (Q1T41F007)-(4 ) 4 F'uel Handling Area Ventilation Exhaust Damper

, (Q1T42F003)-CB) 4 Fuel Handling Area Ventilation Exhaust Damper (Q1T42F004) -(4 ) 4 Fuel Handling Area Ventilation Supply Damper (Q1T42F011) -Ca) 4 Fuel Handling Area Ventilation Supply Damper (Q1T42F012) -(ta ) 4 !

, Fuel Pool Sweep Ventilation Supply Damper 3 i

(Q1T42F019) . (4 ) .- 4 o n

Fuel Pool Sweep Ventilation Supply Damper (Q1T42F020) - (. 8) 4 ;

Containment & Drywell Area Ventilation Supply Damper 4 (QlM41F007) -[3) 4 Co'ntainment & Drywell Area Ventilation Supply Damper 4 (Q1M41F008) -(4)

Containment & Drywell Area Ventilation Exhuast Damper 4 ;

(Q1M41F036)-(4 ) i Containment & Drywell Area Ventilation Exhaust Damper 4 (Q1M41F037) -(3)

J ,

The "-- ( A), - (8)' des 1ydoes on -/hc valve /da,,,per- f, !

numbers ihdicate associated e/ectrical divisions . ,

GRAND GULF-UNIT 1 3/4 6-48 Amendment No. 4, 7, Q, -

TABLE 3.6.6.2-1 (Continued)

SECONDARY CONTAINMENT VENTILATION SYSTEM AUTOMATIC ISOLATION OAMPERS/ VALVES MAXIMUM ISOLATION TIME VALVE FUNCTION (Li.ier) (Seconds)

b. Valves Plant Chilled Water System Aux. Bldg. Isol. Valve (P71-F306) -(A) 30 Plant Chilled Water System Aux. Bldg. Isol. Valve (P71-F304) - ( A) 30 Plant Chilled Water System Aux. Bldg. Isol. Valve

' (P71-F302) -( A) 4 Plant Chilled Water System Aux. Bldg. Isol. Valve (P71-F300) - (A) 4 Plant Chilled Water System Aux. Bldg. Isol. Valve (P71-F307) - (B) ,

30 Plant Chilled Water System Aux. Bldg. Isol. Valve (P71-F305) -(g) 30

~

Plant Chilled Water System Aux. Bldg. Isol. Valve (P71-F303) -(B) 4 Plant Chilled Water System Aux. Bldg. Isol. Valve (P71-F301) -(8) 4 Se'rvice Air System Aux. Bldg. Isol. Valve .

(P52-F221A) - (A) 4 Service Air System Aux. Bldg. Isol. Valve *

(P52-F160A) -(A) 4 Service Air System Aux. Bldg. Isol. Valve (PS2-F221B) - (B) , 4 Service Air System Aux. Bldg. Isol. Valve -

(P52-F160B) - (B) .

4 Instrument Air System Aux. Bldg. Isol. Valve ,-

4 ~-

(P53-F026A) -( A) ,

Instrument Air System Aux. Bldg. Isol. Valve (P53-F026B) - ( B) 4 l FPCC Filt-Demin System Backwash Aux. Bldg. 1501.

Valve (G46-F253) 30

( A 4 B) .

. GRAND GULF-UNIT 1 3/4 6-49 Amend,* cat do. l U . _ . -- . _ - . _ . - - - .-.

TABLE 3.6.6.2-1 (Continted)

SECONDARY CONTAINMENT VENTILATION SYSTEM AUTOMATIC ISOLATION'OAMPERS/ VALVES MAXIMUM ISOLATION TIME VALVE FUNCTION ( E d ;r) (Seconds)

Valves (Continued)

RWCU Backwash RCVG Tk. F.ex. Biog. Isol. Valve (G36-F108)-(A) 30 RWCU Backwash RCVG Tk. Aux. Bldg. ! sol. Valve (G36-F109) - ( B) 30 Nuclear Boiler System Aux. Bldg. Isol. Valve (B21-F113) - ( A) 30 Nuclear Boiler System Aux. Bldg. Isol. Valve (B21-F114) - ( 8) 30 RWCU Aux. Bldg. Isol. Valve (G3?-F235)-(A) 30 RWCU Aux. Bldg. Isol. Valve (C33 F234)-(B) 30 SPCU Aux. Bldg. Isol. Vaba (fen F003) -( A) 30 SPCU Aux. Bldg. Isol. Valse (P6C .rgo4; -(s) 30 to SPCU Aux. Bldg. Isol. Vaive (PSO-F007) -(s) 30 SPCU Aux. Bldg. Isol. Valve (PSG- F008) - ( A) 30 l

Fire Protection System t.ux. Bldg. Isol. Valve l (P64-F282A) - ( A) 4 l .

Fire Protection System Aux. Bldg. Isol. Valve (P64-F283A) - ( A) 4 Fire Protection System Aux. Bldg. Isol. Valve (P64-F332A) - ( A) 4

, Fire Protection System Aux. Bldg. Isol. Yalve

'(P64-F2828)-(8) '4 Fire Protection System Aux. Bldg. Isol. Valve (P64-F2838) - ( B) 4 Fire Protecti6n System Aux. Bldg. Isol. Valve (P64-F3328) - (8) 4 Cond. & Refuel Water Transfer Aux. Bldg. Isol. Valve

.9 (P11-F062) -( A)

I

~

GRANO GULF-Ul:Il 1 3/4 6-50 '

TABLE 3.6.6.2-1 (Continued)

SECONDARY CONTAINMENT VENTILATION SYSTEM AUTOMATIC ISOLATION DAMPERS / VALVES MAXIMUM .

. ISOLATION TIME VALVE FUNCTION @ Cic) (Seconds)

Valves (Continued)$

Cond. & Refuel Water Transfer Aux. Bldg. Isol. Valve (P11-F064)-(A) 4 Cond. & Refuel Water Transfer Aux. Bldg. Isol. Valve (P11-F066) cA) 4 Cond. & Refuel Water Transfer Aux. Bldg. Isol. Valve (P11-F047) -f 4) 4 Cond. & Refuel Water Transfer Aux. Bldg. Isol. Valve (P11-F063) -( 8 ) 4 Cond. & Refuel Water Transfer Aux. Bldg. Isol. Valve (P11-F065) -ca ) 4 Cond. & Refuel Water Transfer Aux. Bldg. Isol. Valve (P11-F067)-Ca) 4 ,

Cond. & Refuel Water Transfer Aux. Bldg. Isol. Valve a

(P11-F061) -( t) 4 o m

Floor and Equipment Drains System Aux. Bldg. Isol. Valve (P45-F158)- (4) 9 Floor and Equipment Drains System Aux. Bldg. Isol. Valve (P45-F160)- (4 ) -9

!. Floor and Equipment Drains System Aux. Bldg. Isol. Valve j (P45-F163)-(4,3 ) 9 Floor and Equipment Drains System Aux. Bldg. Isol. Valve 9

(P45-F159)-(3) i Floor and Equipment Drains System Aux. Bldg. Isol. Valve 9

(P45-F161) -(3 )

Makeup Water Treatment Sys. Aux. Bldg. Isol. Valve 30 (P21-F024) -(4)

Domestic Water System Aux. Bldg. Isol. Valve (P66-F029A)-(A*gj 4 PSW Aux. Bldg. Isol. Valve (P44-F121) -(A ) 100

- GRAND GULF-UNIT 1 3/4 6-51 Amendmed No.

