Text

Proposed TS Re Primary & Secondary Containment Integrity
	ML20056G582
Person / Time
Site:	Cooper
Issue date:	08/31/1993
From:	; NEBRASKA PUBLIC POWER DISTRICT
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ML20056G573	List: ML20056G572; ML20056G582;
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	NUDOCS 9309030295
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' NSD931023 i Page 1 of 11 REVISED TECHNICAL SPECIFICATIONS Revised Pares New Pares 11 166 165b 5 167 166a 165 182 167a 165a 205a ;

i I. INTRODUCTION The Nebraska Public Power District (District) requests that the NRC approve the proposed changes to the Cooper Nuclear Station (CNS) Technical Specifications described below. The requested changes include:

1) Revision of the Technical Specifications pertaining to the Standby Gas Treatment System (Section 3/4.7.B) for consistency with NUREG- i 1433, Standard Technical Specifications, BWR/4.

2) The clarification of the definition for Secondary Containment ;

Integrity and revision of the Technical Specification and Bases section pertaining to secondary containment (Section 3/4.7.C and !

associated Bases) for consistency with NUREG-1433, Standard Technical Specifications, BWR/4.

3) The clarification of the definition for Primary Containment Integrity and revision of Technical Specifications pertaining to {

primary containment automatic isolation valves (Section 3/4.7.D) for consistency with hTREG- 1433, Standard Technical Specifications.

BWR/4. L

4) Miscellaneous editorial changes including, but not limited to, movement of Technical Specification sections between pages, capitalization of defined terms, and making consistent the use of the term " containment automatic isolation valves".

l The individual propa -T changes are detailed in Section II and in the attached Summary or . anges. The bignificant Hazards determination is ,

provided in Section III. The District's analysis has determined that ,

approval of this proposed change involves no significant hazards !

consideration. !

II. DESCRIPTION OF CHANGES 1

1. Standby Cas Treatment System The first part of Proposed Change No.106 involves the revisiot of Section 3/4.7.B, Standby Gas Treatment System, to incorporate applicable Limiting j Conditions for Operations (LCOs) and Surveillance Requirements WRs) from '

NUREG-1433, BWR/4 Standard Technical Specifications (STS). This part of the proposed change includes the addition of a new SR (No. 4.7.B.1.c) that l

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. , I Attachment to NSD931023 Page 2 of 11 requires, at least once per operating cycle, that each Standby Gas Treatment (SGT) subsystem be demonstrated to maintain 2 0.25 inches of l water vacuum for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months at a flow rate of $ 1780 cubic feet per minute !

(CFM), which is the design basis flow of each of the SGT subsystems. This l part of the proposed change also includes the rewording of an existing LCO !

(No. 3.7.B.4), to be more explicit regarding actions to be taken when [

operability requirements of SGT cannot be met during reactor power !

operation.

[

Other minor changes associated with first part of the proposed change ;

consists of a revision to a SR (No.4.7.B.2.d) to require that each SGT subsystem be operated with the heaters on for at least 10 continuous hours each month, as opposed to 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months , and a clarification of a SR (No. !

4.7.B.4.b) that both SGT subsystems require operation of the bypass valve i for filter cooling. l, With the exception of the proposed new SR (No. 4.7.B.l.c) addressing the .

demonstration of the SCT for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months (which represents a new requirement !

based on STS), these changes represent clarifications to existing l requirements by adopting wording and terminology from the STS. The LCO l (No. 3.7.B.4) which describes actions to be taken if SGT system is found i to be inoperable, was rewritten in order to be consistent with the ,

corresponding STS action statements and to be similar with proposed wording contained in the LCO (No. 3.7.C.3 - to be discussed later) addressing actions to be taken should Secondary Containment Integrity not be maintained during reactor power operation. All of the above described I individual changes represent enhancements to Section 3/4.7.B of the CNS ;

Technical Specificati,ns and are more restrictive than those requirements currently in place. No Technical Specification requirements have been ,

deleted. !

I r

2. Secondary Containment The second part of Proposed Change No. 106 is in support of a District ;

commitment contained in correspondence (NSD930779) to the NRC, dated June i 22, 1993, and involves the revision of Section 3/4.7.C, Secondary !

Containment, to incorporate applicable LCOs and SRs from STS. This part i of the proposed change includes clarification of the definition of ;

Secondary Containment Integrity (Definition No. 1.0.V) to require the ,

secondary containment automatic isolation valves required to be closed during design basis accident conditions to be operable, or the affected l penetration is isolated. A new ILO (No. 3.7.C.2) has been added to !

explicitly state actions (including time limits based on STS) necessary' to maintain Secondary Containment Integrity when one or more secondary ;

containment automatic isolation valve (s), required to be closed during ,

accident conditions, is found to be inoperable. To correspond with this ;

new LCO, a new SR (No. 4.7.C.2) has been added to require a verification, at least every 31 days,'that all secondary containment penetrations not ;

capable of being closed by operable containment automatic isolation valves '

are closed by at least one valve, blind flange, or de-activated automatic ,

valve secured in the closed position. The new 31-day SR is more !

restrictive in that it provides explicit instruction regarding actions to ,

be taken when a secondary containment automatic isolation valve (s) is [

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Attachment to i NSD931023 t Page 3 of 11 l (are) found to be inoperable. This will provide a higher level of i assurance that Secondary Containment Integrity will be maintained, when !

required. The above described changes are consistent with wording !

contained in STS. (

l l For the above described LCO changes, a footnote has been added to state that isolation valves closed to satisfy this requirement (LCO No. 3.7.C) !

may be reopened on an intermittent basis under administrative control. !

This footnote is consistent with wording contained in STS and in Section i 3/4.7.D of the CNS Technical Specifications. i 3

This part of the proposed change also involves the replacement of the SR :

(existing SR No. 4.7.C.1.d) for the demonstration of Secondary Containment i Integrity (the ability to maintain 2 0.25 inches of water vacuum) by l immediately operating the SGT system once a secondary containment ;

violation has been determined and corrected. The proposed SR wording .!

would, instead, require the affected area, identified during the plant conditions of Reactor Power Operation or Hot Shutdown, of the pressure i retaining boundary of secondary containment to be verified through :

qualitative leak testing or evaluation prior to declaring Secondary [

Containment Integrity restored. ,

There are two reasons for this individual change. First, the secondary j containment at CNS is not designed so that the affected zone (s) can be ;

isolated. From the standpoint of maintaining Secondary Containment (

Integrity, the secondary containment constitutes a single zone. ;

Therefore, the provision to isolate the affected zone (s) in the subject j i surveillance requirement cannot be implemented at CNS. The second reason

for the replacement of the subject SR is that CNS design does not allow

for the concurrent isolation of secondary containment (Group 6 isolation) and starting of SGT during reactor power operation while maintaining the l portion of the ventilation system which cools the Reactor Recirculation i 2 Motor-Generator (RRMG) sets. Isolation of this portion of the ventilation j system would result in an high-temperature trip of the RRMG sets, which 5 would force a subsequent reactor scram.

The District proposes an additional SR which is contained in STS, but not currently contained in CNS Technical Specifications. SR 4.7.C.3 would {

require each secondary containment automatic isolation valve to actuate to :

its isolated position, within its specified isolation time, at least once i per operating cycle. Proposed SR 4.7.C.3 varies from STS in that such )

i demonstration be done at least once per operating cycle as opposed to once l per quarter. This proposed time frame is consistent with the existing primary containment automatic isolation valve testing requirements :

contained in SR (No. 4.7.D.1.a), and' takes into account the ventilation j requirements of plant equipment during reactor power operation. l

. i The District also proposes that two conditions be added to those !

conditions that must be met in order to not require Secondary Containment j Integrity; that no core alterations are taking place with irradiated fuel in the vessel, and no operations with a potential for draining the reactor ;

vessel are taking place with irradiated fuel in the vessel. The inclusion ;

of these two conditions affect existing LCO No.s 3.7.C.1.d, 3.7.C.3 l 1 (renumbered), and 3.10.E and effectively replace the condition regarding ll 1

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the movement of loads which could potentially damage irradiated fuel (This ,

condition is not contained in STS). These changes make consistent the ;

conditions and terminology contained in STS. The term "with irradiated fuel in the vessel" is not in STS, but is stated in the proposed CNS LCOs j 4 as a clarification. t

3. Primary Containment Automatic Isolation Valves The third part of Proposed Change 106 is part of an overall corrective !

action initiative related to License Event Report (LER) No. 50/285 91-008, ,

dated September 25, 1991. This LER describes the events leading up to a {

Technical Specification violation for the failure to de-activate an f inoperable containment isolation valve. One of the root causes of this i event was identified as an inconsistency in ths Technical Specification requirements for inoperable automatic isolation valves. The action to de- ;

activate inoperable isolation valves is currently contained in the definition for " Primary Containment Integrity," rather than in the action !

required by the LCO. l In response to the above described LER, the District proposes changes to ,

both Definition No. 1.0.P (Primary Containment Integrity) and Section !

3/4.7.D of the CNS Technical Specifications in order to be consistent with

STS, 'The change to the definition involves the clarification of an ;

existing condition to require that all containment automatic isolation valves required to be closed during accident conditions are operable or the affected penetration is isolated, as opposed to requiring that the specific inoperable valve be de-activated. This clarification is ,

necessary due to the fact that it is not always possible to de-activate the specific inoperable valve in its isolated condition, but it is possible to isolate the affected penetration. l Section 3/4.7.D (Primary Containment Automatic Isolation Valves) was l rewritten to provide explicit action.(including time limits specified in STS) requirements Wen a given primary containment automatic isolation i valve is determined to be inoperable. LCO 3.7.D.2 would allow for ,

continued reactor power operation provided the line, having an inoperable valve, is isolated by means of at least one closed manual valve, blind flange, or de-activated automatic valve secured in the closed position.

This new requirement is consistent with the revised definition of Primary Containment Integrity, and wi'.1 further assure that events, similar to that described in LER 285/50 91-008 do not recur. The existing SR 4.7.D.2 has been replaced by a new SR that requires the' verification, at least every 31 days, that primary containment penetrations, not capable of being isolated by operable containment automatic isolation valves, are isolated by valves, blind flanges, or de-activated automatic valves secured in the closed position. Similar to part of the proposed change for Secondary Containment Integrity, this part of the proposed change provides an option for either securing the inoperable valve, in the closed position, or securing another valve in the same line, 1

Also included in this part of the proposed change is a clarification to an .l existing LCO (No. 3.7.D.1) to specify that all primary containment automatic isolation valves listed in Table 3.7.1 of the Technical i

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Attachment to NSD931023 j Page 5 of 11 :

Specifications are to be operable when Primary Containment Integrity is required, as opposed to during reactor power operating conditions. This !

proposed wording is consistent with STS.

With the exception of the new 31-day SR, the changes to Definition 1.0.P l 4

and Section 3/4.7 D represent clarifications to existing requirements as opposed to the addition, deletion, or changes to requirements. The new 1.CO and 31-day SR is based on similar wording contained in STS. The new 31-day SR is more restrictive in that it provides explicit instruction l regarding actions to be taken when a primary containment automatic isolation valve (s) is (are) found to be inoperable. This will provide a higher level of assurance that Primary Containment Integrity will be maintained, when required.

l

4. Miscellaneous Editorial Channes

The fourth part of Proposed Change No. 106 involves numerous editorial changes not directly related to the Standby Cas Treatment, Secondary containment, or Primary Containment Automatic Isolation Valve portions of i this proposed change. These changes include, but are not limited to, page ;

renumbering, capitalization of defined terms, making consistent the use of i the terms " containment automatic isolation valves" and " instrument line excess flow check valves". The specific changes are identified in the attachment to this submittal.

