ML023300276

Request for Amendment to the Technical Specifications Changes to Post-Accident Monitoring Instrumentation Requirements

ML023300276
Person / Time
Site:	Pilgrim
Issue date:	08/19/2002
From:	Bellamy R Entergy Nuclear Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
2.02.072
Download: ML023300276 (37)

Text

? Entergy Entergy Nuclear Generation Co.

Pilgrim Nuclear Power Station 600 Rocky Hill Road Plymouth, MA 02360 Mike Bellamy Site Vice President August 19 ,2002 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Station O-P1 -17 Washington, DC 20555-0001

SUBJECT:

Entergy Nuclear Operations, Inc.

Pilgrim Nuclear Power Station Docket No. 50-293 License No. DPR-35 Request for Amendment to the Technical Specifications Changes to Post-Accident Monitoring Instrumentation Requirements

REFERENCE:

NUREG 1433, Standard Technical Specifications for General Electric Plants, BWR/4 LETTER NUMBER: 2.02.072

Dear Sir or Madam:

In accordance with the provisions of 10 CFR 50.90, Entergy Nuclear Operations, Inc. (Entergy) is submitting a request for an amendment to the Technical Specifications (TS) for Pilgrim Nuclear Power Station. This request is to change the requirements associated with post accident monitoring (PAM) instrumentation and make Pilgrim's Technical Specifications more consistent with the requirements in the standard technical specifications for boiling water reactors.

The Enclosure contains supporting information and a safety assessment of the proposed changes, including a determination of no significant hazards consideration. Attachment 1 is a marked-up version of the current technical specifications.

Entergy has reviewed the proposed changes using the criteria of 10 CFR 50.92 and concludes that the proposed changes do not involve a significant hazards consideration.

Entergy requests approval of the proposed amendment by March 1, 2003 to support Pilgrim's upcoming refueling outage (scheduled to commence on April 19, 2003). Once approved, the amendment will be implemented within 60 days.

202072

Entergy Nuclear Operations' Inc .'LetterNumber: 2.02.072 Pilgrim Nuclear Power Station Page 2 if you have any questions, please contact Mr. Bryan Ford at (508) 830-8403.

I decll~e under penalty of perjury that the foregoing is true and correct. Executed cn the z7-L day of August 2002.

Sinzerely, eobe~r melZy

Enclosure:

Evaluation of the Proposed Changes -11 pages : Marked-up Proposed Technical Specifications and Bases Changes 20 pages

2. List of Regulatory Commitments- 1 Page cc:

Regional Administrator, Region I Mr. Robert Hallisey U.S. Nuclear Regulatory Commission Radiation Control Program 475 Allendale Road Commonwealth of Massachusetts King of Prussia, PA 19406 Executive Offices of Health and Human Services Mr. Travis Tate, Project Manager 174 Portland Street Project Directorate I Boston, MA 02114 Division of Licensing Project Management Office of Nuclear Reactor Regulation Mr. Steve McGrail, Director U. S. Nuclear Regulatory Commission Massachusetts Emergency Management Mail Stop: O-8B-1 Agency Washington, DC 20555-0001 400 Worcester Road Framingham, MA 01702 Senior Resident Inspector U. S. Nuclear Regulatory Commission Pilgrim Nuclear Power Station 600 Rocky Hill Road M/S 66 Plymouth, MA 02360 202072

ENCLOSURE Evaluation of the Proposed Changes

Subject:

Changes to Post-Accident Monitoring Instrumentation Requirements

1. DESCRIPTION

2. PROPOSED CHANGES

3. BACKGROUND 3.1 Post-Accident Monitoring Instrumentation 3.2 10 CFR 50.36(c)(2)(ii) and NRC Final Policy on the Improved Technical Specifications

4. TECHNICAL ANALYSIS

5. REGULATORY SAFETY ANALYSIS 5.1 - No Significant Hazards Consideration

6. ENVIRONMENTAL CONSIDERATION

7. COORDINATION WITH OTHER PENDING CHANGES

8. REFERENCES

A

1. DESCRIPTION Entergy Nuclear Operations, Inc. (Entergy) proposes to amend the Technical Specifications (TS) to change the requirements associated with post-accident monitoring (PAM) instrumentation.