6-TABLE 3.6.6.2-1 (Continued)

SECONDARY CONTAINMENT VENTILATION SYSTEM AUTOMATIC ISOLATION OAMPERS/ VALVES

_ MAXIMUM ISOLATION TIME VALVE FUNCTION ( ;-i,; r ) (Seconds)

Valves (Continued)

PSW Aux. Bldg. Isol. Valve (P44-F122)-(4) 100 PSW Aux. Bldg. Isol. Valve -

(P44-F117) -(4 ) 100 PSW Aux. Bldg. Isol. Valve (P44-F11B) -(4) 100 PSW Aux. Bldg. Isol. Valve .a

, (P44-F120) -(3 ) 100 0 PSW Aux. Bldg. Isol. Valve 100 (P44-F123)- CD)

PSW Aux. Bldg. Isol. Valve (P44-F116)-(s) 100 PSW Aux. Bldg. Isol. Valve (P44-F119)-(8) , 100 RHR "A" Loop Discharge To Liquid Radwaste Valve 30 (E12-F203).g) i l

l l

l GRAND GULF-UNIT 1 3/4 6-52 Amendmed No. I

CONTAINMENT SYSTEMS 3/4.6.7 ATMOSPHERE CONTROL CONTAINMENT ANC CRiWELL HYOROGEN RECOMBINER SYSTEMS jo LIMITING CONDITION FOR OPERATION 3.6.7.1 Two independent containment and dry-ell hydrogen recomoiner systems lk ,

shall be OPERABLE.

l APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.

ACTION:

'I t With one containment ;nd dryu;il h'ydrogen recombiner system inoperable, restore lk the inoperable system to OPERABLE status within 30 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 . F SURVEILLANCE REQUIREMENTS o

4.6.7.1 Each containment and d ya:11 hydrogen recomoiner system shall be lE cemonstratec OPERAELE:

a. At least once per 6 months by verifying during a recombiner system functional test that the minimum heater sheath temperature increases !

to greater than or equal to 700 F within 90 1minutes. Maintain >700 F for at least 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . - L- >-

b. At least once per 18 months by:

1. Performing a CHANNEL CALIBRATION of all control room recomsiner instrumentation and control circuits.

2. Verifying the integrity of all heater electrical circuits my performing a resistance to ground test within 30 minutes following the above required functional test. The resistance to grounc for any heater phase shall be greater than or equal to 10,000 ohms.

3. Verifying during a recomciner system functional test that the heater sheath temoerature increases to greater than or ecual to 1200*F within 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months and is maintained between 1150*F anc 1300*F for at least 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

4. Verifying through a visual examination that there is no evidence of abnormal conditions within the recombiner enclosure; i.e, loose wiring or structural connections, deposits of foreign i materials, etc.

c. [0ELETED)

GRAND GULF-UNIT 1 3/4 6-55 Amendment No. 7, l

~

/

CONTAINMENT SYSTEMS I CO AINMENT AND DRYWELL HYDROGEN IGNITION SYSTEM i LIMITIN CONDITION FOR OPERATION i

3.6.7.2 Two ndependent containment and drywell hydrogen i ition system sub -

systems shall e.0PERABLE. .

APPLICABILITY: 0 ERATIONAL CONDITIONS 1 and 2.

ACTION: ,

With one containment an drywell hydrogen ignitio subsystem inoperable, ;

restore the inoperable s system to OPERABLE st us within 7 days or be in at least HOT SHUTDOWN with' the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SURVEILLANCE REGUIREMENTS s e 4.6.7.2 Each containment and drywe ydrogen ignitign subsystem shall,.be e-sa ,

demonstrated OPERABLE:

O

a. At least once per 92 ays by ene izing the supply breakers and .

verifying that at 1 ast 41 glow p gs are energized.

b. At least once pe 18 months by: i

1. Verifyin the cleanliness of each ow plug by a visual -

inspec on.

2. Ener izing each glow plug and verifying surface temperature of at least 1700*F.

i l

GRAND GULF-UNIT 1 3/4 6-57 Amendment No. 7

CONTAINMENT SYSTEMS CONTAINMENT AND DRYWELL HYDROGEN IGNITION SYSTEM LIMITING CONDITION FOR OPERATION 3.6.7.2 The containment and drywell hydrogen ignition system consisting of the following:

a. At least two igniter assemblies in each enclosed area specified on Table 3.6.7.2-2,

b. All igniter assemblies adjacent to any inoperable igniter assembly in each open area specified on Table 3.6.7.2-2, and

c. Two independent containment and drywell hydrogen ignition subsystems each consisting of two circuits (as listed on Table 3.6.7.2-1) with no more than two igniter assemblies inoperable per circuit, shall be OPERABLE.

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2, ACTION:

a. With less than two igniter assemblies OPERABLE in any enclosed area specified in Table 3.6.7.2-2, restore at least two igniter assemblies 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 .

b. With any adjacent igniter assemblies within an open area as specified on Table 3.6.7.2-2 inoperable restore the igniter assemblies in that open area so that all igniter assemblies adjacent to an inoperable igniter []

assembly are OPERABLE 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 .

c. With one containment and drywell hydrogen igniter 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 .

SURVEILLANCE REQUIREMENTS 4.6.7.2 The containment and drywell hydrogen ignition system shall be demonstrated OPERABLE:

a. At least once per 92 days by energizing the supply breakers and:

1. Verifying a visible glow from the glow plug tip of each normally accessible igniter assembly specified in Table 3.6.7.2-2,

2. Verifying that each circuit of each containment and drywell hydrogen i igniter subsystem is conducting sufficient current to energize the minimum required number of igniter assemblies specified on Table 4.6.7.2-1.

GRAND GULF -UNIT 1 3/4 6-57 Amendment No.

Z69abm1

b. At every COLD SHUTDOWN, but no more frequently than once per 92 days, by energizing the supply breakers and verifying a visible glow from the glow plug tip of each normally inaccessible igniter assembly specified in Table 3.6.7.2-2.

c. At least once per 18 months by:

1. Verifying the cleanliness of each glow plug by a visual inspection.

2. Energizing each glow plug and verifying a surf ace temperature of at least 1700*F.

o-4 0

GRAND GULF - UNIT 1 3/4 6-57a Amendment No.

Z69abm2

Table 3.6.7.2-1 Hydrogen Igniter Circuits Division I Division II Circuit 2 Circuit 1 Circuit 2 Circuit 1 D124 D107 D125 D106 D109 D127 D108 D126 D111 D129 D110 D128 D112 D136 D113 D130 D114 D138 D115 D132 D116 D140 D117 D134 D119 D149 D118 D137 D121 D151 D120 D139 D123 D153 D122 D141 D148 D161 D131 D143 D150 D162 D133 f D145 I D152 D165 D135 D147 D154 D166 D142 r D155 D159 D168 D144 D157 D160 D170 D146 l D172 '

D163 D171 D156 D174 D164 D178 D158 D176 D167 D180 D173 -

D183 I D169 D182 D175 q_

D192 D179 D187 D177 \D D185 D189 D184 l i D186 D181 D191 D193 O l D188 D190 D194 D195 l t

! L i

l i

l l

i 1

3/4 6-57b Amendment No.

GRAND GULF - UNIT 1 t Z69 abs 3 i

Table 3.6.7.2-2 Hydrogen Igniters and Locations Dist. From Center Line

Igniter Div./ Circuit Elevation Azimuth of Reactor NORMALLY ACCESSIBLE Open Areas Containment D124 I/1 136'-0" 21 57'-0" D125 11/1 132'-10" 47 53'-0" D126 I/1 132'-10" 75 51'-9" D127 11/1 132'-10" 107 51'-9" D128 I/1 132'-10" 135 51'-9" D129 II/1 132'-10" 165 51'-9" D130 I/1 132'-10" 195 51'-9" D131 11/2 145'-7" 220 60'-0" D132 I/1 134'-4" 253 51'-9" D133 11/2 134'-4" 285 51'-9" D134 I/1 134'-4" 317 52'-8" D135 II/2 136'-0" 349 51'-9" D137 1/1 160'-4" 36 53'-6" D138 11/1 157'-10" 70 51'-9" D139 1/1 157'-10" 100 51'-9" D140 11/1 160'-4" 135 51'-2" CF 51'-9" UD D141 I/1 155'-10" 164 C)

D142 II/2 155'-10" 196 51'-9" D143 I/1 165'-0" 226 61'-4" D144 II/2 160'-4" 260 54'-2" D145 I/1 159'-4" 285 51'-5" D146 11/2 159'-4" 321 51'-5" D148 I/2 182'-9" 30 61'-0" D149 11/1 167'-8" 41 42'-0" D154 I/2 182'-4" 136 51'-9" l D155 I/1 182'-4" 254 55'-9" l