I As an administrative note, Page 11 is subject to revision by Proposed

. Technical Specification Change No. 117. In order to facilitate the orderly revision of affected plant procedures, the District requests that l the amendment, associated with the proposed change, become effective 60 6 2 days after its issue date.

III. SIGNIFICANT HAZARDS DETERMINATION

! 10 CFR 50.91(a)(1) requires that licensee requests for operating license ,

j amendments be accompanied by an evaluation of significant hazards posed by 4

the issuance of the amendment. This evaluation is to be performed with respect to the criteria given in 10 CFR 50.92(c). The following analysis '

meets these requirements. ;

This submittal is judged to consist of four changes. ;

1. The clarification of existing requirements and the addition of new l requirements regarding the Standby Gas Treatment (SGT) system

(Section 3/4.7.B).

2. The clarification of existing requirements and the addition of new l requirements to both the definition of Secondary Containment l Integrity and its associated Technical Specifications (Sections 3/4.7.C) and Bases (4.7.B and 4.7.C). j

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3. The clarification of existing requirements and the addition of new requirements to both the definition of Primary Containment Integrity

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Attachment to NSD931023 .

Page 6 of 11 'I i

and the associated Technical Specifications for Primary Containment ;

Automatic Isolation Valves (Sections 3/4.7.D).

4. Miscellaneous editorial changes including, but not limited to, making consistent the use of valve terminology, page renumbering, and capitalization of defined terms.

Evaluation of this Amendment with Resnect to 10 CFR 50.92 i The enclosed Technical Specifications change is judged to involve no significant hazards based on the following:

A. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Evaluation

1. The first part of Proposed Change No. 106 revises Section 3/4.7 B, Standby Cas Treatment (SGT) System, in order to be.

consistent with NUREG-1433, Standard Technical Specifications ,

(STS) for General Electric Plants (BUR /4). This portion of the proposed change consists of adding a new surveillance requirement (SR) to demonstrate, at least once per operating ?

cycle, that each SGT subsystem can maintain 2 0.25 inches water vacuum for at least 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months at a flow rate of f 1780 cubic feet per minute (CFM) . Also included is a rewording to the Limiting Condition for Operation (LCO) governing actions t to be taken if the SGT system is made or found inoperable, and two clarifications based on STS. :

i All of the above changes are based on suggested wording ,

contained in STS and represent requirements that are more '

explicit or restrictive than what are currently in place.

These individual changes do not involve any. physical modification of the plant or delete any Technical Specification requirements currently in place. They do not i involve a change in plant settings and do not affect any ident initiators. For the reasons given above, the ,

b. strict concludes that this part of the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. The second part of Proposed Change No.106 revises a condition contained in the definition of Secondary Containment Integrity, Section 3/4 7.C, Secondary Containment, and the associated Bases section in order to be consistent with STS. .

This portion of the proposed change includes a clarification of a condition contained in the definition of- Secondary Containment Integrity and a creation of an LCO to clearly i

, specify actions to be taken when a given secondary containment automatic isolation valve becomes inoperable. A new 31-day-surveillance requirement has been added to verify that secondary containment penetration lines containing inoperable

Attacpment to NSD931023 Page 7 of 11 valves, are verified isolated.

Two additional SRs are also proposed; 1) that isolation time of the individual automatic isolation valves will be demonstrated at least once per operating cycle and, 2) to verify Secondary Containment Integrity through leak testing or evaluation of the affected area of the pressure retaining boundary prior to declaring Secondary Containment Integrity restored. Two new conditions (described as " operations with a potential for draining the reactor vessel with irradiated fuel in the vessel" and " core alterations with irradiated fuel in the vessel"), for determining when Secondary Containment Integrity is required, have also been added to the LCOs. The term "with irradiated fuel in the vessel" is not contained in STS, but is added to simply provide a clarification. These two new conditions effectively replace an existing condition regarding the movement of loads which could potentially damage irradiated fuel.

All of the above described changes, with the exception of the second new SR proposed, are based on STS. The second SR replaces an existing SR that is not implementable due to the fact that CNS secondary containment, by design, cannot be compartmentalized from a Secondary Containment Integrity standpoint. Furthermore, the ventilation system serving the Reactor Recirculation Motor-Generator (RRMG) sets would have to isolate in order to utilize SGT to create the required vacuum. Isolation of this portion of the ventilation system could result in a RRMG high-temperature trip, thus leading to a plant transient. Replacement of the old SR with the new SR removes the potential of creating a plant transient through implementation of the old SR, thus reducing the probability of an accident previously evaluated. Also, the LCO prohibiting continued reactor operation following a loss of secondary containment greater than 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months (when secondary containment is required) is unaffected by this SR change, and remains the dominant requirement. All of the above described changes provide additional Technical Specification controls on the management of secondary containment, and therefore will provide additional assurance that Secondary Containment Integrity continues to be met.

The individual changes contained within this proposed change do not involve any physical modification of the plant, do not affect any accident initiators, nor do they change any assumptions in the accident evaluations. There are no changes in plant settings that affect plant operation response. For the reasons given above, the District concludes that this part of the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

3. The third part of Proposed Change No. 106 revises a condition contained in the definition of Primary Containment Integrity l

l l

Attachment to NSD931023 Page 8 of 11 and Section 3/4.7.D, Primary Containment Automatic Isolation Valves, in order to be consistent with STS. This portion of the proposed change revises the subject definition to include additional conditions for inoperable primary containment automatic isolation valves. This change revises Section

'3/4.7.D to specify actions, including the establishment of i time limits based on STS, to be taken when a given primary containment automatic isolation valve becomes inoperable.

This part of the proposed change does not involve any physical modification of the plant or delete any Technical Speciff cation requirements currently in place. There are ne '

changes in plant settings that affect the plant operation response, nor are there any changes that affect any ,

assumptions in the accident evaluation.

The new 31-day surveillance requirement places a specified '

time period for the verification of one closed manual valve, ,

blind flange, or de-activated automatic valve secured in the closed position, in lines containing an inoperable valve. :

This surveillance requirement replaces an existing !

surveillance requirement with a more explicit verification ,

requirement and provides a higher assurance that Primary ,

Containment Integrity is being met. For the reasons given t above, the District concludes that this part of the proposed change does not involve a significant increase in the '

probability or consequences of an accident previously evaluated.

4. The fourth part of Proposed Change No. 106 involves numerous editorial changes not directly related to the Standby Gas Treatment, Secondary Containment, or Primary Containment Automatic Isolation Valve portions of this proposed change.

These changes include, but are not limited to, page ;

renumbering, capitalization of defined terms, making ;

consistent the use of the terms " containment automatic isolation valves" and " instrument line excess flow check ,

valves". These changes are editorial in nature, and have no ;

impact on plant equipment, plant design, or operations. These editorial changes do not modify or add any initiating parameters. Therefore, the District concludes that this part ;

of the proposed change does not involve a significant increase '

in the probability or consequences of an accident previously evaluated. -

B. Does the proposed change create the possibility for a new or !'

different kind of accident from any accident previously evaluated?

Evaluation

1. The first part of Proposed Change No. 106 revises Section 3/4.7.B. Standby Gas Treatment (SGT) System, in order to be :

consistent with STS. This portion of the proposed change adds i a new SR regarding the demonstration of SGT to maintain 2 0.23 4 i a

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Attachment to NSD931023 !

Page 9 of 11 l i

e inches of water vacuum for at least I hour at s 1780 CFM, a revision to the LCO governing actions to be taken if the SCT system is made or found inoperable, and two clarifications ;

I based on STS. All of the above changes are based on suggested wording contained in STS and represent requireuents that are

- more explicit or restrictive than what are currently in place. l These individual changes do not constitute any changes or 4 additions to any hardware or changes in plant configuration. ,

These individual changea do not introduce any new modes of '

plant operation. Both the revised LCO and the proposed SR are more restrictive than current Technical Specification requirements. Therefore, the District concludes that this part of the proposed change does not create the possibility for a new or different kind of accident from any accident previously evaluated. ;

2. The second part of Proposed Change No.106 revises a condition contained in the definition of Secondary Containment ,

Integrity, provides a new LCO (Section 3/4.7.C) to clearly l

, specify actions to be taken when a given secondary containment automatic isolation valve becomes inoperable, adds a new 31-day SR to verify that secondary containment penetration lines containing inoperable valves, are isolated, along with two additional SRs and two new conditions that must exist in order '

to not require secondary containment. All of the above '

described individual changes are based on STS except for a new SR which requires the affected area of the pressure retaining l boundary to be verified through qualitative leak testing or l evaluation prior to declaring Secondary Containment Integrity '

restored. l The SR that is not based on STS replaces an existing SR that is not implementable during normal plant operations, and if

implemented, could result in a RRMG high-temperature trip, thus leading to a plant transient. Implementation of the proposed SR does not require system lineups or tests that have not been previously analyzed; thus cannot crecte the possibility for a new or different accident from any accident previcusly evaluated.

The individual c.hanges contained in this portion of the proposed change do not involve a change in plant design, do not introduce a new mode of plant operation, nor will they i contribute to a change in the plant's transient response.

Therefore, the District concludes that this part of proposed change does not create the possibility for a new or different kind of accident from any accident previously evaluated.

3. The third part of Proposed Charge No. 106 revises the definition of the term Primary Containment Integrity, revises section 3/4.7.D to rnore clearly specify actions regarding inoperable containment automatic isolation valves. These

, individual changes do not constitute any hardware changes, additions, or changes in plant configuration. These changes l

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Page 10 of 11 !

r do not introduce any nes modes of plant operation, or contribute to a change in the plant's transient response. ,

There are no technical changes as to the limiting conditions for operations that must be satisfied. The new 31-day !

surveillance requirement is more restrictive than current Technical Specification requirements in that it provides

explicit instruction for ensuring that a given penetration (s)

! is isolated. Therefore, the District concludes that this part ,

of the proposed change does not create the possibility for a new or different kind of accident from any accident previously .

evaluated.

4 The fourth part of Proposed Change No. 106 involves numerous editorial changes not directly related to the Standby Gas Treatment, Secondary Containment, or Primary Containment ,

Automctic Isolation Valve portions of this proposed change.

Ther.e individual changes do not involve any alteration to the plaat design, setpoints, or operating parameters, nor do they introduce or change any mode. of plant operation. Therefore, this part of the proposed change does not create the r possibility for a new or different kind of accident from any accident previously evaluated.

i C. Does the proposed change create a significant reduction in the margin of safety? l Evaluation

1. The first part of Proposed Change No. 106 revises Section 3/4.7.B. Standby Gas Treatment (SGT) System, in order to be i consistent with STS. This portion of the proposed change j consists of the addition of a new SR (from STS), revision to l

an existing LCO to be more explicit regarding actions to be ;

, taken if the SGT system is made or found inoperable, and two j clarifications. All of the above changes are based on !

suggested wording contained in STS and represent requirements that are more explicit or restrictive than what are currently .

in place. These indivi'lai changes do not involve any change !

to plant design, equipment, instraent setpoint settings, or operation. Therefore, the District concludes that this part of .

the proposed change does not create a significant reduction in !

the margin of safety. -)

2. The second part of Proposed Change No.106 revises a condition ,

contained in the definition of Secondary Containment !

f Integrity, provides a new LCO (Section 3/4.7.C) to clearly specify actions to be taken when a given secondary containment automatic isolation valve becomes inoperable, adds a new 31- i day SR to verify that secondary containment penetration lines !

containing inoperable valves, are isolated, along with two e 4 additional SRs and two new conditions that must be met-in order to not require secondary containment All of the above l described individual changes are based on STS except for a new ;

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Attachment to >

NSD931023 Page 11 of 11 )

SR which requires qualitative leak testing or evaluation of the affected secondary containment pressure retaining boundary prior to declaring Secondary Containment Integrity restored.