Entergy proposes to update the TS to:

a. reflect post-accident instrumentation for the TS funct;ons meetng the guidance of NRC Regulatory Guide 1.97; and

b. adopt standard limiting conditions for operations, action statements, allowable outage times, surveillance requirements, and reporting requirements for PAM instrumentation.

These changes will make Pilgrim's TS more consistent with the requirements in the Standard Technical Specifications for General Electric Plants, BWR/4, NUREG-1433 (Reference 1).

2. PROPOSED CHANGES The following changes are proposed as part of this TS amendment.

A. Titles Change the title of Specification 3/4.2.F from "Surveillance Information Readouts" to "Post-Accident Monitoring Instrumentation." Modify instrument function names.

B. Applicability of the Requirements Change the conditions during which the requirements of Specification 3/4.2.F are applicable to when the unit is in the Run Mode or the Startup Mode.

C. Changes to Columns on Table 3.2.F Delete the "Instrument #" and "Type Indication and Range" Columns from Table 3.2.F.

Also add an identifying number to each Function and move the "Minimum # of Operable Instrument Channels" column to following the Function name column.

D. Required Action for One Inoperable Channel out of Two Channels (Excluding the Drywell High Radiation Function)

Change the required Action for one of two requirci- channels of a Function being inoperable. The current Actions (current Table 3.2.F Notes (1) and (3)) require restoring the inoperable channel to operable status within 30 days or placing the unit in Cold Shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> while Specification 3.7.A.7.c requires restoration within 7 days or placing the unit in Hot Shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. These Actions are changed in proposed Table 3.2.F Note (1) to restoring the inoperable channel to operable status within 30 days or submitting a special report within 14 days.

E. Required Action for One Inoperable Channel out of Two Channels for the Drywell High Radiation Function Change the required Action for one of two required channels of the Drywell High Radiation Function being inoperable. The current Action (current Table 3.2.F Note (7))

requires restoring the inoperable channel to operable status within 7 days or submitting 1

a special report within 14 days. This Action is changed in proposed Table 3.2.F Note (1) to restoring the ijo'pe'erable channel to operable statfu.within 30 days or submitting a special report within 14 days. Two inoperable Drywell High Radiation Function channels are addressed by the proposed Note (3).

F. Required Action for No Control Room Indication of a Funct;on (Excluding the Containment H2 Analyzer and Drywell High Radiation Functions)

Change the required Action for having no indication in the control room for a Function.

The current Actions (current Table 3.2.F Notes (2) and (3)) require restoring control room indication within 7 days or placing the unit in Cold Shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. This Action is changed in proposed Table 3.2.F Note (2) to restoring control room indication within 7 days or placing the unit in Hot Shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

G. Required Action for No Control Room Indication of the Containment H2 Analyzer Function Revise the limiting condition for operation from 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, consistent with standard Technical Specifications. Move the current requi.-ed Action for having no indication in the control room (both required channels inoperable) of the Containment H2 Analyzer Function from Specification 3.7.A.7.c to proposed Note (4).

H. Deletion of Containment H2 Analyzer Function from TS 3/4.7.A.7.c Delete the Containment H2 Analyzer from Specification 3/4.3.7.A.7.c and add the requirements to Table 3/4.2.F.

Non-Regulatory Guide 1.97, Type A or Category 1, non -Type A, Variables Delete the Suppression Chamber Air Temperature, Drywell/Torus Differential Pressure, Safety/Relief Valve Position, Safety Valve Position Indicator, Tail Pipe Temperature Indication, Reactor Building Vent, Main Stack Vent, and Tu;bine Building Vent Functions from Specification 3/4.2.F.