D156 II/2 183'-4" 274 48'-0" D157 I/1 182'-4" 293 58'-11" D158 11/2 183'-4" 320 53'-2" D160 1/2 202'-0" 35 46'-0" D161 - 11/1 207'-9" 59 44'-0" l

D170 11/1 207'-7" 135 55'-8" D171 II/1 206'-0" 216 46'-9" E172 I/1 204'-11" 252 26'-0" D173 II/2 204'-4" 256 53'-8" l D174 I/1 204'-11" 284 53'-8" D175 11/2 201'-11" 298 26'-8" l 56'-6" D176 I/1 207'-9" 310 GRAND GULF - UNIT 1 3/4 6-57c Amendment No.

l l Z69abm4

Dist. From Center Line

Igniter Div./ Circuit Elevation Azimuth of Reactor Open Areas Containment (Continued)

D178 11/1 262'-0" 6 55'-5" D179 I/2 262'-0" 48 55'-5" D180 11/1 262'-0" 91 55'-0" D181 1/2 262'-0" 140 55'-0" D182 II/1 262'-0" 183 55'-0" D183 I/1 262'-0" 225 55'-0" D184 11/2 262'-0" 268 55'-0" D185 1/1 262'-0" 333 55'-0" D186 I/1 283'-10" 349 39'-9" D187 11/1 283'-10" 34 39'-9" D188 I/2 283'-10" 81 39'-9" D189 11/1 283'-10" 127 39'-9" D190 1/2 283'-10" 152 39'-9" D191 11/1 283'-10" 199 39'-9" D192 I/1 283'-10" 242 39'-9" D193 11/2 283'-10" 286 39'-9" D194 11/2 295'-0" 349 15'-3" D195 1/2 295'-0" 158 15'-3" as S3 NORMALLY INACCESSBILE

[

Open Areas Drywell D106 11/2 146'-3" 0 26'-6" D107 1/2 145'-7" 63 29'-3" D108 11/2 146'-2" 120 29'-8" D109 I/2 147'-1" 180 26'-3" D110 11/2 145'-7" 240 29'-2" D111 I/2 145'-7" 313 25'-1 1/4" D112 I/2 160'-6" 0 27'-4" D113 II/2 160'-6" 60 29'-9" D114 I/2 160'-6" 135 27'-1" D115 II/2 160'-6" 180 26'-10" D116 I/2 160'-6" 232 26'-1" l D117- 11/2 160'-6" 324 26'-5"

! D118 11/2 179'-0" 0 26'-4" l

D119 I/2 179'-0" 65 26'-4" D120 11/2 179'-0" 125 26'-4" D121 .

1/2 179'-0" 180 26'-4" D122 11/2 179'-0" 245 26'-4" D123 1/2 179'-0" 305 26'-4" GRAND GULF - UNIT 1 3/4 6-57d Amendment No. l i

Z69abm3

Dist. From Center Line

Igniter Div./ Circuit Elevation Azimuth of Reactor NORMALLY INACCESSIBLE (Continued)

Enclosed Areas Main Steam Tunnel D136 11/1 166'-0" 16 51'-9" D147 I/1 166'-0" 344 51'-9" RWCU Backwash Room D150 1/2 168'-10" 70 42'-0" D151 11/1 168'-10" 105 42'-0" D152 1/2 178'-10" 70 46'-2" D153 11/1 178'-10" 109 51'-5" RWCU Heat Exchanger Room D159 I/2 202'-0" 21 50'-4" n, D177 II/2 202'-0" 341 55'-0" s3 Q

Filter /Demineralizer Room D162 11/1 202'-0" 74 55'-8" D163 1/2 202'-0" 88 48'-0" D164 I/2 202'-0" 92 48'-0" D165 11/1 202'-0" 106 55'-8" RWCU Pump Area D166 11/1 202'-0" 93 45'-0" D167 I/2 202'-0" 86 37'-6" l

RWCU Sample Area D168 II/1 202'-0" 86 34'-0" D169 I/2 202'-0" 96 22'-6"

!

System Prefix is E61 for all igniters I

3/4 6-57e Amendment No. l GRAND GULF - UNIT 1 269abm6

i Table 4.6.7.2-1 i

NUMBER OF IGNITERS BY CIRCUIT Division I Minimum Required Total on Circuit !

Circuit 1 19 21 Circuit 2 22 24 >

Division 11 Circuit 1 20 22 Circuit 2 21 23

?

i e-M O

t I

I l l -

I i.

t' s

GRAND GULF - UNIT 1 3/4 6-57f Amendment No. l l

Z69abm7 ,

l

CONTAINMENT SYSTEMS DRYWELL PURGE SYSTEM LIMITING CONDITION FOR OPERATION 3.6.7.3 Two independent drywell purge system subsystems shall be OPERABLE.

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2. ,

t ACTION:

P With one drywell purge system subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 30 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 . ,

SURVEILLANCE REQUIREMENTS C;ntin;;d

, i 4.6.7.3 Each drywell purge system subsystem shall be demonstrated DPERABLE: l l

a. At least once per 92 days by:

1. Starting the. subsystem from the control room, and i l

2. Verifying that the system operates for at least 15 minutes.

b. . At least once per 18 months by:

1. Verifying a subsystem flow rate of at least 1000 cfm during subsystem operation for at least 15 minutes.

2. Verifying the pressure differential required to open the vacuum .

l breakers on the drywell purge compressor discharge lines, from the closed position, to be less than or equal to 1.0 psid. I

c. Verifying the OPERABILITY of the drywell purge compressor discharge line vacuum breaker isolation valve differential pressure actuation instrumentation with an opening setpoint of 0.0 to 1.0 psid (Drywell minus Containment) by performance of a: ,

1. CHANNEL CHECK at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ,

2. 'CHANNIL FUNCTIONAL TEST at least once per 31 days, and i

3. CHANNEL CALIBRATION at least once per 18 months.

l t t

l GRAND GULF-UNIT 1 3/4 6-58 Amendment No. 8, j

-.n,__,,,,,,n-_- -

, , _ - , ,,n----,--+_, . , . ,, v., , ,, , - - , . _ , - - - , , . - , . ,

a = .m - ~

s 3.4.6 CONTAINMENT SYSTEMS BASES 3r/4.6.1 PRIMARY CONTAINMENT 3/4.6.1.1 PRIMARY CONTAINMENT INTEGRITY

. PRIMARY CONTAINMENT INTEGRITY ensures that the release of radioactive materials from the containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the accident analyses. This restric-tion,inconjunctionwiththeleakageratelimitation,willlimitthesite boundary radiation doses to within the limits of 10 CFR Part 100 during accident conditions.

3/4.6.1.2 CONTAINMENT LEAKAGE The limitations on containment leakage rates ensure that the total containment leakage volume will not exceed the value assumed in the accident analyses at the. peak accident pressure of 11.5 psig, P . As an added conser-vatism,themeasuredoverallintegratedleakagerateilfurtherlimitedtoless than or equal to 0.75 L a during performance of the periodic tests to account for possible degradation of the containment leakage barriers between leakage tests.

Operating experience with the main steam line isolation valves has indicated that degradation has occasionally occurred in the leak tightness of the valves; therefore the special requirement for testing these valves.

The surveillance testing for measuring leakage rates is consistent with exemptionthe(s req)uirements of Appendix J to 10 CFR 50 with the exc the airlocks after each opening.

3/4.6.1.3 CONTAINMENT AIR LOCKS The limitations on closure and leak rate for the containment air locks are required to meet the restrictions on PRIMARY CONTAINMENT INTEGRITY The and the l

containment leakage rate given in Specifications 3.6.1.1 and 3.6.1.2.

specification makes allowances for the fact that there may be long periods of

} time when the air locks will be in a closed and secured position during reactor operation. Only one closed door in each air lock is required to maintain the integrity of the containment. q g g 3/4.1.1.4 MSIV LEAKAGE CONTROL SYSTEM Calculated doses resulting from the maximum leakage allowance for the main steamline isolation valves in the postulated LOCA situations would be a small fraction of the 10 CFR 100 guidelines, provided the main steam line system from the isolation valves up to and including the turbine condenser remains intact.