None of the new requirements result in operation or testing -

that is different than what is currently being performed.

i Therefore, the District concludes that this part of the proposed change does not create a significant reduction in the i margin of safety.

r

3. The third part of the proposed change revises the definition of the term Primary Containment Integrity to clarify conditions to be met regarding inoperable primary containment l automatic isolation valves. This change revises Section 3/4.7.D to specify actions to be taken when a given primcry containment automatic isolation valve becomes inoperable. Se individual changes do not change the operating requirov.nts 1 specified in the Technical Specifications, but are more restrictive in that they provide explicit instruction .

regarding actions to be taken when a primary containment ,

automatic isolation valve is found inoperable. By placing these requirements into LCO 3.7.D and providing a 31-day >

surveillance requirement for lines containing inoperative-valves, the margin of safety is not reduced. None of the

, proposed individual changes favolve any change to the plant ,

design, equipment, instrument setpoint settings, or operation. l' Therefore, the District concludes that the proposed change does not create a significant reduction in the margin of '

safety.

4. The fourth part of Proposed Change No. 106 involves numerous editorial changes not directly related to the Standby Gas Treatment, Secondary Containment, or Primary Containment l Automatic Isolation Valve portions of this proposed change.

These individual changes do not involve any change to plant ,

design, equipment, instrument setpoint settings, or operation. [

Therefore, the District concludes that this part of the i proposed change does not create a significant reduction in the j margin of safety.

l IV. CONCLUSION The District has evaluated the proposed changes described above against j the criteria given in 10 CFR 50.92(c) in accordance with the requirements

of 10 CFR 50.91 (a)(1), This evaluation has determined that this proposed change will D2.t 1) involve a significant increase in the probability or ;

consequences of an accident previously evaluated, 2) create the !

possibility for a new or different kind of accident from any accident ,

previously evaluated, or 3) create a significent reduction in'the margin of safety. Therefore, for the reasons detailed above, the District requests NRC approval of this Proposed Change No. 106. !

Point of Contact for further information: David N. Madsen (402) 563-5005 l

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NSD931023 !

Page 1 of 5 !

SUMMARY

OF CHANGES The following is a list of changes to CNS Technical Specifications associated with Proposed Change No. 106.

Chances Associated with Standby Cas Treatment System ,

a

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Page 165 - Add new SR 4.7.B.1.c requiring demonstration of Standby Gas i Treatment (SGT) subsystems to maintain 2 0.25 inch water vacuum for l one hour at 5 1780 CFM during refueling outage this SR is consistent with similar SR contained in the Secondary Containment portion of Standard Technical Specifications (STS).

Page 165a - Reword LCO 3.7.B.4 to explicitly state - that if SGT system operability requirements are not met, then 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 COLD SHUTDOWN within the ;

following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . This new wording is being added to be consistent with STS regarding SGT and POWER REACTOR OPERATION.

- Reword SR 4.7.B.2.d to require each Standby Gas Treatment (SGT) l subsystem to be operated 10 continuous hours every month as opposed to 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months every month in order to be consistent with wording contained in STS. ,

I

- Reword SR 4.7.B.4.b to clarify that OPERABILITY of the bypass valve for filter cooling to each SGT subsystem shall be demonstrated. The ,

requirements have not changed; wording changed to be consistent with STS wording.

Page 205a - For LCO 4.10 E. replace the term " handling of irradiated fuel" with the STS term " movement of irradiated fuel assemblies". Replace the Non-STS statement " movement of loads which could potentially damage l 3 irradiated fuel in the secondary containment" with the STS l conditions " core alterations with irradiated fuel in the vessel, and 1 operations with a potential for draining the reactor vessel with irradiated tuel in the vessel". This change is necessary to make the LCO condicions consistent with STS and changes made to proposed -

LCOs 3. 7. C.1. d and 3. 7. C. 3.b (discussed below). The term "with irradiated fuel in the vessel" is not in STS, but is added as a clarification.

Changes Associated with Secondary Containment Page 5 - Reword Condition No. 3 of Definition 1.0.V to explicitly state that all containment automatic isolation valves required to be closed during accident conditions are operable or the affected penetration is isolated.

Page 166 - For LCO 3.7.C.1.d, change wording to use the STS term " fuel assemblies" as opposed to fuel. Replace the word " handled" with the STS word " moved". Remove the statement "no loads which could ;

potentially damage irradiated fuel in the secondary containment" !

(not in STS) and replace with two additional conditions contained in STS: "no core alterations are taking place with irradiated fuel in !

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.2tachment to NSD931023 .

Page 2 of 5 the vessel", and "no operations with a potential for draining the reactor vessel are taking place with irradiated fuel in the vessel".

Because none of these activities are allowed when secondary ;

containment integrity does not exist (see LCO 3.7.C.1 wording), the i term "and" has been retained. The term "with irradiated fuel in the ;

vessel" is not in STS, but is added as a clarification.

- Add new LCO 3.7.C.2 to require that secondary containment I penetrations having inoperable valves which are required to be l closed during accident conditions, are isolated. In order to be consistent with STS wording, new Tech Spec time durations to isolate ;

the penetration are established: 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months limit for flow paths with one secondary containment isolation valve inoperable, 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for ;

flow paths with two secondary containment isolation valves inoperable.

- Add footnote to the bottom of the page to allow for the intermittent opening of the subject valves under administrative controls (footnote currently exists for primary containment isolation valves and is also contained in STS).

- For LCO 3.7.C.3.b (renumbered), replace statement " irradiated fuel handling operations" with " movement of irradiated fuel assemblies",

add the term "IMMEDIATELY" . Delete the statement " movement of loads

which could potentially damage irradiated fuel in the secondary

containment" and insert an additional action to suspend operations ,

with the potential for draining the reactor vessel. All of these ;

changes are made in order for the LCO to be consistent with STS i wording.

- Replace wording contained in SR 4. 7.C. l .d with new SR wording requiring SECONDARY CONTAINMENT INTEGRITY to be verified through l qualitative leak testing or evaluation of the affected area of the pressure retaining boundary prior to declaring SECONDARY CONTAINMENT INTEGRITY restored (SR applicable to REACTOR POWER OPERATION or HOT ,

The design of CNS does not allow the i SHUTDOWN condition).

demonstration of SECONDARY CONTAINMENT INTEGRITY by use of SGT ,

, system during REACTOR POWER OPERATION. Isolation of the SGT system would result in the isolation of the ventilation system which serves .

the Reactor Recirculation Motor Generator (RRMG) sets, which could l ultimately result in a high-temperature trip of the RRMG sets. l Also, it is not possible to isolate a given area of the reactor ,

building from the balance of the reactor building. The new SR i allows for immediate verification of secondary containment through leak testing and the demonstration of SECONDARY CONTAINMENT !

INTEGRITY, at the soonest practical time. STS does not contain a !

requirement for verification of SECONDARY CONTAINMENT INTEGRITY following major maintenance activities of secondary containment. ;

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- Add new SR 4.7.C.2 to demonstrate SECONDARY CONTAINMENT INTEGRITY, by verifying, at least every 31 days, that secondary containment i penetrations are isolated when associated inoperable isolation valves are identified. This proposed SR is based on similar wording contained in STS.

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Attachment to !

NSD931023 !

Page 3 of 5 j

- Add new SR 4.7.C.3 to be consistent with STS with one variation. [

Isolation time will be demonstrated once per OPERATING CYCLE, as j opposed to once per quarter. This time frame is consistent with the

existing primary containment isolation valve SRs provided in SR , 4.7.D.1.a.

Page 182 - Add statement to the 5th line of the 1st paragraph to 4.7.B & 4.7.C, Bases to support the new SR 4.7.C.3 which states that " timing of the isolation valves verify they are closing within the design basis l requirements as specified in the USAR".

Chances Associated with Primary Containment Automatic Isolation Valves ;

Page 5 - Reword Condition No. 3 of Definition 1.0.P to explicitly state that i all containment isolation valves required to be closed during ;

accident conditions are operable or the affected penetration is isolated. 'l I

Page 167 - For LCO 3.7.D.1, clarify specification to state that all primary containment isolation valves listed in Table 3.7.1 to be operable ,

when PRIMARY CONTAINMENT INTEGRITY is required, as opposed to during !

reactor power operating conditions. This proposed wording is ;

consistent with STS.

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- For LCO 3.7.D.2, replace the existing LCO with a requirement that l primary containment penetration flow paths, having one or more j inoperable automatic isolation valve (s), are isolated. In order to -!

be consistent with STS wording, new Tech Spec time durations are established: 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months limit for lines containing a single inoperable ;

valve (8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months for main steam line); 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months if both the inboard and i i

outboard isolation valves are inoperable. j

- For SR 4.7.D.2, replace current surveillance requirement with a new !

surveillance requirement that requires the verification, at least once every 31 days, that lines containing inoperable primary containment isolation valves are isolated. This proposed SR is {

based on similar wording contained in STS. !

Miscellaneous Editorial Changes j Page 11 - Change page number references for Section 3.7.C, " Secondary l Containment" from Page 165a to Page 166, and for Section 3.7.D, i

" Primary Containment 1 solation Valves", from Page 166 to Page 167, j to reflect changes in the body of the Tech Specs. l

- Remove index entry for Section 3.5.H. ' This section was removed per l Amendment No. 163, but was inadvertently reintroduced in Amendment [

164. There are no changes to the corresponding Technical Specifications. ,

Add the word " Automatic" between the words " Containment" and i

" Isolation" in the reference to Section 3.7.D. -l l

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Attachment to NSD931023 Page 4 of 5 Page 5 - For Definition 1.0.P. Condition 3, reword the term " automatic '

containment isolation valves" as " containment automatic isolation valves". ;

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- For Definition 1.0.V, Condition 3, replace the term " automatic l ventilation system" with " containment automatic".

Page 165 - Capitalize the following existing defined terms: OPERABLE (2 l locations), SECONDARY CONTAINMENT INTEGRITY (1 location), and !

OPERATING CYCLE (1 location). -l 1

- For LCO 3.7.B.2.a, last line, replace "$ 0.25 Wg" with "2 0.25 inch of water vacuum". ;

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- Move LCO 3.7.B.2.a and SR 4.7.B.2.a to Page 165a. ,

i Page 165a - Capitalize the following existing defined terms: OPERABLE (5 locations), OPERABILITY (3 locations), OPERATING CYCLE (2 locations), IMMEDIATELY (1 location), SECONDARY CONTAINMENT ,

INTEGRITY (1 location). !

- For LCO 3.7.B.2, change the term "CMF" to the correct units "CFM".

- For LCO 3.7.C.1, Line 3, insert "as specified in 3.7.C.2 or" between i the terms "except" and "when". This change simply accommodates the addition of LCO 3.7.C.2 (previously discussed).

- For LCO 3.7.B. 3, Line 4, change and capitalize the term " reactor ,

operation" to " REACTOR POWER OPERATION".