J. Instrument Check Frequency Change the required frequency of instrument checks to once per 31 days.

K. Calibration Frequency Presentation Present the calibration frequencies consistently as once per 24,months for those functions with current requirements that efiectively require the cal;bration once per 24 months.

L. Number of Required Neutron Monitoring Channels The number of required neutron monitoring channels is increased from 1 to 2.

M. Instrument Channels Addressed y the Specification 3/4.2.F requirements With the removal of the "Instrument #" column fom Table 3.2.F, the TS will no longer control the specific instrument numbers comprising each required Function. The specific instruments comprising each required Function is proposed to be modified as necessary 2

to reflect the currently credited Regulatory Guide 1.97;,Type A or Category 1, non-Type A, instruments '*"*." "*

N. Delete Note (2) on page 3/4.2-41 Delete Note (2) on TS page 3/4.2-41 to reflect that it is no Iongor used.

These proposed TS changes are indicated in Attachment 1. Also included in Attachment 1 for information are the conforming Bases chaiiges, which will be made afte," approval of the proposed changes.

3. BACKGROUND

1. Post-Accident Monitoring Instrumentation The primary purpose of the PAM instrumentation is to display plant variables that provide information required by the control room operators during accid&-nt situations. This information provides the necessary support for the operator to take the manual actions for which no automatic control is provided and that are required for safety systems to accomplish their safety functions for design basis events.

The purpose of PAM instrumentation for Regulatory Guide 1.97, Type A variables is to provide information so that the control room operating staff can:

" Perform the diagnosis specified in the Emergency Operating Procedures (EOPs).

These variables are restricted to preplanned actions for the primary success path of Design Basis Accidents (DBAs), (e.g., loss of coolant accident (LOCA)), and

"* Take the specified, preplanned, manually controlled actions for which no automatic cortrol is provided, which are required for safety systems to accomplish their safety function.

The purpose of PAM instrumentation for Regulatory Guide 1.97, Category 1, non-Type A, variables is to Drovide information so that the control room operating staff can:

" Determine whether systems important to safety are performing their intended functions;

" Determine the potential for causing a gross breach of the barriers to radioactivity release;

"* Determine whether a cross bac1i-of-a barrier has occurred; and

"* Initiate action necessary to proiect the public and for an estimate of the magnitude of any impending threat.

Information on Pilgrim's post-accide~nt monitoring instrumentation can be found in Pilgrim's Final Safety Analysis Rerort (FSAR) Section 7.8.10, "Post-Accident Monitoring."

FSAR Table 7.8-3 lists PAM panel instrument specifications. The NRC issued two SERs 3

for Pilgrim regarding 'post-accident instrumentatiori anl Regulatory Guide 1.97 (References 2 and 3).

The operability of? the accident monitor~ng instrumeni'tation ensures th3t there is sufficient information available on selected plant parameters to monitor and asses' plant status and behavior following an accident. This capability is consistent with th.

recommendations of Regulatory Guide 1.97.

2. 10 CFR 50.36(c)(2)(ii) and NRC Final Policy Statement The NRC Final Policy Statement on the Improved Technical Specifications (ITS), states that the purpose o the technical specifications is to impose those conditions or limitations on reactor operations necessary to obviate the possibility of an abnormal situation or event giving rise to an immediate threat to the public health and safety by identifying those featuies which are of controlling importance to safety and establishing on them certain conditions of operation which cannot be changed without prior Commission approval. To achieve that purpose, the final policy statement provides four criteria that delineate those constraints on design and operadion of nuclear power plants that must be retained in the technical specifications. These criteria were subsequently incorporated into 10 CFR 50.36(c)(2)(ii).