Operating experience has indicated that degradation has occasionally occurred in the leak tightness of the MSIV's such that the specified leakage requirements have not always been maintained continuously. The requirement for the leakage control system will reduce the untreated leakage from the MSIVs when isolation of the primary system and containment is required.

GRAND GULF-UNIT 1 B 3/4 6-1

Insert to Bases 3/4.6.1.3, Page B3/4 6-1 Verification that each air lock door inflatable seal system is OPERABLE by the performance of a local leak-detection test for a period of less than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months is permissible if it can be demonstrated that the leakage rate can be i accurately determined for this shorter period. This is in accordance with Section 6.4 and 7.6 of ANSI N45.4-1972.

L

[

l I

i

.I ZZ2sd1

CONTAINMENT SYSTEMS BASES CONTAINMENT PURGE SYSTEM (Continued)

Leakage integrity tests with a maximum allowable leakage. rite for purge supply and exhaust isolation valves will provide early indication of resilient eaterial seal degradation and will allow the opportunity for repair before gross leakage failures dev.elop. The 0.60 L leaking limit shall not be l exceededwhentheleakageratesdetermined$ytheleakageintegritytestsof these valves are added to the previously determined total for all valves and penetrations subject to Type B and C tests.

3/4.6.2 ORWELL 3/4.6.2.1 ORWELL INTEGRITY Drywell integrity ensures that the steam released for the full spectrum of drywell pipe breaks is condensed inside the primary containment either by

the suppression pool or by containment spray. By utilizing the suppression i pool as a heat sink, energy released to the containment is minimized- ' and the !

severity of the transient is reduced.

Imert A 3/4.6.2.2 DRWELL BYPASS LEAKAGE ures that the maxinum leakage The limitation on drywell bypass leakage rate e which could bypass the suppression sool during an acc1 ent would not rescit in the The irtegrated drywell r

containment exceeding its design prassure of 15.0 psig.

"1l cap -- r + / rase t B l N

Ls.stc.akage value is limited to 10%mw4d le of the d: f n dr- iAk.ge.e:ks&tg, 4 C

\ The limiting case accident is a very . mall reactor coolant The system break which long term di?-

will not automatically result in a reactor depressurization.

farential pressure created between the drywell and containment will result in a significant pressure buildup in the containment due to this bypass leakage. ,

l 3/4.6.2.3 DRWELL AIR LOCKS The limitations on closure for the drywell air locks are recuired to-meet ,'

the restrictions on ORWELL INTEGRITY and the drywell leakage rate giver in The specification makes allowances for Specifications 3.6.2.1 and 3.6.2.2.the fact thtt there Only mayone be closed long periods 4cor of time a closed and secured position during reactor operation.

y in each air lock is required to maintain the integrity of the dryweli gEHssnr m 3/4.6.2.4 DRWELL STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the drywell will -

be maintained comparable to the original design specification for the life of !

the unit. A visual inspection in conjunction with Type A leakage tests is r sufficient to demonstrate this capability. ,

l 3/4.6.2.'5 ORWELL INTERNAL PRESSURE . r The limitations on drywell-to-containment differential pressure ensure that the drywell peak pressure of 22.0 psig dees not exceed the design pressure of 30.0 psig and that the containment peak pressure of 11.5 psig does not exceed the design pressure of 15.0 psig during LOCA conditions. The eaximum external ;

drywell pressure differential is limited to +1.0 psid, we11'below the 2.3 psid '

at which suppression pool water will be forced over the wier wall and into the drywell. The limit of 2.0 psid for initial poi,itive drywell to containment _

drp:11 pre;;ur; tc 22.0 p;ig eich i: _ .. . 127;,

=d is design pressurewilljimitth: consistent with the safety analysis.

pr:sem ,

8 3/4 6-3 Amendment No. - 1 ,

GRANO GULF-UNIT 1 not allow cleadn3 of the top vent wMd

_.J

Insert to Bases 3/4.6.2.2, Page B3/4 6-3 Insert "A" ThedesigndrywellleakagerateisexpressedasA/[liandhasavalueof 0.90 fta, gffkisdependentonlyonthegeometryofdrywellleakage 2 k is a lumped paths where A = flow area of leakage paths in ft and constant which considers geometric and friction loss coefficients such as discontinuities and Reynolds numbers. At a 3 psi differential pressure fromdrywelltocontainmentanA/Ykhasanequivalentmassflowof 35,000 scfm.

Insert "B" which is equivalent to 3500 scfm at 3 psid drywell to containment.

Insert "C" The A/ [k'value of 0.90 f t is derived from the analysis of " bypass 2

capability with containment spray and heat sinks". (FSAR 6.2.1.1.5.5)

I l

{

\

ZZ51a/abm7 f

i Insert to Bases 3/4.6.2.3, Page B3/4 6-3 Verification that each air lock door inflatable seal system is OPERABLE by the performance of a local leak-detection test for a period of less than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months is permissible if it csn be demonstrated that the leakage rate can be

curately determined for this shorter period. This is in accordance with Section 6.4 and 7.6 of ANSI N45.4-1972.

4 l

h 5

ZZ2sd2

- - , ~ . - __

_ . . . .~_

CONTAINMENT SYSTEMS BASES 3/4.6.2.6 DRYWELL AVERAGE AIR TEMPERATURE o

The limitation on drywell average air temperature ensures that peak drywell temperature does not. exceed the design temperature of 330 F during LOCA conditions and is consistent with the safety analysis.

3/4.6.3 DEPRESEURIZATION SYSTEMS T'he specifications of this section ensure that the drywell and containment pressure will not exceed the design pressure of 30 psig and 15 psig, respectively, during primary system blowdown from full operating pressure. ,ogo g g ,,

The suppression pool water volume must absorb the associated d d o N

structural sensible heat released during a reactor blowdown from 10.. p 4 "

Using conservative parameter inputs, the maximum calculated containment pressure during and following a design basis accident is below the containment design pressure of 15 psig. Similarly the drywell pressure remains below the design pressure of 30 ps19 The maximum and minimum water volumes for the suppression pool are 138,851 cubic feet and 136,146 cutiic feet, respectively.

These values include the water volume of the containment pool, horizontal .mur A g and weir annulus.# Testing in the Mark III Pressure Suppression Test w N

vents,ty Facili and analysis have assured that the suppression pool temperature will not rise above 185 F for the full range of break sizes.

Should it be necessary to make the suppression pool inoperable, this shall only be done as specified in Specification 3.5.3.

Experimental data indicates that effective steam condensation without excessive loads on the contai'n' ment pool walls will occur with a quencher device and pool temperature below 200 F during relief valve operation.

Specifications have been placed on the envelope of reactor operating condi-tions to assure the bulk pool temperature does not rise above 185 F in compliance with the containment structural design criteria.

In addition to the limits on temperature of the suppression pool water, operating procedures define t% cction to be taken in the event a safety-relief valve inadvertently opens or su cks open. As a minimum this action shall include: (1) use of all available means to close the valve, (2)' initiate suppression pool water cooling, (3) initiate reactor shutdown, and (4) if other safety-relief valves are used to depressurize the reactor, their discharge shall be separated from that of the stuck-open safety relief valve to assure mixing and uniformity of energy insertion to the pool.

The containment spray system consists of two 100". capacity trains, each with three spray rings located at different elevations about the inside circumference of the containment. RHR A qump supplies one train and RHR pump ,g B supplies the other. OliR pump C cannot supply the containment spray syste Dispersion of the flow of water is effected by 350 nozzles in each train, g enhancing the condensation,of water vapor in the containment volume and - Insert c b -

preventing overpressurization. fHeat rejection is through the RHR heat exchangers. The turbulence caused by the spray system aids in mixing the containment air volume to maintain a homogeneous mixture for H2 control.