- For LCO 3.7.C.3, change the statement "If secondary containment ,

integrity cannot be maintained" with the statement "If Specification 3.7.C.1 and 3.7.C.2 cannot be met". This change is simply to recognize the addition of LCO 3.7.C.2 and does not change the requirements. It is also made in order to be consistent with !

, similar LCO 3.7.D.3 and is consistent with other action statements throughout the CNS Tech Specs.

- Move Section 3/4.7.B.4 to new Page 165b. ,

- Move Section 3/4.7.C, Secondary Containment, to Page 166. Make corresponding change to Table of Contents.

Page 166 - Capitalize the following existing defined terms: SECONDARY l CONTAINMENT INTEGRITY (1 location). HOT SHUTDOWN (1 location), COLD j SHUTDOWN (1 location), OPERATING CYCLE (1 location), and REFUELING l OUTAGE (1 location).

- For SR 4.7.C.1 a, last'line, rephrase the term (p- wind speed)" to

"(p- average wind speed)". This change is simply to be consistent with implementing procedures and represents no chance - in the requirements.

- Renumber LCO 3.7.C.1.e as 3.7.C.3 to reflect the addition of new 140

Attachment to NSD931023 Page 5 of 5 ,

1 3.7.C.2.

- Move new LCOs 3.7.C.2 & 3 and SRs 4.7.C.2 & 3 to new Page 166a.

- For SR 4.7.C.1.a, Line 10, and SR 4.7.C.1.c, Line 2, replace the l term "1/4" with "0.25". !

Page 167 - Move Section 3/4.7.D, Primary Containmtnt Isolation Valves, to Page {

No. 167. Make corresponding change to Table of Contents. ;

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- Capitalize the following existing defined terms: OPERABLE . (2 i locations), OPERATING CYCLE (3 locations), OPERABILITY (1 location), .

SHUTDOWN (2 locations), and COLD SHUTDOWN (1 location). j

- For the headings to Sections 3.7 D and 4.7.D, insert the word [

" Automatic" between the words " Containment" and " Isolation". ;

- For LCO 3. 7.D.1, Line 3, change the term " isolation valves" to

, " primary containment automatic isolation valves". For Line 5, j change the term " instrument line flow check valves" to " instrument j line excess flow check valves".

- For SR 4.7.C.1, Line 1, add the word " automatic" between the words

" containment" and " isolation". -

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- For SR 4.7.D.1.c, change term " instrument line flow check valves" to ;

" instrument line excess flow check valves". !

- For SR 4.7.D.1.a, Line 4, replace the term " initiated" with the term

" actuated". For Line 5, replace the term " initiation" with :

" ACTUATION" and capitalize the term " SIMULATED AUTOMATIC". i t

- Move revised Section 3/4.7.D.2, associated footnote, and 3.7 D.3 to ;

new Page 167a.

Page 182 - For 3.7.B & 3.7.C, Paragraph 2 Line 4, replace the term " reactor ,

operation" with the term " REACTOR POWER OPERATION". i s

- For 4.7.B and 4.7.C, second paragraph (due to addition of new first ,

paragraph), Line 2, replace the term "1/4" with the term "0.25".

Page 205a - Capitalize the following existing defined terms: OPERABLE (8 !

locations), OPERABILITY (3 locations), FHUTDOWN (1 location), {

REFUELING OUTAGE (1 location), SPIRAL REIDAD (1 location) . !

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REVISED TECHNICAL SPECIFICATION PAGES i

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I TABLE OF CONTENTS (cont'd)

Pare No.

SURVEILLANCE LIMITING CONDITIONS FOR OPERATION REOUIREMENTS 3.5 CORE AND CONTAINMENT COOLING SYSTEMS 4.5 114 - 131 '

A. Core Spray and LPCI Systems A 114 B. RER Service Water System B 116 C. HPCI System C 117 D. RCIC System D 118 ;

E. 119 Automatic Depressurination System E F. Minimum Low Pressure Cooling System Diesel i Generator Availability F 120 G. Maintenance of Filled Discharge Pipe G 122 -

1 3.6 PRIMARY SYSTEM BOUNDARY 4.6 132 - 158 A. Thermal and Pressurization Limitations A 132 B. Coolant Chemistry B 133a C. Coolant Leakage C 135 D. Safety and Relief Valves D 136 E. Jet Pumps E 137 F. Recirculation Pump Flow Mismatch F 137 :

G. Inservice Inspection G 137 H. Shock Suppressors (Snubbers) H 137a ;

3.7 CONTAINMENT SYSTEMS 4.7 159 - 184 ,

A. Primary Containment A 159 B. Standby Cas Treatment System B 165 C. Secondary Containment C 166 D. Primary Containment Automatic Isolation Valves D 167 i

3.8 MISCELLANEOUS RADI0 ACTIVE RATERIAL SOURCES 4.8 185 - 190 i 3.9 AUXILIARY ELECTRICAL SYSTEMS 4.9 191 - 202 A. Auxiliary Electrical Equipment A 191 B. Operation with Inoperable Equipment B 192 3.10 CORE ALTERATIONS 4.10 203 - 209 1 A. Refueling Interlocks A 203 ,

B. Core Monitoring B 205 )

C. Spent Fuel Pool Water Level C 205 D. Time Limitation D 206 )

E. Spent Fuel Cask Handling E 206 ;

3.11 FUEL RODS 4.11 210 - 214e A. Average Planar Linear Heat Generation Rate (APLHGR) A 210 B. Linear Heat Generation Rate (LHCR) B 210 C. Minimum Critical Power Ratio (MCPR) C 212 1

f0R l^)FORmnon Outy - Nor AMT o, Pwosa CHAns /06 Mode - The reactor mode is established by the mode selector-switch. The modes L.

include refuel, run shutdown and startup/ hot standby which are defined as follows:

1. Refuel Mode - The reactor is in the refuel mode when the mode switch is in the REFUEL position. Then the mode switch is in the REFUEL position, the refueling interlocks are in service.

2. Run Mode - In this mode the reactor system pressure is at or above 625 psig and the reactor protection system is energized with APRM protection (excluding the 15% high flux trip) and RBM interlocks in service.

3. Shutdown Mode - The reactor is in the shutdown mode when the mode switch is in the SHUTD0'Wi position.

4. Startup/ Hot Standbv - In this mode the reactor protection scram trips initiated by the main steam line isolation valve closure are bypassed, the low pressure main steam line isolation valve closure trip is bypassed, the reactor protection system is energized with APRM (15% SCRAM) and IRM neutron monitoring system trips and control rod withdrawal interlocks in service.

L.' ';orral "entilation - ':ormal ventilation is the controlled process of discharging and replacing air from/to a confinement to maintain temperature, humidity, or other conditions necessary for personnel safety and entry. The contents of the atmosphere being discharged from the confinement will have been established prior to establishing normal ventilation following a purging / venting operation.

M. Offsite Dose Assessment Manual I ODAm - An OFFSITE DOSE ASSESSMENT MANUAL (ODAM) shall be a manual containing the methodology and parameters to be used in the calculation of offsite doses due to radioactive gaseous and liquid effluents, calculation of gaseous and liquid effluent monitoring instrumentation alarm / trip setpoints, and describes the Environmental Radiation Monitoring Program.

S. Operable - Operability - Ope ra ti n g ,

1. Operable - Operabilits - A system, subsystem, train, component or device j shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function (s) . Implicit in this definition shall be the assumption that all necessary attendant instrumentation, controls, normal and j emergency electrical power sources (except as specified in Section 1.0.J and 3.9), cooling or seal water, lubrication or other auxiliary equipment that are required for the system. subsystem, train. component or device to perform its function (s) are also capable of performirg their related

. support function (s).

A verification of OPF.RABILITY is an administrative check, by examination of appropriate plant records (logs, surveillance test records), to determine that a system, subsystem, train, component or device is not inoperable. Such verification does not preclude the demonstration (testing) of a given system, subsystem, train, component or device to determine OPERABILITY.

2. Operating - Operating means a system, subsystem, train, component, or i device is performing its intended function in its required manner.

O. Operatine Cvele - Interval between the end of one refueling outage and the end of the next subsequent refueling outage. j P. Primarv Containment Interrirv - Primary containment integrity means that the !

drywell and pressure suppression chamber are intact and all of the following conditions are satisfied:

1. All manual containment isolation valves on lines connected to the reactor coolant system or containment, and which are not required to be open during accident conditions, are closed.

2. At least one door in each airlock is closed and sealed.

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3. All containment automatic isolation valves required to be closed during accident conditions are operable or the affected penetration is isolated. I 1

4. All blind flanges and manways are closed.

P.A furre - Purrinc - Purge or Purging is the controlled process of discharging air ;

or gas from a confinement to establish temperature, pressure, humidity, ,

concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.

P.B Process control Protram -

The Process Control Program outlines the solidification of radioactive waste from liquid systems. It does not substitute for station operating procedures, but provides a general description of equipment, controls, and practices to be considered during waste solidification to assure solid wastes.

Q. Rated Power - Rated power refers to operation at a reactor power of 2381 megawatts thermal. This is also termed 100% power and is the maximum power level authorized by the operating license. Rated steam flow, rated !

coolant flow. rated neutron flux, and rated nuclear system pressure refer to !

the values of these parameters when the reactor is at rated power. )

R. Reactor Power Operation - Reactor power operation is any operation with the -]

mode switch in the "Startup/ Hot Standby" or "Run" position with the reactor l critical and above 1% rated power.

i S. Reactor Vessel Pressure - Unless otherwise indicated, reactor vessel pressures i listed in the Technical Specifications are those measured by the reactor vessel steam space detectors. ,

T. Refuelinn Outare - Refueling outage is the period of time between the shutdown .

of the unit prior to a refueling and the startup of the plant after that !

refueling.

U. Saferv Limits - The safety limits are limits within which the reasonable maintenance of the fuel cladding integrity and the reactor coolant system integrity are assured. Violation of such a limit is cause for unit shutdown and review by the Nuclear Regulatory Commission before resumption of unit operation. Operation beyond such a limit may not in itself result in serious consequences but it indicates an operational deficiency subject to regulatory review.

1

, V. Secondarv Containment Interrity - Secondary containment integrity means that the reactor building is intact and the following conditions are met:

1. At least one door in each access opening is closed.

2. The standby gas treatment system is operable.

3. All containment automatic isolation valves required to be closed during l ]

accident conditions are operable or the affected penetretion is isolated. l j 1

W. Shutdcwn - The reactor is in a shutdown condition when the mode switch is in j the " Shutdown" or " Refuel" position.

1. Hot Shutdown means conditions as above with reactor coolant temperature greater than 212*F.

2. Cold Shutdown means conditions as above with reactor coolant temperature equal to or less than 212*F and the reactor vessel vented.

5-

T IMIT!"c r0NDITION FOR OPrD ATTON CUDVrTT ' Wer DrOTTTDr"FNT 3.7. A (cont' d. ) 4.7.A (cont *d.)

6. Low-Low Set Relief Function 6. Low-Low Set Relief Function

a. The low-low set function of the a. The low-low set safety / relief valves safety-relief valves shall be shall be tested and calibrated as l OPERABLE when there is irradiated specified in Table 4.2.B.

fuel in the reactor vessel and the reactor coolant temperature is

> 212*F, except as specified in 3.7.A 6 a.1 and 2 below.