Criterion1 - Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary. (10 CFR 50.36(c)(2)(ii)(A))

Criterion2- A process variable, design feature, or operating restriction that is an initial condition of a design basis accident (DBA) or transient analysis that either assumes the failure of or presents a challenge to the -ntggrity of a fission product barrier. (10 CFR 50.36(c)(2)(ii)(U))

Criterion.'- A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a DBA or tr insient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. (10 CFR 50.36(c)(2)(ii)(C))

Criterion4 - A structure or component which operating experience o., probabilistic safety assessment has shown to be significant to public hralth and safety. (10 CFR 50.36(c)(2)(ii)(D))

As discussed in NUREG-1433, the PAM instrumentation thAt mo, itors the Regulatory Guide 1.97, Type A and Category 1, non-Type A, varir.ble, is the monitoring instrumentation which meets thl 9creening criteria in the final po~icy statement and 10 CFR 50.36(c)(2)(ii).

4. TECHNICAL ANALYSIS The following is the justification for the changes propc-:'* *" part of this TS amendment.

A. Titles The changes in the title of Specification 3/4.2.F froni Surveillance Information Readouts to Post-Accident Monitoring Instrumentation and function names are editorial changes to reflect the function of the instruments controlled. These changes do not result in any change in requirements.

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B. Applicability of the Requirements The current Specifications have various implied conditi6ns during which the requirements were applicable by having different required Action end states for different Functions. These various implied applicabilities are made consistent and changed to requ: ",. t;le Functions to be operable when the unit is in the Run Mode or the Startup Mode.

Having the PAM instrumentation required to be operable in the Run and Startup Modes provides the appropriate controls on the instrumentation. These variables a:e related to the diagnosis and preplanned actions required to mitigate DBAs. The appicable DBAs are assumed to occur in Run or Startup Modes. In the other plant operational modes, plant conditions are such that the likelihood of an event that would requir3 PAM instrumentation is extremely low; therefore, PAM instrumentation should not be required to be operable in these modes.

C. Columns on Table 3.2.F Deleting the "Instrument #" and "Type Indication and Range" Columns from Table 3.2.F, removes design details which are more appropriately controlled by plant procedures.

Removal of these details is consistent with NUREG-1 433.

Adding an identifying number to each Function and moving the "Minimum # of Operable Instrument Channels" column following the Function name column are editorial changes.

These changes do not result in any changes in the TS requirements.

D. Required Action for One Inoperable Channel out of Two Channels (Excluding the Drywell High Radiation Function)

The proposed 30-day allowed out-of-service time for one inoperable channel out of two required channels for a Function with subsequent reporting to the NRC provides the appropriate required Action for the condition. The 30-day completion time is appropriate based on operating experience and takes into account the remaining operable channel, the passive nature of the instrument (no critical automatic action is assumed to occur from these instruments), and the low probability of an event requiring PAM instrumentation during this interval. The required action of submitting a report in lieu of the current shutdown requirement is appropriate, since alternative actions are identified as part of the report, there is no loss of functional capability, and given the likelihood of plant conditions that would require information provided by this instrumentation.

E. Required Action for One Inoperable Channel out of Two Channels for the Drywell High Radiation Function The proposed 30-day allowed out-of-service time for one inoperable channel out of two required channels for the Drywell High Radiation Function instead of the current 7 days prior to reporting the condition to the NRC provides the appropriate required Action for the condition. The 30-day completion time is appropriate based on operating experience and takes into account the remaining operable channel, the passive nature of the instrument (no critical automatic action is assumed to occur from these instruments), and the low probability of an event requiring PAM instrumentation during this interval.

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F. Required Action for No Control Room Indication of a Function (Excluding the Containment H2 Analyzer and DrywlI High Radiation Functions)

The proposed required Actions modify the end state of the'unit if there is nc indication in the con~rol room for a Function and control room indication is not restored with'n 7 days. The current required Actions require placing the unit in Cold Shutdown within 2-4 hours while the proposed required Action is to place the unit in Hot Shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. This change is consistent with the proposed change in when the PAM instrumentation is required to be operable.