The suppression pool cooling function is a mode of the RHR system and functions as part of the containment heat removal system.' The purpose of the system is to ensure containment integrity following a LOCA by preventing

~

G ND GULF-UNIT 1 B 3/4 6-4 l Admemdment Mo.

Insert A to Bases 3/4.6.3, Page B 3/4 6-4 and have corresponding pool water depths of 18'4-1/12" and 18' 9-3/4" respectively.

The minimum suppressio.n pool volume of 135,2918 ft is based on satisfying the Mark 111 suppression pool sizing criteria established 8in a GE Design Review File (DRF T23-0408). The 135,291 ft8 and 138,701 ft pool water volumes were used in the Grand Gulf containment analysis (GE DRF 699-0010) to verify their adequacy to perform all required functions following the design basis Limiting Condition of a main steam line break from 105% of full power. (FSAR 6.2.1.1.3.3.3)

The 18'4-1/12" and 18' 9-3/4" pool water depths are nominal values derived analytically, considering pool geometry, from the above pool volumes which were used in the pool design analyses. The pool levels (depth) satisfy criteria or constraints imposed by: (1) 2-foot minimum post-LOCA horizontal vent coverage to assure steam condensation / pressure suppression, (2) adequate ECCS pump NPSH, (3) adequate depth for vortex prevention, (4) adequate depth for minimum recirculation volume, (5) adequate weir-wall free-board for inadvertent upper pool dump and (6) to limit hydrodynamic loads on submerged structures during SRV and vent steam discharges.

The suppression pool temperatures are based as follows:

95*F Is the initial condition for the analysis determining pool volume adequacy and satisfies the post-LOCA long term peak pool temperature of 185'F.

120*F 1s analytically based and is derived to satisfy the 170*F post-blowdown peak pool temperature assuming a LOCA when the reactor is isolated.

105'F Are derived from the analytically bas'd 95'F and 120*F using and engineering judgement considering operator response time, reactor 110*F pressure vessel energy and pool heat capacity to meet the 170*F limit and also to avoid unnecessary scrams and/or depressurizations.

Z59sd3

> t i.

L Insert B to Bases Section 3/4.6.3, Page B3/4 6-4 .

i The surveillance requirements, which include system losses for surveillance !

I testing, provide adequate assurance that the containment spray system will be OPERABLE when required.

I l

i 1

i-k r

I, t

P A

I A

i i

b 4

ZZ51a/abm8 ,

-~q 3 g-e-n---, w , mp,m ,,,~-,,-w - e --w.,- - . , . - - ,s--.---- -cm- ,---m-g-, ,.g--n- -- --, g,-re---..-. - - - - -w av-.,s.- ,m.gy--., , . , -.

I Insert C to Bases 3/4.6.3, Page B 3/4 6-4 Whenever maintenance activities are performed that could result in nozzle obstruction, a surveillance will be performed to verify that flow through the nozzles is unobstructed.

i t

I t

ZZ51a/abm5 f

r-y-rWrr< wrww-emvw *r vyW'yr g- wwww w-vWwry-y y---e- ww -wwys'-rw T s wyy-'ww---- e'g-eeyyp*w6qvi-,g-w39 m - . _ 7 m,rN= - - - -------yn---+----M-=+---y++-+-

CONTAfNMENT SYSTEMS BASES DEPRESSURIZATION SYSTEMS (Continued) excessive containment pressures and temperatures. '

The suppression pool cooling mode is designed to limit the long term bulk temperature of the pool to 185'F considering all of the post-LOCA energy additions. The suppression pool cooling ,

trains, being an i.ntegral part of the RHR system, are redundant, safety-related component systems that are initiated. fo' lowing tne recovery of the reactor vessel water level by ECCS flows from the RHR system. Heat rejection to the standby service water is accomplished in the RHR heat exchangers.

~

The suppression pool make up system provides water from the upper contain-ment pool to the suppression pool by gravity flow through two 100% capacity dump 1.ines following a LOCA. The quantity of water prosided is sufficient to account for all conceivable post accident entrapment vol mes, ensuring the long term energy sink capabilities of the suppression pool anc maintaining the water coverage over the uppermost drywell vents.

The minimum freeboard distance above the suppression pool high water level to the top of the weit wall is adequate

.te preclude flooding of the drywell in the event of an inadvertent dump. During i refueling, neither automatic nor manuti a: tion can open the make-up dump valves.

3/4.6.4 CONTAINMENT AND DRYWELL ISOLATION VALVES The OPERABILITY of the containment isolation valves ensures . hat the con- 1 tainment atmosphere will be isolated from the outside environment in the event, '

- - ' of a release of radioactive material to the containneat ateosphe're or pres-surization of the containment and is consistent with :ne requirefnents of GDC 54 through 57 of Appendix'A to 10 CFR Part 50. Containmert isolation within the

?

time limits specified for those isolation valves designed to close automatically ensures that the release of radioactiw material tc the environment will be consistent with the assumptions used in the analyses for a LOCA.

The operability of the drywell isolation valves enseres that the drywell atmosphere will be directed to the suppre:,sion pool for the full spectrum of pipe breaks inside the drywell. Since the allowable value of drywell leakage is so las ge, individual drywell penetration leakage is not measured. By checking valve operability on any penetration which c;uld contribute a large fraction of the design leakage, the total leakage is miintained at less than the design value.

i"o het --*The maximum isolation times foi containmer.: and drywell automatic isolation I' valves tical are thetimes.

closing times used in the FSAR accident analysis for valves with analy-For automatic isolation valves not having analytical closing times, closing times are derived by applying margins to previous valve closing test data obtained by using ASME Section X3 criteria. Maximum closing times for these valves was determined by using a factor of no times the allow-able (from previous test closure to ne)t test closurz) ASME Section XI margin and adding this to the previous test closure time.

_3/4.6.5 DRYWELL POST-LOCA VACUUM BREAKERS The post-LOCA drywell vacuum breaker system is provided to relieve the vacuum in the drywell due to steam condansaLion following bioe down. Contain,-

ment air is drawn through the vacuum breaker che:k v.-Aven in the two branches of the separate post-LOCA vacuum relief line and in e braach c:f each drywell purge compressor discharge line. Vacuum relief initiain Jt a riifferential pressure of one psi. This vacuum relief, in conjunction with the rest of the GRAND GULF-UNIT 1 B 3/4 6-5 Amendment No. 9, l 6

Insert to Bases 3/4.6.4, Page B 3/4 6-5 l

l i t 1 Table 3.6.4-1 lists the Containment and Drywell Isolation Valves in four sections. Section 1 contains the Automatic Isolation Valves which are those valves that receive an automatic isolation signal from Table 3.3.2-1 instrumentation and are located on the Containment or Drywell penetrations.

The valves included in Section 2 are Manual Isolation Valves which receive a remote manual signal from a handswitch and are located on the Containment or Drywell Penetrations. Some of the valves in Section 2 may receive automatic signals, but not automatic isolation signals from instrumentation in Table 3.3.2-1. The valves included in Section 3 are those which do not receive isolation signals from instrumentation listed in Table 3.3.2-1 and do not utilize a remote manual handswitch. Section 5 includes check valves, local manual operated valves and power operated valves that do not utilize a handswitch. Section 4 of Table 3.6.4-1 contains test connection valves.

ZZ51a/abm3

CONTAfNMENT SYSTEMS BASES l -

DRYWELL POST-LOCA VACUUM.BREAXERS (Continued) drywell purge system, is necessary to insure that the post-LOCA drywell2 H concentration does not exceed 4% by volume.

forcing noncondensibles through the horizontal vents an at a rate designed to maintain the H2 concentraticn below the flammable limits.

There are two 100% vacuum relief systems so that the plant may continue .

operation with one system out of service for a limited period of time. !

3/4.6.6 SECONDARY CONTAINMENT Secondary containment is designed to minimize any ground level release of radioactive material which may result from an accident.

and Enclosure Building provide secondary containment during normal operation! "

when SS the containment is sealed and in service. When the reactor is in COLD

'an,qTDOWN or REFUELING, the containment may be open and the Auxiliary Build; c, Enclosure Building then become the only containment.