1. With the low-low function of one safety / relief valve (S/RV) inoperable, restore the inoperable LLS S/RV to OPERABLE within 14 days or be in the HOT STANDBY mode 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 low-lov set function of both S/RVs inoperable, be in at '

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

b. The pressure switches which control the low-low set safety / relief valves shall have the following settings. _

NBI-PS-51A Open Low Valve

1015 20 psig (Increasing)

NBI-PS-51B Close Low Valve 875 20 psig (Decreasing)

NBI-PS-51C Open High Valve j 1025 20 psig (Increasing) ;

NBI-PS-51D Close High Valve l 875 20 psig (Decreasing) l B. Standbv Gas Treatment System B. Standbv Cas Treatment System

1. Except as specified in 3.7.B.3 1. At least cnce per OPERATING CYCLE l ,

below, both Standby Gas Treatment the following conditions shall be subsystems shsll be OPERABLE at all demonstrated.

times when SECONDARY CONTAINMENT INTECRITY is required. a. Pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 6 inches of water at the system design flow rate.

b. Inlet heater input is capable of reducing R.H. from 100 to 70% R.H. ;

c. Demonstrate that each Standby Gas l Treatment subsystem can maintain 2 0.25 inch of water vacuum in the secondary containment for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months _at a flow rate of 5; 1780 CFM under calm j wind conditions (2<F<5 mph where ,

F- average wind speed). !

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. t LIMITING CONDITION FOR OPERATION SURVEILLANCE REOUIREMENT i 3.7 3 (cont'd) 4.7.B (cont'd) 2.a. The results of the in-place cold DOP 2.a. The tests and sampie analysis of I leak tests on the HEPA filters shall Specification 3.7.B.2 shall be '

show 299% DOP removal. The results performed at least once every of the halogenated hydrocarbon leak 18 months for standby service or tests on the charcoal adsorbers after every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of system shall show 299% halogenated operation and following significant ,

hydrocarbon removal. The DOP and painting, fire or chemical release halogenated hydrocarbon tests shall in any ventilation zone be performed at a Standby Gas communicating with the system.

Treatment flowrate of $1780 CFM and at a Reactor Building pressure of ,. Cold DOP testing shall be performed l 2 0.25 inch of water vacuum, after each complete or partial i replacement of the HEPA filter bank

b. The results of laboratory carbon or after any structural maintenance ,

sample analysis shall show 299% on the system housing.

radioactive methyl iodide removal with inlet conditions of: velocity c. Italogenated hydrocarbon testing -

227 FPM, 21,75 mg/m 3 inlet methyl hall be performed after each [

iodide concentration, 270% R.H. and complete or partial' replacement of i s30*C. the charcoal adsorber bank or after I any structural maintenance on the

c. Each fan shall be shown to provide system housing.

1780 CFM i10%.

d. Each subsystem shall be operated with the heaters on at least t 10 continuous hours every month. l

e. Test sealing of gaskets for housing [

doors downstream of the HEPA filters i and charcoal adsorbers shall be i performed at, and in conformance ;

with, each test performed 'for compliance with Specification ,

4.7.B.2.a and Specification l 3.7.B.2.a. ;

3. From and after the date that one 3. System drains where present shall be I Standby Gas Treatment subsystem is inspected quarterly for adequate l made or found to be inoperable for water level in loop-seals.

l any reason, REACTOR POWER OPERATION -i is permissible only during the l succeeding seven days unless such l l subsystem is sooner made OPERABLE, [

provided that during such seven days '

all active components that affect l OPERABILITY of the OPERABLE Standby .l Cas Treatment subsystem, and its

~

associated diesel generator, shall l be OPERABLE.

Fuel handling requirements are ,

specified in Specification 3.10.E.

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'T.IMITING CONDITIONS FOR OPERATION Sl'PNEILIANCE REOUIREMENTS ,

t 3.7.B (cont'd) 4.7.B (cont'd) l

4. If the conditions of LCO 3.7.B.1, 4.a. At least once per OPERATING CYCLE l l 2, or 3 cannot be met, be in at automatic initiation of each- ;

least HOT SHUTDOWN within the next . Standby Gas Treatment subsystem 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and COLD SHUTDOWN within shall.be demonstrated. .j the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . p

b. At least once per OPERATING CYCLE i manual OPERABILITY of the bypass !

valve for filter cooling to each l Standby Gas Treatment subsystem 3 shall be demonstrated. !

c. When one Standby Gas Treatment subsystem becomes inoperable, the t OPERABLE Standby Gas Treatment l .,

subsystem shall be verified to be OPERABLE IMMEDIATELY and. daily l .l thereafter. A demonstration of ,

diesel generator OPERABILITY is l.

not required by this :

specification. ;

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LIMITING CONDITIONS FOR OPERATTON SURVEILLANCE REOUIREMENTS 3.7.C Secondary Containment 4.7.C Secondary Containment '

1. SECONDARY CONTAINMENT INTEGRITY 1. Secondary containment surveillance shall be maintained during all shall be performed as indicated '

modes of plant operation except as below:

specified in 3.7.C.2 or when all i of the following conditions are a. A preoperational secondary [

met. containment capability test shall - i be conducted after isolating the

a. The reactor is suberitical and reactor building and placing Specification 3.3.A is met, either Standby Gas Treatment '

subsystem filter train in

b. The reactor water temperature is operation. Such tests shall ,

below 212*F and the reactor demonstrate the capability to !

coolant system is vented. maintain 0.25 inch of water vacuum l ,

calm wind (2<F<5 mph) under

c. No activity is being performed conditions with a filter train which can reduce the shutdown flow rate of not more than 100% of margin below that specified in building volume per Specification 3.3.A. (p- average wind speed) day.l '

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d. No irradiated fuel assemblies are b. Additional tests shall be j being moved in the secondary performed during the first 3 containment, no core alterations OPERATING CYCLE under an adequate l l are taking place with irradiated number of different environmental .

fuel in the vessel, and no wind conditions to enable valid operations with a potential for extrapolation of the test results.

draining the reactor vessel are ;

taking place with irradiated fuel c. Secondary containment capability I in the vessel. to maintain 0.25 inch of water vacuum under calm wind (24p<5 mph) conditions with a filter train flow rate of not more chan 100% of building volume per day, shall be demonstrated at each REFUELING OUTAGE prior to refueling.

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d. When the integrity of the pressure )

retaining boundary of secondary containment has been lost during POWER REACTOR .0PERATION or HOT SHUTDOWN, then SECONDARY CONTAINMENT INTEGRITY--shall be verified through qualitative leak e testing or- ' evaluation of the affected area prior .to declaring SECONDARY CONTAINMENT INTEGRITY restored.

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1.1MITING CONDITIONS FOR OPERATION SURVEILLANCE FEOUTREMENTS j 3.7.C (cont'd) 4.7.C (cont'd) i

2. Whenever SECONDARY CONTAINMENT 2. When it is determined that one or l INTEGRITY is required, but one or more secondary . containment more penetration flow paths automatic isolation valve (s) ,

contain an inoperable secondary required to be closed during l containment automatic isolation accident conditions is inoperable i valve required to be closed during when SECONDARY CONTAINMENT.

accident conditions, then isolate INTEGRITY is required, the ;

the affected penetration (s), affected penetration (s) shall be l having an inoperable automatic terified isola'ted, at least every I isolation valve within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months by 31 days, by at least one closed i use of at least one closed manual manual valve, blind flange, or ;

valve, blind flange, or de-activated automatic valve, !

de-activated automatic valve secured in the closed position. [

secured in the closed position.* i l

If both secondary containment !

automatic isolation valves are 't inoperable in the affected {

penetration (s), then the above l action shall be taken within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .* This requirement may be I satisfied by de-activating the i inoperable valve (s) in the closed 3. At least once per OPERATING CYCLE,- j position. demonstrate each OPERABLE j secondary containment' automatic j

3. If Specification 3.7.C.1 and isolation valve required . to be ,

3.7.C.2 cannot be met, restore closed during accident conditions SECONDARY CONTAINMENT INTEGRITY actuates to the isolated position, i within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or; on an actual. or. simulated AUTOMATIC actuation signal, within j

a. Be in at least HOT SHUTDOWN within the specified isolation time. i the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD j SHUTDOWN within the following ,

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

b. IMMEDIATELY suspend movement of i irradiated fuel assemblies in the secondary containment, all cure !

alterations, activities which :

could reduce the shutdown margin, [

and initiate actions to suspend ;

operations with the potential for draining the reactor vessel. The

-l provisions of Specification 1.0.J ,

are not applicable. j i

~

Isolation valves closed to satisfy this l requirement may be reopened on an intermittent basis under administrative ,

control.

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1 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS 3,7 D Primary Containment Automatic 4.7.D Primary Containment Automatic Isolation Valves Isolation Valves l 1. When PRIMARY CONTAINMENT INTEGRITY 1. The primary containment automatic is required, all primary isolation valves surveillance containment automatic isolation- shall be performed as follows:

valves listed in Table 3.7.1 and all instrument line excess flow a. At least once per OPERATING-CYCLE check valves shall be OPERABLE the OPERABLE isolation valves that except as specified in 3.7.D.2. are power operated and j automatically ' actuated shall - be ~

tested for simulated automatic ;

ACTUATION and closure times. j i

b. At least once per quarter: j (1) All normally open power operated isolation valves (except for the main steam line power-operated j isolation valves) shall be fu13y j closed and reopened. ;

'I (2) With the reactor power less.than 75%, trip main steam isolation .

valves individually and verify l closure time. j

c. At least once per OPERATING CYCLE.

the OPERABILITY of the reactor coolant system instrument line j excess flow check valves shall be l l verified, t

d. At least once per OPERATING CYCLE, l while SHUTDOWN, the devices that limit the maximum opening angle to 60* shall be verified = functional for the following . valves- !

PC-230MV, PC-231MV, PC-232MV, and PC-233MV. l l

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l TTuTTING CONDITIONS FOR OPFPATION SURVEILIANCE REOUIREMENTS 3.7.D (cont'd) 4.7.D (cont'd) l

2. Whenever PRIMARI CONTAINMENT 2. When it is determined that one or ,

INTEGRITY is required, but one or more primary containment automatic more penetration flow paths isolation valve (s) specified in contain an inoperable primary Table 3.7.1 required to be closed .

containment automatic isolation during accident conditions is ,

valve specified in Table 3.7.1, inoperable when PRIMARY ;

REACTOR POWER OPERATION may CONTAINMENT INTEGRITY is required, !

continue provided the the affected penetration (s) shall penetration (s), having an be verified isolated, at least ,

inoperable automatic valve, is every 31 days, by at least one (are) isolated within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months closed manual valve, blind flange, 1 (8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months for main steam line) by or de-activated automatic valve, >

the use of at least one closed secured in the closed position. ,

manual valve, blind flange, or de-activated automatic valve ;

secured in the closed position.*

If both inboard and outboard ,

containment isolation valves are inoperable in the affected :

penetration (s), then the above .

action shall be taken within one l hour.* This requirement may be ,

satisfied by de-activating the j inoperable valve (s) in the closed j position. ,

t

3. If Specification 3.7.D.1 and 3.7.D.2 cannot be met, an' orderly ;

SHUTDOWN shall be initiated and the reactor shall be in the COLD I SHUTDOWN condition within 24 !

hours.

1

Isolation valves closed to satisfy these requirements may be reopened on an intermittent basis under administrative control.