Having the unit exit the Run and Startup Modes provides the appropriate controls for the subject conditions. These variables are related to the diagnosis and preplanned actions required to mitigate DBAs. The applicable DBAs are assumed to occur in Run or Startup Modes. In the other plant operational modes, plant conditions are such that the likelihood of an event that would require PAM instrumentation is extremely low; therefore, exiting the Run and Startup Modes provides the appropriate action under the evaluated conditions.

G. Required Action for No Control Room Indication of the Containment H2 Analyzer Function This revision changes the LCO from 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, consistent with standard technical specifications 3.3.3.1 and B 3.3.3.1 D.1. The 72-hour Completion Time is based on the low piobability of the occurrence of a LOCA that wculd generate hydrogen in amounts capable of exceeding the flammability limit.

H. Deletion of Conta;nment H2 Analyzer from TS 3/4.7.A.7.c This is an editorial change only. Any non-editorial changes ar- discussed in other sections.

I. Non-Regu;atory Guide 1.'7, Type A or Category 1, non-Type A, Variables The Functions proposed to be deleted are not instrumentation that monito;s the Regulatory Guide 1.97, Type A and Category 1 non-Type A, variables. Therefore, this instrumentation does not meet the screening criteria in the final policy statement and 10 CFR 50.36(c)(2)(ii) and removal from the TS is appropriate.

J. Instrument Check Frequency The current TS requirements have various frequencies identified fo. performing the required Instrument Check (i.e., each shift, once each day, once every 30 days, and once per month).

These required frequencies are made consistent with NUREG-1433.

Performance of the Instrument Check once every 31 days ensures that a gross failure of instrumentation has not occurred. An Instrument Check is normally a comparison of the parameter indicated on one channel against a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations batween instrument channels could be an indication of excessive instrument drift an one of the channels or something even more serious. An Instrument Check will d:tac*. gro:;s channel failure; thus, it is key to verifying the instrumentation continues to operate properly between calibrations. The proposed Frequency of 31 days is based upor, plant operating experience, with regard to channel operability and drift, which demonstrates that failure of more than one channel of a given Function in any 31-day interval is rare. The Instrument Check supplements less 6

formal, but more frequent, checks of channels during normal operational use of those displays associated vith~required channels.

K. Calibration Frequency Presentation The current TS requirements have various presentations of the requiredc c libration frequencies all of which effectively require the calibration a minimum of once per 24 months (i.e., each refueling outage, once each operating cycle, and eaclh start ip.. The presentation of these required frequencies are made consistent with NUREG-;,43? as once per 24 months. This change in presentation of the required interval does not change the maximum interval between calibrations required by the TS for these instruments because the current refueling interval for Pilgrim is 24 months. This change in presentation does remove restrictions on the plant conditions under which the calibration can be performed. Removal of these restrictions is acceptable because implementation of the controls of 10 CFR 50.65 will continue to provide the necessary assurance that the calibrations are not performed during inappropriate plant conditions.

L. Number of Required Neutron Monitoring Channels Increasing the number of required neutron monitoring channels from 1 to 2 is a more restrictive TS requirement. This change does not modify the design or operation of the plant but only results in more conservative TS requirements.

M. Instrument Channels Addressed by the Specification 3/4.2.F requirements With the removal of the "Instrument #" column from Table 3.2.F the TS will no longer control the specific instrument numbers comprising each required Function. But only instrumentation that meets the design requirements for Regulatory Guide 1.97, Type A or Category 1, non-Type A, instruments should be used to meet the minimum required channels requirements of the TS. Therefore, the specific instruments comprising each required Function is proposed to be modified as necessary to reflect the credited Regulatory Guide 1.97, Type A or Category 1 non Type A instruments. The following are the credited Regulatory Guide 1.97, Type A or Category 1, non-Type A, instruments for the proposed TS required Functions.