The maximum isolation times for secondary containment automatic isolation !

dampers / valves c e the times used in the FSAR accident analysis for dampers /

valves with analytical closing times. l For automatic isolation valves not having analytical closing times, closing times are derived by applying margins '

to previous valve closing test data obtained by using ASME Secti'dn XI criteria. ,

Maximum closing times for these valves was determined by using a' factor of two '

times the allowable ( yom previous test closure to next test closure) ASME Section XI margin and adding this to the previous test closure time.

l Establishing and maintaining a vacuum in the Auxiliary Building and i

Enclosure Building wita the standby gas treatment system once per 18 months,

along with the surveillance of the doors, latches, dampers,end-valves,ais ade- lgg I quate to ensure that there are no violations of the integrity lof the secondary containment.

blind -Floaps aaa entus dins, i The OPERABILITY of the standby gas treatment systems ensures that l sufficient iodine removal capability will be available in the event of a LOCA.

The reduction in containment ioding . inventory reduces the resulting site boundary radiation doses associated with containment leakage. The operation of this system and resultant iodine remova.1 capacity are consistent with the assumptions used in the LOCA analyses. Cumulative operation of the system with the heaters OPERABLE for 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months over a 31 day period is sufficient to l

reduce the buildup of moisture on the absorbers and HEPA filters.

~

{

The surveillance testing ior verifying heat dissipation for the Standby ;

Gas Treatment System heaters is performed in accordance with ANSI N510-1975 !

with the exception of the 5% current phase balance criteria of Section 14.2.3. !

The offsite power system for the Grand Gulf Nuclear Station consists of a non-transpositional 500 KV grid. The grid has an inherent unbalanced load distri '

bution which results in unbalanced voltages in the plant. Voltage unbalances exceeding the ANSI N510-1975 5% criteria are not atypical. -

l o

GRAND GULF-UNIT 1 B 3/4 6-6 AmendmentNo.9,-l l

5.0 DESIGN FEATURES 5.1 SITE

.s EXCLUSION AREA 5.' 1.1 The exclusion area shall be as shown in Figure 5.1.1-1.

LOW POPULATION ZONE 5.1.2 The low population zone shall be as shown in Figure 5.1.2-1.

UNRESTRICTED AREA BOUNDARY FOR GASEOUS EFFLUENTS AND FOR LIOUID EFFLUENTS 5.1.3 The unrestricted area boundary for gaseous effluents and for liquid effluents shall be as shown in Figure 5.1.3-1. The gaseous effluent release points are shown in Figure 5.1.1-1.

5.2 CONTAINMENT 1

CONFIGURATION 5.2.1 The containment is a steel lined, reinforced concrete structure composed of a vertical right cylinder and a hemispherical dome. Inside and at the bottom of the containment is a reinforced concrete drywell composed of a vertical right cylinder and a steel head which contains an approximately eighteen to nineteen foot deep water filled suppression pool connected to the drywell through a series of horizontal vents. The containment has a minimum net free air volume of 1,400,000 cubic feet. The drywell has a minimum net free air volume of 270,000 cubic feet.

DESIGN TEMPERATURE AND PRESSURE 5.2.2 The containment and drywell are designed and shall be maintained for:

l a. Maximum internal pressure:

l 1. Drywell 30 psig.

2. Containment 15 psig.

b. Maximum internal temperature:

1 1. Drywell 330*F.

j 2. Suppression pool 185'F.

' c. Maximum external-to-internal differential pressure:

1. Drywell 21 psid.

2. Containment 3 psid.

SECONDARY CONTAINMENT Auxifrary 5.2.3 Thesecondarycontainmentconsistsofthe-RcactorBuildingandthel$ "

Enclosure Building, and has a minimum free volume of 3,640,000 cubic feet.

l l

ATTACHMENT 3 PROPOSED CHANGES TO THE GRAND GULF NUCLEAR STATION TECHNICAL SPECIFICATIONS NRC TECHNICAL REVIEW BRANCH: MATERIALS ENGINEERING 258mm1 L

Attachm:nt 3/Msterials Engineering AECM-84/0330 (6/17/84)

Page 2 Listing of Item Numbers by Technical Specification Problem Sheet (TSPS) Number TSPS No. Item Nos.*

160 3.A.01 219 3.D.01 319 3.B.01 4

Item number format: 1.A.02 Item number within category Category designator Attachment number Z58mm2 L

Attachmint 3/ Materials Enginrering AECM-84/0530 (6/17/84)

Page 3 A. TYP0 GRAPHICAL ERRORS , EDITORIAL CHANGES, AND CLARIFICATIONS This proposed change corrects obvious typographical errors, implements editorial changes such as correction of spelling errors, punctuation errors, and grammatical errors or provides clarification of the basic meaning or intent of the subject technical specification.

MP&L has determined that the proposed change does not; o Involve a significant increase in the probability or consequences of an accident previously evaluated; or o Create the possibility of a new or different kind of accident from any accident previously evaluated; or o Involve a significant reduction in a margin of safety.

Therefore, the proposed change does not involve a significant hazards consideration.

A description of this change including necessary justification for the change is provided below:

CLARIFICATIONS A clarification to the technical specifications to improve understanding and readability is discussed below:

1. (TSPS 160), Reactor Coolant Pressure / Temperature Limits, Technical Specification 3/4.4.6, Bases 3/4.4.6 The proposed change to this specification is to clarify that reactor vessel metal temperature rather than reactor coolant system tempera-ture is to be used in conjunction with Figure 3.4.6.1-1. This will make the specification consistent with the figure which is based on vessel metal temperature. The specification is further revised to refer specifically to reactor coolant temperature for the maximum heatup/cooldown rates. This is conservative, since the coolant temperature will change more rapidly than the vessel metal tempera-ture. Additionally, it is proposed to revise Surveillance Require-ment 4.4.6.1.3 to indicate that reactor vessel material specimens will be removed and examined in accordance with Appendix H of 10 CFR 50 and to renumber Surveillance Requirements 4.4.6.1.3 and 4.4.6.1.4 so that requirements dealing with temperature measurement and requirements dealing with materials testing are grouped together appropriately. This proposed change also adds Specification 4.4.6.1.5 to require removal and examination of the reactor flux wire specimens during the first refueling outage for determination of pressure vessel fluence as a function of time and power level.

This is consistent with current industry practice, since flux wires contain fewer impurities than material specimens and therefore provide a more accurate indication of neutron fluence. Also, for further clarification, Figure 3.4.6.1-1 is revised to indicate that Z58mm3 L

Attachmsnt 3/ Materials Engineering AECM-84/0330 (6/17/84)

Pagn 4 the region of acceptable operation is to the right of the curves and that B' and C' are coincident with B and C, respectively. Bases 3/4.4.6 is also revised to maintain consistency with the proposed change and to clarify that the revised figure complies with 10 CFR 50, Appendix G. This proposed revision provides consistency with GCNS FSAR Section 5.3.2.1. This change improves the safety of the plant by clarifying certain requirements, assuring conformance with current regulations, and assuring consistency with the plant as described in the safety analyses. (Pages 3/4 4-17, 3/4 4-18, 3/4 4-19, B 3/4 4-4, and B 3/4 4-5)

Z58mm4 m

Attachernt 3/ Materiels Engin22 ring AECM-84/0330 (6/17/84)

Page 5 B. TECHNICAL SPECIFICATION /AS-BUILT PLANT CONSISTENCY The following change is proposed to render the technic 1 specification consistent with the as-built plant. In all such cases, the as-built plant is consistent with the safety analyses and the licensing basis.

In that this proposed change is inherently consistent with the safety analyses and the licensing basis, it is concluded that the proposed change does not:

o Involve a significant increase in the probability or consequences of an accident previously evaluated; or ,

o Create the possibility of a new or different kind of accident from any accident previously evaluated; or o Involve a significant reduction in a margin of safety.