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3.7.B & 3.7.C PASES (cont'd) l t

High efficiency particulate absolute (HEPA) filters are installed before and after ;

the charcoal adsorbers to minimize potential release of particulates to the ;

, environment and to prevent clogging of the iodine adsorbers. The charcoal adsorbers !

are installed to reduce the potential release of radioiodine to the environment. The !

in-place test results should indicate a system leak tightness of less than 1 percent bypass leakage for the charcoal adsorbers and HEPA filters The laboratory carbon !

sample test results should indicate a radioactive methyl iodide removal efficiency l of ai least 99 percent for expected accident conditions. If the performance of the l HErA filters and charcoal adsorbers are as specified, the resulting doses will be ;

less than the 10 CFR 100 guidelines for the accidents analyzed.

Only one of the two Standby Gas Treatment subsystems is needed to cleanup the reactor ;

building atmosphere upon containment isolation. If one subsystem is found to be inoperable, there is no immediate threat to the containment system performance and ;

REACTOR POWER OPERATION or refueling operation may continue while repairs are being l '

. made. If both subsystems are inoperable, the plant is brought to a condition where the standby Gas Treatment System is not required. 6 i !

4.7.B & 4.7.C BASES !

Standby Gas Treatment System and Secondary Containment Initiating reactor building isolation and operation of the Standby Cas Treatment System to maintain at least a 0.25 inch of water vacuum within the secondary l l containment provides an adequate test of the operation of the reactor building l isolation valves, leak tightness of the reactor building and performance of the !

Standby Gas Treatment System. - Timing of the isolation valves verify they are closing !

within the design basis requirements as specified in the USAR. Functionally testing [

the initiating sensors and associated trip channels demonstrates the capability for i automatic actuation. Periodic testing gives sufficient confidence of reactor building ;

integrity and Standby Cas Treatment System performance capability. l l

Pressure drop across the combined HEPA filters and charcoal adsorbers of less than l 6 inches of water at the system design flow rate will indicate that the filters and j adsorbers are not clogged by excessive amounts of foreign matter. A 7.8 kw heater -

is capable of maintaining relative humidity below 70%. Heater capacity and pressure 1 drop should be determined at least once per operating cycle to show system performance capability.

The frequency of tests and sample analysis are necessary to show that the HEPA filters and charcoal adsorbers can perform as evaluated. Tests of the charcoal adsorbers with halogenated hydrocarbon refrigerant shall be performed in accordance >

with ANSI N510-1980. The test canisters that are installed with the adsorber trays ,

should be used for the charcoal adsorber efficiency test. Each sample should be at l least two inches in diameter and a length equal to the thickness of the bed. If test !

results are unacceptable, all adsorbeat in the system shall be replaced >

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.. 1 LTMITING CONDITIONS FOR OPERATION SURVEILIANCE REOUIREMENTS l 3.10.B (Cont'd) 4.10 (Cont'd)

4. During SPIRAL RELOAD, SRM ,

OPERABILITY will be verified by i using a _ portable external source .>

every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months until the required amount of fuel is loaded to maintain ;

3 cps. As an alternative to the [

above, two fuel assemblies will be ;

loaded in different cells containing !

control blades around each SRM to Ej obtain the required 3 cps. Until ;

these two assemblies have been '

loaded, the 3 cps requirement is not o necessary. ,

C. Svent Fuel Pool Water Level C. Spent Fuel Pool Water Level ->

Whenever irradiated fuel is stored ?

in the spent fuel pool, the pool When irradiated fuel is stored in .l water level shall be maintained at the spent fuel pool, the water level ;

or above 8b' above the top of the shall be recorded daily. l fuel. j D. Time Limitation Irradiated fuel shall not be handled ~

in or above the reactor prior to ;

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

E. Standby Gas Treatment System E. Standby Gas Treatment System l

^

From and after the date that one Standby Gas Treatment subsystem is When one Standby Gas Treatment _

made or found to be inoperable for subsystem becomes . inoperable , the- l any reason, movement of irradiated OPERABLE. Standby Gas Treatment l ;

fuel assemblies, core alterations subsystem shall be verified to be 'I with irradiated fuel in the vessel, OPERABLE immediately and daily l '

and operations with a potential for thereafter. A demonstration of E draining the reactor vessel with diesel generator OPERABILITY is not l _l irradiated fuel in the vessel is required by this specification. ;

permissible only during the

succeeding seven days unless such ,

subsystem is sooner made OPERABLE, i provided that during such seven days 4 all active components that affect l OPERABILITY of the OPERABLE Standby )

Gas Treatment subsystem, and its associated diesel generator, shall l be OPERABLE.

At least one diesel generator shall 'l l be OPERABLE during fuel handling l operations. This one diesel shall be capable of supplying power to an l l OPERABLE Standby Cas Treatment I subsystem.

F. . Core Standby Coolinn Systems l During a REFUELING OUTAGE, refueling operation with fue1~in the reactor vessel may continue with one Core Spray and one LPCI subsystem ,

inoperable, or with both Core Spray 1 subsystems inoperable. Refueling is permitted with the suppression

{ chamber drained provided an OPERABLE y

Core Spray or LPCI subsystem is aligned to take a suction en the condensate storage tank containing at least 150,000 gallons-(214 ft.

indicated level).

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TABLE OF CONTENTS (cont'd) l Pace No.

SURVEILLANCE LIMITING CONDITIONS FOR OPERATION REOUIREMENTS 3.5 CORE AND CONTAISMENT COOLING SYSTEMS 4.5 114 131 A. Core Spray and LPCI Subsystems A 114 B. Containment Cooling Subsystem (RRR Service Water) B 116 C. HPCI Subsystem C 117 D. RCIC Subsystem D 118 E. Automatic Depressuritation System - E 119 F. Minimum Low Pressure Cooling System Diesel Generator Availability F 120 G. Maintenance of Filled Discharge Pipe G 122 u e ng;nooroa enrogn=rse vnwnnre~ovre ennia ng u ,23 3.6 PRIMARY SYSTEM BOUNDARY 4.6 132 - 158 A. Thermal and Pressurization Limitations A 132 B. Coolant Chemistry B 133a C. Coolant Leakage C 135 D. Safety and Relief Valves D 136 E. Jet Pumps E 137 F. Recirculation Pump Flow Mismatch F 137 G. Inservice Inspection G 137 H. Shock Suppressors (Snubbers) H 137a 3.7 CONTAINMENT SYSTEMS 4.7 159 - 184 A. Primary Containment j,,,$a, A 159 B. Standby Gas Treatment System B 165 C. Secondary Containmen C /bb 1CSa D. Primary Containment solation Valves D 4.8 3.8 MISCELLANEOUS RADI0 ACTIVE MATERIAL SOURCES 185 - 190 f

3.9 AUXILIARY ELECTRICAL SYSTEMS 4.9 191-202j{

A. Auxiliary Electrical Equipment A 191 B. Operation with Inoperable Equipment B 192 3.10 CCRE ALTERATIONS 4.10 203 - 209 A. Refueling Interlocks A 203 B. Core Monitoring B 205 C. Spent Fuel Pool Water Level C 205 D. Time Limitation D 206 E. Spent Fuel Cask Handling E 206 3.11 FUF~u RODS 4.11 210 - 214e -

f A. Average Planar Linear Heat Generation Rate (APLHCR) A 210 l B. ulnear Heat Generation Rate (LHCR) B 210 C. Minimum Critical Power Ratio (MCPR) C 212

_ , , , , _ _ . _ _ i

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,-,i

h TuroKmAT30y &LY ) N fm 07 &m &wg L. Mode - The reactor mode is established by the mode selector-switch. The modes include refuel, run, shutdown and startup/ hot standby which are defined as follows:

1. Refuel Mode - The reactor is in the refuel mode when the mode switch is in the REFUEL position. When the mode switch is in the REFUEL position, the refueling interlocks are in service.

2. Run lio_de - In this mode the reactor system pressure is at or above 825 psig '

and the reactor protection system is energi zed with APRM protection (excluding the 15% high flux trip) and RBM interlocks in service.

3. Shutdown Mode - The reactor is in the shutdown mode when the mode switch is ,

in the SHUTDOWN position.

i 4 Startup/ Hot Standby - In this mode the reactor protection scram trips initiated by the main steam line isolation valve closure are bypassed, the low pressure main steam line isolation valve closure trip is bypassed, the '

reactor protection system is energized vith APRM (15% SCRAM) and IRM neutron monitoring system trips and control rod withdrawal interlocks in service.

i L.A Rormal Ventilation - Normal ventilation is the controlled process of discharging and replacing air from/to a confinement to maintain temperature, humidity, or other !

conditions necessary for personnel safety and entry. The contents of the atmosphere j being discharged from the confinement will have been established prior to establishing normal ventilation following a purging / venting operation.

M. Offsite Dose Assessment Manual (ODA!il - An OFFSITE DOSE ASSESSMENT MANUAL (ODAM) l shall be a manual containing the methodology and parameters to be used in the !

calculation of offsite doses due to radioactive gaseous and liquid effluents, calculation of gaseous and liquid ef fluent monitoring instrumentation alarm / trip setpoints, and describes the Environmental Radiation Monitoring Program.

N. Operable - Onerability - Operating l

1. Operable - Operability - A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function (s). Implicit in this definition shall be the assumption l that all necessary attendant instrumentation, controls , normal and emergency elect rical power sources (except as specified in Section 1.0.J and 3.9),

cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its function (s) are also capable of performing their related support function (s). l A verification of OPERABILITY is an administrative check, by examination of appropriate plant records ( I n r.s . turveillance test records), to determine that a system, subsystem, 1. r a i n , enmponent or device is not inoperable. Such verification does not preclude t he demonstration (testing) of a given system, subsystem, train, component et device to determine OPERABILITY.

2. Operating - Operating means a syst em, subsystem, train, component, or device is performing its intended function in its required manner.

I

0. Operatinn Cvele - Interval between the end of one refueling outage and the end of the next subsequent refueling outage.

P. Primary Containment Intrn_r_Lty - Primary cont ainment integrity means that the drywell and pressure suppression chamber are intact and all of the following conditions are satisfied:

1. All manual containment isolation valves on lines connected to the reactor coolant system or containment, and which are not required to he open during accident conditions, are closed.

? At least one door in each airlock is closed and sealed.

0a/20/91

Aeaf6 (ceMdte,]w, e Jed3_0 hA d'ce[ deti^J I

3. All CutomatihErntainmen6 isolation valves are operable

~ --

or d: 2:11 md

_n J

e isolated -" UP

Or N Penth.ltm ~- !

4. All blind flanges and manways are closed.

P.A Purce - Purcinn - Purge or Purging is the controlled process of discharging air or gas from a confinement to establish temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.

P.B Process Control Procram - The Process Control Program outlines the solidification of radioactive waste from liquid systems. It does not substitute for station operating procedures, but provides a general description of equipment, controls , and practices to be considered during waste solidification to assure solid wastes.

Q. Pated Power - Rated power refers to operation at a reactor power of 2381 megawatts thermal. This is also termed 100% power and is the maximum power level authorized by the operating license. Rated steam flow. rated coolant flow, rated neutron flux, and rated nuclear system pressure refer to the values of these parameters when the reactor is at rated power.

R. Paactor Pnwer Operation - Reactor power operation is any operation with the mode switch in the "Startup/ Hot Standby" or "Run" position with the reactor critical and above 1% rated power. ,

i S. Peactor vessel Prersure - Unless otherwise indicated, reactor vessel pressures listed in the Technical Specifications are those measured by the reactor vessel steam space detectors.