Function Current TS Regulatory Guide i.97 Instruments Instrumentr,

1. Reactor Water Level 640-29A LI-263-1 OA 640-79  !'1-263-1 00B LI-263-106A LI-263-106B LI-1001-650A LI- 1 001-650B LR-1001-604A LR-1001-604B

2. Reactor Pressure 640-25A PI-253-49A 640-25B PI-263-49B PR-1001-600A PR-1001-600B 7

Function Current TS ;,,. Regulatory Guide 1.97 Instruments Instruments

3. Drywell Temperature TRU-9044 TRU-9044 TI-9019 - TI-9019

4. Neutron Monitoring APRM APRM IRM SRM SRM

5. Suppression Chamber Water TI-5021-2A TI 5021-02A Temperature TI-5022-2B TI 5022-02B TRU-5021-OIA TRU-5022-01 B

6. Suppression Chamber Bottom Pressure PID-5067A PI-1001-69A PID-5067B PI-1001-69B PR-1001-69

7. Suppression Chamber Water Level LR-5038 LI-1001 604A LR-5049 LI-1001 604B LR-1001-604A LR-1001-604B

8. Drywell Pressure (Wide Range) PI-1001-600A PI-1001-600A PR-1001-600A PR-1001-600A PI-1001-600B PI-1001-600B PR-1001-600B PR-1001-600B

9. Drywell Pressure (Narrow Range) PI-1001-601A P1-1001-601A PR-1001-600A PR-1001-600A PI-1001-601B PI-1001-601B PR-1061-600B PR-1001-600B

10. Drywell High Radiation RIT-1001-606,A RIT-1001-606A RIT-1001-606B RIT-1001-606B RR-1001-606A RR-1001-606A RR-1001 606B RR-1001 606B

11. Containment H2 Analyzer AR-1001-612A AR.1001-612A AR-1001-612B AR-1 001 -612B AI-1-5184A Al- ;-5184B N. Delete Note (2) on page 3/4.2-41 With the revised presentation of tho calibration frequency requirements, Note (2) is no longer K*

used on Table 4.2.F. Deletion of this Note is dependant upon the prior approval of the proposed TS change concerning the relocation of certain Control Rod Block functions submitted via Reference 5. If the changes requested in Reference 5 are not approved prior to the approval of this proposed change, then Note (2) should not be deleted as part of this change.

In Reference 4, the NRC approved required actions similar to those proposed by this submittal.

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• t'*

5. REGULATORY SAFETY ANALYSIS "1. No Significant Hazards Analysis ,

Entergy Nuclear Operations, Inc. (Entergy) is proposing to mo.';fy requirments associated with Post-Accident Monitoring Instrumentation.

Pilgrim has evaluated whether or .1ot a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1. Does the change involve a significant increase in the probability or consequences of an accident previously analyzed?

Response: No.

Post-Accident Monitoring (PAM) Instrumentation is not an initiator of any previously evaluated accident because there is no credib!e failure of PAM instrumentation that could initiate previously evaluated accidents. Therefore, the proposed changes do not involve a significant increase in the probability of an accident previously analyzed.

The availability and use of PAM instrumentation help to ensure that the manual operator actions for mitigating an accident will be taken, and that the operator will be able to verify that automatic actions have occurred. The proposed changes make the requirements in the Technical Specifications more consistent with assumed operator actions. The proposed required actions, allowed out-of-service times, and surveillance intervals are appropriate based on operating experience, other instiumentation available, the passive nature of the instrument (no critical automatic action is assumed to occur from these instruments), and the low probabilityof an event requiring PAM instn'mentation. Therefore, the proposed changes do not-w involve a significant increase in the consequences of an accident previously analyzed.

Therefore, this proposed amendment does not involve a significant increase in the probability of occurrence or consequences of an accident previously analyzed.