Therefore, the proposed change does not involve a significant hazards consideration. ;

A description of this change including justification fre the change is provided below:

1. (TSPS 319), Reference Code for Rx Vessel, Technical Specification Bases 2.1.2 and 2.1.3 This proposed change modifies Bases 2.1.3 to reflect the appropriate Edition and Addenda of the ASME Code. This proposed change makes the ;

technical specifications consistent with the as-built plant and is consistent with Section 5.3.3.1.1.1 of the FSAR. (Page B 2-5)

Z58mm5

Attachmsnt 3/Mstarials Enginsering AECM-84/0330 (6/17/84)

Pagt 6 C. ENHANCEMENTS THAT ARE CONSISTENT WITH THE SAFETY ANALYSES No technical specification changes in this category are included with !

this attachment.

i I

4 t

[

l t

I L

n 1

e i

?

I b

I il

}, L 4

(

yI .

f i

4 h

Z58mm6 1 ,-,..<..,b-- m.- . ,.- , ,y._.-w-m,.,-.,.- .,,.%- - - , , ---.m. ___,...,m.;,.--,---,,,,m---~7y.,-.--,_, ---

,ev r*v--"

" Attechment 3/Materitis Engineering AECM-84/0330 (6/17/84)

Page 7 D. REGULATORY REQUIREMENTS / REQUESTS / RECOMMENDATIONS The following change is proposed to render the technical specification consistent with recent changes in NRC policy and the Code of Federal Regulations, as well as to implement changes or enhancements recently requested or recommended by NRC reviewers.

This proposed change is required to render the technical specification consistent with recent NRC guidance, and it has been concluded based on a review of this item that the proposed change does not; o Involve a significant increase in the probability or consequences of an accident previously evaluated; or o Create the possibility of a new or different kind of accident from any accident previously evaluated; or o Involve a significant reduction in a margin of safety.

Therefore, the proposed change does not involve a significant hazards consideration.

A description of this change including justification for the change is provided below:

1. (TSPS 219) Reactor Pressure vs. Metal Temperature Curves, Technical Specification 3/4.4.6, Bases 3/4.4.6 This proposed change revises the curves of Figure 3.4.6.1-1 to conform to the revised requirements of 10CFR50, Appendix G, para-graph IV.A.2 and 3 (May, 1983); also, a statement is added to Bases 3/4.4.6 to indicate that the RT for welds and base material in theclosureregionisequaltoggTiess than 10*F and that the initial hydrostatic test pressure was 1563 psig. The curves presently in the technical specifications conform with the require-ments of the previous revision of Appendix G, with the exceptions noted in GE BWR Licensing Topical Report NED0-21778-A. The proposed curves are in compliance with the revised Appendix G requirements to maintain the appropriate margins (as defined in the revised IV.A.2) above the bolt-up reference temperature at pressures exceeding 20 percent of the preservice system hydrostatic test pressure. For pressures below 20 percent of the preservice hydrostatic test pressure, the proposed curves meet the requirement to maintain a 60*F margin above the bolt-up reference temperature, as defined in paragraph IV.A.3. It is therefore concluded that this change is in compliance with 10CFR50, Appendix G requirements and is in accordance with recent NRC guidance to implement the May,1983 revisions to Appendix G in the Grand Gulf Technical Specifications.

(Page 3/4 4-19 and B 3/4 4-4)

Z58mm7 L

' Attach =snt 3/Msterials Engineering AECM-84/0330 (6/17/84)

Page 8 E. PROPOSED TECHNICAL SPECIFICATION CHANGES (AFFECTED PAGES ARE PROVIDED IN THE ORDER OF ASCENDING PAGE NUMBERS.)

258m8

5 SAFETY LIMITS BASES 2.1.3 REACTOR COOLANT SYSTEM PRESSURE 1971 The Satety Limit for the reactor coolant system essure has been selected such that it is at a pressure below which it can be s own that the integrity of the system is not endangered. The reactor pressure ssel is designed to Sec-unden if 7A. tion III of the ASME Boiler and Pressure Vessel Code 1974-Edition, including D Addenda throughP5;;;;r 1075, which permits a maximum pressure transient of 110%, M 1375 psig, of design pressure, 1250 psig. The Safety Limit of 1325 psig, as measured by the reactor vessel steam dome pressure indicator, is equivalent to 1375 psig at the lowest elevation of the reactor coolant system. The pressure Safety Limit is selected to be the transient overpressure allowed by the ASME Boiler and Pressure Vessel Code,Section III, Class I. ;

2.1.4 REACTOR VESSEL WATER LEVEL With fuel in the reactor vessel during periods when the reactor is shutdown, consideration must be given to water level requirements due to the effect of decay heat. If the water level should drop below the top of the active irradiated fuel during this period, the ability to remove decay heat is reduced. This reduc-tion in cooling capability could lead to elevated cladding temperatures and clad -

perforation in the event that the water level became less than two-thirds of the core height. The Safety Limit has been established at the top of the active irradiated fuel to provide a point which can.be monitored and also provide adequate margin for effective action.

a l

A*6*d### '" l GRAND GULF-UNIT 1 B 2-5 L m

REACTOR COOLANT SYSTEM 3/4.4.6 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION phd reeck ocsse{ melsf tethperatute 3.4.6.1 The reactor coolant syctem t q craturc = d pressure ^shall be limited kq in accordance with the limit lines shown on Figure 3.4.6.1-1 (1) curve A for hydrostatic or leak testing; (2) curve B for heatup by non-nuclect means, cooldown following a nuclear shutdown and low power PHYSICS TEST 5; and (3) curve C for operations with a critical core other than icw power PHYSICS TESTS, with:

tracks coolauf

a. A maximumdheatup of 100?F in any one hour period, lk reacka coolant

b. A maximumdcooldown of 100'F in any one hour period, l

c. A maximum temperature change of less than or equal to 10?F in any one hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves, and

d. The reactor vessel flange and head flange temperature greater than or equal to 70'F when reactor vessel head bolting studs are under tension.

APPLICABILITY: At all times.

ACTION:

With any of the above limits exceeded, restore the temperature and/or pressure to within the limits within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the reactor coolant system; determine that the reactor coolant system

remains acceptable for continued operations or 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 SHUTOOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE REOUIREMENTS 4.4.6.1.1 During system heatup, cooldown and inservice leak and hydrostatic testing operations, the reactor coolant system temperature and pressure shall be determined to be within the above required heatup and cooldown limits anda to o the right of the limit lines of Figure 3.4.6.1-1 curves A or 4',4as applicabla, y at least once per 30 minutes, g , j g, lbe deadon *coolaa& s s fem flerrute and teacfot. vessel mebeY fempree fune ska0

}e n'tir&utvec/ 40 be.

GRAND GULF-UNIT 1 3/4 4-17 Aee~ din e*f No. --.- l

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) hp todot vessel maf*f temWoluRK o 4.4.6.1.2 The reactor coolant system t:ger: tere end pressured shall be deter- %

i

' mined to be to the right of the criticality limit line of Figure 3.4.6.1-1 curves C and C' within 15 minutes prior to the withdrawal of control rods to bring the reactor to criticality and at least once per 30 minutes during system heatup.

4.4.6.1.f4 The reactor vessel material specimens shall be removed and examined l kg '

te 6te--b rer tar presswe-vessel fluence as a function of time and THERMAL POWER as required by 10 CFR 50, Appendix H in accordance with the schedule in (

Table 4.4.6.1.3-1. -The-results-of-these-f4eence-deteminations-shall be :::d

-to-update the cerves of Figure 3.4.4 1. The edjusted-reference-t;=per:tur: ;

-cesulting-fra- neutren irradiation 05:11 bc :+1eulated b:: d en the gr;;t:r ef tM fel!*;- ;

f0# **t*'4*T* i" t # C'E3"1** '; d*#i";d Y t2 O

. Actual-shift in th: RTDT

-CVN-impact t::t. s 3

4. Tr:dicted-shift-in-ATm-fer-plate-C25aA-2 :nd =14-627260/8322A27AE f i

-Seat /4et) :: determiM4 by Regulatory Guide 1.99r"-Effect: Of 20 id2:1

-E1:::nte en Predicted-Radiation h-:ge te R:::t:r Y::::1 ".:teefals!'-e 4.4.6.1./3 The reactor vessel flange and head flange temperature shall be ".

l i

verified to be greater than or equal to 70'F:

a. In OPERATIONAL CONDITION 4 when reactor coolant system temperature l is: l l

~

1. $ 100*F, at least once per 17 hours1.967593e-4 days 0.00472 hours 2.810847e-5 weeks 6.4685e-6 months . l

2. S 80*F, at least once per 30 minutes. ;

i

b. Within 30 minutes prior to and at least once per 30 minutes during tensioning of the reactor vessel head bolting studs.