T. gjefunlinc Oura re - Refueling outage is the period of time between the shutdown of

"'"the uni t prior to a refueling and the startup of the plant after that refueling.

t U. Saferv 1.imits - The safety limits are limits within which the reasonable maintenance of the fuel cladding integrity and the reactor coolant system integrity are assured.

Violation of such a limit is cause for unit shutdown and review by the Nuclear Regulatory Commission before resumption of unit operation. Operation beyond such a limit may not in itself result in serious consequences but it indicates an operational deficiency subject to regulatory review v Seenndarv Gntainn nt Interrim - Secondary containment integrity means that the reactor building is intact and the following condi: ions are met:

1 At least one door in cach access openinr is c l c,s e d .

ope rablTrshra

+ . b.wsw,lle M_*

n- h _bd h"iaf- - -

2. The standby gas treatment s. stem is -

All * - * ' - t *r- isolation .alvesiarc[ operable or _.__ __ i-

m1a:ed --

c,,gg{ { m4, Q;g,g[y [,,J7, m 9utdown The reactor is in a shutdown conditier when the mode switch is in the

" Shutdown" or " Refuel

position.

1. Hot Shutdcwn means conditions as above with resctor coolant temperaturc greater than 212'F.

2. Cold 3hutdown means conditions as above uith reactor coolant temperature equal to or less than 212*F and the reactor vessel tented.

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i LlMITTUG CONDITION FOR OPERATION 97RVEILLANCE REOUTREMENT ,

3.7.A (cont'd.) 4. 7. A (cont' d. )

6. iow-Low Set Relief Function 6. Low-Low Set Relief Function ,

a. The low-low set function of the a. The low-low set safety / relief valves saferv-relief valves shall be shall be tested and calibrated as ;

/Sheraole) when there is irradiated specified in Table 4.2.B.

-tuel in the reactor vessel anc tne ;

reactor coolant temperature is 2 212*F, except as specified in 3.7.A.6.a.1 and 2 below. 3

1. With the low-low function of one safety / relief valve (S/RV) inoperable, restore the inoperable !

6 LLS S/RV to OPERABLE within 14 days or be in the HOT STANDBY mode 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 i 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

2. With the low-low set function of both S/RVs inoperable, be in at l least HOT STANDBY within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months i

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

b. The pressure switches which control ;

the low-low set safety / relief valves shall have the following settings.

NBI-PS-51A Open Low Valve 1015 20 psig (Increasing)

NBI-PS-51B Close Low Valve 875 20 psig (Decreasing)

NBI-PS-51C Open High Valve /M 4.7.8 J. c'- o,- m;M J.7 8 - j 1025 20 psig (Increasing) .

-u !

B. Standbv Cas Trearw nt su rem NBI-PS-51D Close High Valve 875 20 psig (Decreasing)

1. At~1 east once per W erat N cycle] !

de f dng condmons snau be -

B. Standbv Cas Treatment Svstem ALL. dera nstrated.

CAPF,

1. Except as specified in 3.7.B.3 ! a. Pressure drop across the combined g below, both Standby Cas Treatment HEPA filters and charcoal adsorber p subsystems shall be threradiDat al banks is'less than 6 inches of water t hes when (seconcarv cont;ainmeg at the system design flow rate.

QegritvMs required.

2.a. Ihe results' ot the in-place colci DUr % b. Inlet heater input is capable ? of i leak tests on the HEPA filters shall reducing R.H. from 100-to 70% R.H.

u show >_99% DOP removal. The results C.. - A J A %,. K R. mwt Pm l of the halogenated hydrocarbon leak 2.a. The tests . and sample analysis of tests on the charcoal adsorbers Specification 3.7.B.2 shall be-M e

kW shall show 299% halogenated performed at least once every i 33 871' hydrocarbon removal. The DOP and 18 months for standby service or 9 !

I halogenated hydrocarbon tests shall after every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of system i q.M."AA be performed at a Standby Gas operation and following significant- ,

Treatment flowrate of $1780 CFM and painting, fire . or chemical release .j 4 any zone-

.p p. .

at a Reactor _ Building pressure of in

**""i"**i"E ventilation "ith th* 57****'

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d lo.u n .f d_ W *M -

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- i REVISED WORDING FOR PAGE 165. CNS TECHNICAL SPECIFICATION SECTIONS 3/4.7.B l t

Add new SR 4.7 B.1.c m_ j

c. Demonstrate that each Standby Gas Treatment subsystem can maintain 2 0.25 '

j inch of water vacuum in the secondary containment for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months at a flow '!

rate of s 1780 CFM under calm wind conditions- (2<p<5 mph where 4 . average -;

wind speed), ,

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LIMITING CONDITION FOR OPEFATION SURVEILM'CE PEOUTDWENT ,

I J . 7.B (cont' d) 4.7.B-(cont'd) ,l

b. The results of laboratory carbon b. Cold DOP testing shall be performed sample analysis shall shov 299% after each complete or pertial radioactive methyl iodide removal replacement of t.he HEPA filter bank ;

or after any structural maintenance with inlet conditions of: velocity

227 FPM. 21.75 mg/m 3 inlet methyl on the system housing.

iodide concentration, 270% R.H. and

$30*C, c. Halogen'ated hydrocarbon testing i shall be performed after each '

c. Each fan shall be shown to provide complete or partial replacement of 1780 +1 %. the charcoal adsorber bank or after F N_ b any structural maintenance on the system housing. .

3. From a af ter the d e that one '

( Standby Gas Treatmentj subsystem is made or found to be fnoperable for d. Each subsystem shall be- or,erated with the heaters on ar. ; least .

any reason, creactor v operati onr is ~

1 10 g rs every :

permissible -only during the 4 -

succeeding seven days unless such subsystem is sooner made fonerant0>- e. Test sealing of gaskets for housing .

4 provided that during such seven days doors downstream of the HEPA filters a

all active components that, affect and charcoal adsorbers shall be l Emernoili tvlof the rbinerEoi o standby performed at, and in conformance Gas Treatment subsystem, and its with, each test performed for 7['

compliance with Specification I associated diesel generator, shall J be(perno]G. 4.7.B.2.a and Specification 3.7.B.2.a.

Fuel handling require. :s are ,

specified in Specificatio. 1.10.E. 7

3. System drains where present shall be [

T inspected quarterly for adequate i 4 'T -.d1:1 - . . _:

--r '

N water level in loop-seals

>roceau . shall b m tiatec __

4.a. At least once per Caperannz cve rs D l mmediately t a lish reactoi ondit r which t dby Gas automatic initiation oT each Standby -

7

- . ( . :- - - - .: 1 ::;____', Gas Treatment subsystem shall be i demonstre.ted.

b,&gh Mk $Mpvd neak M.

I b. At'least once ner Foernr W eve lem _

manual (opernniii tv1 of the bypass val"e for~ iilter cooline a shall be 3 KLL (AW demonstrated. ed sND) 64 7<. 4-+ 'tJ datadm i -

c. When one Standby Gas Treatment p 3 subsystem becomes inoperable, the ,

n eraoB0 Standby Gas Treatment i fnen k b 3.7,8.4 19.7.8. 9 4' W subsva em shall _be verified to be f cerab fimmediatelv7 and daily ,

(b Ir#* Me. 165 b. thereaf ter. J' A demonstration of i l

4

%f diesel generatorToperabilit31s not required by this spbcification. ;

ND- -

Secondarv Containment .,

l' Secondarv Containment

l. Secondary containment surveillance

1. 6econdary T_-thent_i inreeritv3 shall be performed as indicated - ,

shall be maintained during all modes below:

1 of plant operation exceptswhen all f 7 p ;

of the followhnditionsjare met.

& sp a w a g.7.c.2. o r )

l l% $xdm 1.2.C $ H.7.c To W :

b

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4 REVISED WORDING FOR PAGE 165a. CNS TECHNICAL SPECIFICATION SECTIONS 3/4.7.B ,

Revise LCO 3.7.B.4 to read

~ _- _ _

4. If the conditions of LCO 3.7.B.1, 2, or 3 cannot be met, be in at least i HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and COLD SHUTDOW within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . -

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oae r. men

a. The reactor is suberitical and m, A preoperational secondary Specification 3.3.A is met. ,gg p].u containment capability test shall be wth p@ onducted after isolating the

b. The reactor water temperature is y A eactor building and placing either below 212*F and the reactor coolant g g tandby Gas Treatment subsystem) system is vented. 1 4 n, filter train in operation. Such orenehn : Avests shall demonstrate the

c. No activity is being performed whien m. robb) capability to maintain le4 -inch of can reduce the shutdown margin below fo r 4,.;,q vacuum under calm wind that specifien in Specification 4 pg, w(ater2<E<5 mph) conditions wit {h a filter 3.3.A. - %^ .ren,g ,re $ rain flow rate of not more than b- W M '8

tu ;

10 of building volume er day.

d. No irradiated fuelb being Lanc.c; i tg fta A;p0%, wind speed) '

gy in the secondary containment,9(d -- [t 6ff *Nr 0 l le * -hieb ceuld petenti ll; damage o. Additional tests shall be performed

'irredicted fuel :: being : c:d c during the first o peratine c ve t e.)

6 c e r ender" r ent2in~: nt n e' an adeouate number oi fnder different environmental wind

- .5ee ne rt f ate. f e #EW LCO 2.7 C

2 3- If : -izi r" - n _ . _ ., .. .

g conditions to enable valid zg7 _-1.

. m. restore extrapolation of the test results.

E e coMia ry m ininment interrit" . a, 5" within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or: .. Secondary c ntainment capability to N M 7.7.C,143 To f9 /A. /s(* Naad-- __ _

maintain 's inch of water vacuum

a. Be in at least Mot Shutdownh under calm wind (20.00333 hours 1.984127e-5 weeks 4.566e-6 months and conditions with a filter train flow in r+ t ri dined _own) wi thin the rate of not more than 100% of building volume per day, shall_be husECIsmpfo11owing 24 urs nstrated at eacN me nne

b. # Suspend dia ed fuel -

utarlD prior to refueling.

k m ., a 1 4 m. -mv.,*'-- in the b'_I secondary containment. d.

m m _.

_._ __ [' N_

- mm - c 1 m m mu .x . ~ . , ,

; ,,u ; G rio

ion is determined, the St- ob

_._.1a3. u ._

O ~i? 4 ,. _ , . , . . ,

das i tment System b - j)}

Perated .ediately ter th' :

% ,,. A .k 4 -

M all c$e Recku f * *

d ** * #* ^^ * *

% 4, m 4. alterations a.r.d activities tardl*8 the secondar-A emainder of

, which coul~d ~ reduce the w I swe.a Yimba shutdown .

.ontainment t confl.~ its abilit' Marginn M The g gg,g- & r of 6

)A.pmr,b..I tu / t w +" ~ provisions of Specz '_ cation 6 wey4 6

ec nd- f containment at 1/ nch o:

& * ' g 1.0.J are not applicatue . P ara . : rat negative pressure under 1:

E Y 5ee O -Le belem l f h e Nk(>a.. - -

AJJ ne e JR. 4.1.C."1 l 3, MM f*W #66a..

. b&

(O_

_ D. Prifrary Contain-enth sol ati on Valve s M D. _Primarv Containment. Isolation Valves Ma(flFW_ G8M AIJJMUT DMEatTTY la ema.O Mrm deCL .