4! 4. 4. l. 5~

'lle werce //uc wite spe comee.s sha// be geonoved at Hie i

(

Nod se he/an9 c A ye ed emmined to de fecan ane see ek" press ute ve sse/ //ue ov ce es a Junchs,a of tione and powese le us / ud u sea' 4 mod <Cy p;yuge g g, q, g, - / , g l 7~he resul/s of 46e flu ence de feemuu a i%.r, <a cauguachou y; ut/4 Fly ute 8 %.44-1, sh all be used h ad gus-f fire ;

e usedes of oQute 3. V. L. /-/, e s d'efu rce d. \

i GRAND GULF-UNIT 1 3/4 4-18 Amendment No. 8,._ ;

r O

A BB' CC' ' L 1400 - i g

A -NITA WSTEM WDROTEE LWIT l Bottom Head B - NITIAL NON+AJCLEAR EATING LIMIT M -+

C - NITIAL NUCLEAR (CORE CRITICAL) 1200 .

LIMIT BASED ON G.E. BWR LICENSING '

TOPICAL REPORT NEDO-21778 A Core Beit OII _

A', B', C'- A, B, C LIMITS AFTER AN l ASSUMED 26*F CORE BELTLIE TEMP.

SHIFT FROM AN INITIAL SHELL PLATE 1000 ' RT NDT OT. A' ,lS E SHOWN ,

(NOT LIMITING) 6 a n d C a. Ye c.o r ** c t o s e r' wtTH 13 an d C, 3 i

}

Z ReStecTivfLY.

& l

>- 800 -

k Curves A, B ond C are predicted to ;

, be applicable for service periods . 0- (

up to 32 EFPY.

g M l

1 600 -

1

-Feedwater l Nozzle ,

h Limlls E . .

l I < $ 0 a.

400 - AccefTO ble ftfloW of etn R AT20N

j',*P o s T o -r'wr! voyh t o F Th e '

, l

,3I29'\9 a. y t t s e A IS LC cuRYG. )

70*r + , ,

O O

$2 !2 k 200 -

{

O , 500 100 200 300 400 O

RPV Metal Temperature (*F) l i

MINIMUM REACTOR PRESSURE VESSEL lETAL TEMPERATURE VS REACTOR l

Figure 3.4.6.1-1 GRAND GULF-UNIT I 3M 4 -19 Ah p'DMCA'I M* ~

-1%c'wR7k7- fu weth and base paterial w the closuee flage. ;

M

/}4Jpg. & lo *F. 75 iar M/ A des y k hc. des + pressue ua.s REACTOR COOLANT SYSTEM ,

i BASES -

j 3/4.4.6 PRESSURE / TEMPERATURE LIMITS All cot,,onents in the reactor coolant system are designed to withstand -

the effects of cyclic. loads due to system temperature and pressure changes. l These cyclic loads are introduced by normal load transients, reactor trips, l and startup and shutdown operations. The various categories of load cycles -

used for design purposes are provided in Section 3.9 of the FSAR. During startup and shutdown, the rates of temperature and pressure changes are limited so that !

the maximum specified heatup and cooldown rates are consistent with the design ;

assumptions and satisfy the stress limits for cyclic operation. ;

During heatup, the thermal gradients in the reactor vessel wall produce thermal stresses which vary from compressive at the inner wall to tensile at the outer wall. These thermal induced compressive stresses tend to alleviate the tensile stresses induced by the internal pressure. Therefore, a pressure- i temperature curve based on steady state conditions, i.e., no thermal stresses, represents a lower bound of all similar urves for finite heatup rates when -

the inner wall of the vessel is treate.d as the governing location. l The heatup analysis also covers the determination of pressure-temperature f limitations for the case in which the outer wall of the vessel becomes the con- ;

trolling location. The thermal gradients established during heetuc produce ;

tensile stresses which are already present. The thermal induces tresses at l the outer wall of the vessel are tensile and are dependent on bcth the rate of l heatup and the time along the heatup ramp; therefore, a lower bound curve similar ;

to that described for the heatup of the inner wall cannot be defined. Subse- i quently, for the cases in which the outer wall of the vessel becomes the stress ;

controlling location, each heatup rate of interest must be analyzed on an ;

individual basis. l The reactor vessel materials have been tested to determine their initial n RT NDT* 1 The results of these tests are shown in Table B 3/4.4.6-1. Reactor g [

operat"on and resultant fast neutron, E greater than 1 Mev, irradiation will ') (

'. cause an increase in the RT NDT. Therefore, an adjusted reference temperature, j based upon the fluence, phosphorus ccatent and copper content of the material !

in question, can be predicted using Bases Figure B 3/4.4.6-1 and the recommenda- i tions of Regulatory Guide 1.99, Revision 1, " Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials." The pressure / tempera- l ture limit curve, Figure 3.4.6.1-1, curves A', B' and C', includes predicted j adjustments for this shift in RT NDT for the end of life fluence as well as l adjustments for possible errors in the pressure and temperature sensing instruments durves A'od C' 4e coincidext 441/4 curves B and c,m,nedwely . hl, The actual shift in RT NDT of the vessel material will be established period- l ically during operation by removing and evaluating in accordance with ASTM E185-73 I and 10 CFR 50, Appendix H, irradiated reactor vessel material specimens installed l near the inside wall of the reactor vessel in the core area. The irradiated !

specimens can be used with confidence in predicting reactor vessel material transition temperature shift. The operating limit curves of Figure 3.4.6.1-1 shall be adjusted, as required, on the basis.of the specimen data and recom-sendations of Regulatory Guide 1.99, Revision 1.

. l GRAND GULF-UNIT 1 B 3/4 4-4 g),g g g l l t

T *

~

REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued)

The pressure-temperature limit lines shown in Figures 3.4.6.1-1, curves C, and C' and A =d ", for reactor criticality and for inservice leak and 4

- hydrostadetestinghavebeenprovidedtoassurecompliancewiththeminimum temperature requirements of Appendix G to 10 CFR Part 50 for reactor criticality and for inservice leak and hydrostatic testing.

3/4.4.7 MAIN STEAM LINE ISOLATION VALVES Double isolation valves are provided on each of the main steam lines to minimize the potential leakage paths from the containment in case of a line break. Only one valve in each line is required to maintain the integrity of the containment. The surveillance requirements are based on the operating history of this type valve. The maximum closure tirr.e has been selected to contain fission products and to ensure the core is not uncovered following

~

line breaks.

3/4.4.8 STRUCTURAL INTEGRITY The irispection programs for ASME Code Class 1, 2 and 3 components ensure that the structural integrity of these components will be maintained at an acceptable level thruughout the life of the plant.

Components of the reactor coolant system were designed to provide access to permit inservice inspections in accordance with Section XI of the ASME Boiler and Pressure Vessel Code, 1977 Edition, and Addenda through Summer 1978.

The inservice inspection program for ASME Code Class 1, 2 and 3 components will be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable addenda as required by 10 CFR Part 50.55a(g) except where specific written relief has been granted by the NRC pursuant to 10 CFR f Part 50.55a(g)(6)(i).

i 3/4.4.9 RESIDUAL HEAT REMOVAL A single shutdown' cooling mode loop provides sufficient heat removal capability for removing core decay heat and mixing to assure accurate tempera-t ture indication; however, single failure considerations require that two loops i be OPERABLE or that alternate methods capable of decay heat removal be demonstrated and that an alternate method of coolant mixing be in operation.

l

l. -

GRAND GULF-UNIT 1 B 3/4 4-5 Amedd***d A!o -

\ /

ML20092B309

Text

Navigation menu

Search