The primary containment rE'r r-r : 7 ; m - i-.; 1. isolation 1.Q""7 Hi mr . alla isolation valves valves surveillance shall be listed in Tabl ' 7.1 and all performed as follows:

instrument line fl ow check valves Ja. east nee pen- eMM m@ j shall be Conerable) except as tne merab13 isolation valves that ;

specified in 3.7.D.2. CWS are power operated and automatically e ,

j '"itirred shall be tested for

- <G.TM v l. , closel 4. s.A:4 41J4 __

~~

' Dims.A5 h init'= tier

/ closure times.

and ;

r**y ka. reep nnel ea, rtyIki,tew t hed mder alminhN tv6 " l l i I e-

% b= 4u - = - '

ACTUATT.0M T h.SEcrzey _TcM.c.tdf Wd V a

3/4. 7. D 4 e Pe w IG 7 -166- , , . , , ,

REVISED WORDING t '1R PAGE 166. CNS TECHNICAL SPECIFICATION SECTIONS 3/4.7.C Add New LCO 3.7.C.2 ~

__ + -

2. Whenever SECONDARY CONTAINMENT INTEGRITY is required, but one or more penetration flow paths contain an inoperable secondary f

)

containment automatic isolation valve required to be closed during l accident conditions, then isolate the affected penetration (s), ,

having an inoperable automatic isolation valve, within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months by use of at least one closed manual valve, blind flange, or de-activated automatic valve secured in the closed position.* If both secondary containment automatic isolation valves are inoperable in j the affected penetration (s), then the above action shall be taken [

within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .* This requirement may be satisfied by de-activating the inoperable valve (s) in the closed position.

- _ .)

Replace existing Surveillance Requirement 4.7.C.l.d wording with new SR wording 7

d. When the integrity of the pressure retaining boundary of secondary 1 containment has been lost during POWER REACTOR OPERATION or HOT SHUTDOWN, then SECONDARY CONTAINMENT INTEGRITY shall be verified through qualitative leak testing or evaluation of the affected area prior to declaring the SECONDARY CONTAINMENT INTEGRITY restored.

L -

Add New Surveillance Requirements 4.7.C.2 and 4.7.C.3 _

q

2. When it is determined that one or more secondary containment !

{

automatic isolation valve (s) required to be closed during accident .

I conditions is inoperable when SECONDARY CONTAINMENT INTEGRITY is required, the affected penetration (s) shall be verified isolated, at I least every 31 days , by at least one closed manual valve, blind

{ flange, or de-activated automatic valve, secured in the closedi '

position. _

m- _

3. At least once per OPERATING CYCLE, demonstrate each OPERABLE

{ secondcry containment automatic isolation valve required to be I closed during accident conditions actuates to the isolated position, on an actual or SIMULATED AUTOMATIC ACTUATION signal, within the specified isolation time.

w ,- _

l

. +. ..

LIMITINC cnNDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS

.2 , . ' . ~ . e .. : .'

b. At least once per quarter: j (1) All normally open power operated isolation valves (except for the i main steam line power-operated .l isolation valves) shall be fully closed and reopened. ,

(2) With the reactor power less than 75%, trip main steam isolation valves individually and verify.

closure time.

c. At_ least once per rove r at inc eveie3 e tne Tonerabilit?) of the reactor coolant system Instrument line flow ll check valves-shall be verified. g

d. At least once per A m m W cvc i M, 8

). ;

while Tshurciowr3 the devices that 4 p limit iihe maximum opening angle to q '> ;

60* shall be verified functional for '

PC-230MV, the following valves:

PC-231MV. PC-232MV. and PC-233MV.

[c

2. '-

- geplau, ulk Nw- keNinge

-: . c.,

are(f*H.

' :1;:_;r .. 2. ' -- : 1r I;:1 M :- 11 :: !!m"\

\) ,

'pe 4ed in Table 3.7.1 mes Ln Ta '

.1 is ino - , the 3 noperab reactor pow - peratiori >osition of at - me other valve l' aay continue . at least one n eac having an .

ble ,

alve in e having ar 'll 1: . : : ; r d ; d i._ il, .

Lnoperab alve shall . in the Rag w;)k w ved;p ,w,/

g aode tresponding to the is ec ,. q l f ~ . . . .

3. If Specification 3.7.D.1 and 3.7.D.2 y 'N pil ,

cannot be met, an orderly G utcown fpt i J[df.7,0.gi3eg shall be initiated and the reactor shall be in the (Gold Shutdown g neh 4e l condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

tu w" W '

h

Isolation valves closed to satisfy IO*-

% these requirements may be reopened on an intermittent basis under y administrative control.

-k 9

-167- . ,, , 2 'p

. u .

REVISED WORDING FOR PAGE 167. CNS TECHNICAL SPECIFICATION SECTIONS 3/4.7.D Replace existing LCO 3.7.D.2 with new LCO 3.7.D.2 _ _ __-

_~ _

2. Whenever PRIMARY CONTAINMENT INTEGRITY is required, but one or more penetration flow paths contain an inoperable primary containment automatic isolation valve specified in Table 3.7.1, REACTOR POWER OPERATION may continue provided the penetration (s), having an inoperable automatic valve, is (are) isolated within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months (8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months for main steam line) by the use of at least one closed manual i valve, blind flange, or de-activated automatic valve secured in the closed position.* If both inboard and outboard containment isolation valves are inoperable in the affected penetration (s), then the above action shall be taken within one hour.* This requirement ,

may be satisfied by de-activating the inoperable valve (s) in the closed position .- _

Replace existing SR 4.7.D.2 with new SR 4.7.D.2

2. When it is determined that one or more primary containment automati isolation valve (s) specified in Table 3.7.1 required to be closedi j

during accident conditions is inoperable when PRIMARY CONTAINMENT INTEGRI'li is required, the affected penetration (s) shall be verified isolated, at least every 31 days , by at least one closed manual )

I valve, blind fictge, or de-activated automatic valve, secured in the closed position.

i s

. u .

P 3.7.B & 3.7.C BASES (cont'd)

High efficiency particulate absolute (HEPA) filters are installed before and after the charcoal adsorbers to minimize potential release of particulates to the environment and to prevent clogging of the iodine adsorbers. The charcoal adsorbers are installed to reduce the potential release of radiciodine to the environment. The in-place test results should indicate a system leak tightness of less than 1 percent bypass leakage for the charcoal adsorbers and HEPA filters. The laboratory carbon sample test results should indicate a radioactive methyl iodide removal efficiency of at least 99 percent for expected accident conditions. If the performance of the HEPA filters and charcoal adsorbers are as specified, the resulting doses will be less than the 10 CFR 100 guidelines for the accidents analyzed.

Only one building of the twoupon atmosphere Standby Gas Treatment containment isolation.subsystems is needed If one subsystem to cleanup is found to be the re inoperable, there is no immediate threat to the containment system performance and "

--ar 3:r**- or refueling operation may continue while repairs are being made.

If both subsystems are inoperable, the plant is brought to a condition where the 9 Standby Gas Treatment System is e m ui 1 E ACTO k They ed a e> W lt* 50^ ** I"'

t!#3 O'O h'

c. 'eg l'q baalj 4.7.B & 4.7.C #ASES. 0 Weft, e#.s.

winin W dngp %A, US Ad.

_ Eft,KtIoM f A9 ad ra mmfr ^*_ if reci-- f f.a k^ in ^=

Standby Gas Treatsent System and Seconcary containhent - - -

Initiating reactor building isola ion Tua operation "

f the Standby Gas Treatment System to maintain at least a , ine wr vacuum within the secondary containment provides an adequate

of the operation of the reactor building isolation valves , leak tightne of the reactor building and performance of the Standby Gas Treatment System. Functionally testing the initiating sensors and actuation. y associated trip channels demonstrates the capability for automatic Periodic testing gives sufficient confidence of reactor building integrity and Standby Gas Treatment System performance capability. M Pressure drop across the combined HEPA filters and charcoal adsorbers of less than 6 inches of water at the system design flow rate will indicate that the filters and-adsorbers are not clogged by excessive amounts of foreign matter. A 7.8 kw heater is capable of maintaining relative humidity below 70%. Heater capacity and pressure drop should be determined at least once per operating cycle to show system performance capability.

The frequency of tests and sample analysis are necessary to show that the HEPA filters and charcoal adsorbers can perform as evaluated. Tests of the charcoal adsorbers with halogenated hydrocarbon refrigerant shall be performed in accordance with ANSI N510-1980. The test canisters that are installed with the adsorber trays should be used for the charcoal adsorber efficiency test. Each sample should be at least two inches in diameter and a length equal to the thickness of the bed. If test remits are unacceptable, all adsorbent in the system shall be replaced

-182- af =

.a.

~. - .

. o'5.. . s l

j~ l

)

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUTREMENTS $

~

3.10.B (Cont'd) - 4.10 (Cont'd) !

4. Durinz [# iral Ieloag SRM CoperabilievT vflI be verified by using a portable external source )

every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months until the required i amount of fuel is loaded to maintain 4 3 cps. As an alternative to the klb l above, two fuel assemblies will be CAfa. j loaded in different cells containing !

control blades around each SRM to i obtain the required 3 cps. Until i these two assemblies have been ;

loaded, the 3 cps requirement is not l necessary. ~

.l C. Spent Fuel Fool Water Level C. Spent Fuel Pool Water Level " l Whenever irradiated fuel is stored .

j in the spent fuel pool, the pool When irradiated fuel is stored .in water level shall be maintained at the spent fuel pool, the, water level or above 8h' above the top of the shall be recorded daily, fuel ,

awra%.

D. Time Limitation Irradiated fuel sh- 1 not be hand in or above th reactor _ priot /o l-24 hours after reactor 6hutoo E. Standbv Gas. reatment System #p E. Standbv Gas Treatment System From and 'fter the date at one

$ Standby as Treatment s system'is When one Standby Gas subsystem becomes inoperable,Treatment the < $l

~

made o found to be i perable for any ason, h2Mling of irradiated W erable3 Standby Gas Treatment 8 3 fueld " n: m nr rf 1: "- -tirh subsistem shall be verified ' to be c al f p:::nti:11; "-- ;: frrr' int " foperab1D immediately' and ' daily ful ir S: ::: ' 7 - " - the4 q tfiereaf ter. A demonstrationL of permissible .only during diesel generatoffoperabilits is~not-succeeding seven days unless_such required by this specification,_ _

y subsystem is sooner made moer_abl h 4 " -

~

provided that during such seven days % 7 all active , components that' affect JQp gg gMMpj

)f Goer 2binty7Dt theWerabTD Standby AMd

\

Gas Treatment sub system, and its j [uAl(A k MI associated ciesel generttor, shall be @erabh t " I Y PA dg "~ 8Mg j I

" l At le}st onedesel generator shall be Querable] during fuel handling drigl dig ~ b -

I operations. This one diesel shall ten d jerad M M g4(4,agaerl be capable of supplying power to an Copera51e3 Standby Gas Treatment f F. subsystem >

Core Standbv Cool'ne Systetr During aCretueline ourmyWefueling operation with fuel in the reactor .

vessel may continue with one Core Spray and one LPCI subsystem

(<[

/

inoperable, or with both Core Spray subsystems inoperable. Refueling is permitted . with the supuression s -

chamber drained provided an/boerabif e Core Spray or LPCI subsystem is aligned to take a suction on the condensate storage tank containing at least 150,000 gallons (214 ft.

indicated level).

-205a-

~

. _ - a X ,

AM..2,7-,'^ E TE ^'

ML20056G582

Text

SUMMARY

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