2. Does the change create the possibility of a new or different kind of accident from any accident previously analyzed?

Response: No The proposed change does not involve the physical modification of structures systems, or components, plant design basis, or the manner in which the plant is operated. PAM instuumentation is passive and does ,ioi initiate automatic actions.

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As a result, there are no credible failures that cbuld initiate a new or different kind of accident from' any accident previously evaluated.';

Therefore, this proposed amendment does not create the possibility of a new or different kind of accident from any previously analyzed.

3. Does the change involve a significant reduction in the margin o salety?

Response: No PAM instrumentation performs no automatic functions. PAM instruments help to ensure that operatos take necessary manual actions to mitigate the consequences of an accident, and that operat3rs have adequate information to confirm the operation of automati- accident mitigation functions have occurred. The proposed required actions, allowed out-of-service times, and surveillance intervals are appropriate based on operating experience, other instrumentation available, the passive nature of the instrument (no critical automatic action is assumed to occur from these instruments), and the low probability of an event requiring PAM instrumentation.

Therefore, this proposed amendment does not involve a significant reduction in margin of safety.

Based on the above, Entergy concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and accordingly, a finding of "no significant hazards consideration" is justified.

6. ENVIRONMENTAL CONSIDERATION A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.

Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need to be prepared in connection with the proposed amendment.

7. COORDINATION WITH OTHER PENDING TS CHANGES The markups of the TS submitted with this change delete Note (2) on TS page 3/4.2-41.

Deletion of this Note is dependant upon the prior approval of the proposed TS change concerning the relocation of certain Control Rod Block functions submitted via Reference 5. If the changes requested in Reference 5 are not approved prior to the approval of this proposed change, then Note (2) should not be deleted as part of this change.

8. REFERENCES

1. NUREG-1433, "Standard Technical Specifications General Electric Plants, BWR/4,"

Revision 2.

10

I -' -

2. NRC letter, S. F. Shankman to G. W. Davis, dated March 13, 1991 regarding "Safety Evaluation of the Pilgrin Nuclear Power Station Response to Conformance to Regulatory Guide 1.97 (TAC NO. 51119).

3. NRC letter, R. B. Eaton to R. A. Anderson, dated December 23, 1992 regarding "Supplemental Safety Evaluation of Conformance to Regulatory Guide (R. G.) 1.97 for Pilgrim Nuclear Power Station (PNPS) (TAC NO. M51119)."

4. NRC letter, R. M. Pulsifer to M. Balduzzi, dated May 10, 2002 regarding 'Vermont Yankee Nuclear Power Station - Issuance of Amendment Re: Allowed Outage Times for Post Accident Monitoring Instrumentation (TAC No. MB3430)."

5. Entergy letter 2.02.064, R.M. Bellamy to NRC, dated August 19, 2002.

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I; ATTACHMENT 1 PROPOSED TECHNICAL SPECIFICATION AND BASES CHANGES (MARK-UP)

LIMITING CONDITION FOR OPERATION f SURVEILLANCE REQUIREMENT 3.2 PROTECTIVE INSTRUMENTATION (Cont) 4.2 PROTECTIVE INSTRUMENTATION (Cont)

E. Drywell Leak Detection E. Drywell Leak Detection The limiting conditions of Instrumentation shall be operation for the instrumentation functionally tested, calibrated that monitors drywell leak and checked as indicated in detection are given in Section*. Section 4.6.C.

3 .6 .C. - -- S u--- o F. Surveillance Information Readouts k_ -- * ' I *Instrumentation shall be/

• '* The

-- limiting conditions calibrated _o and checked as

\ instrumentation that provides indicalte -F, ,/-e surveillance information readouts regie in Table 3.2.F.

/

Revision+/--7 Amendment No. 89;-15-93 3/4.2-3

INSERT 3/4.2-3A F. 'Post-Accident Monitoring instrumentation F. Post-Accidenit Monitoring Instrumentation During the RUN MODE and the STARTUP Instrumentation shcll be calib.ated and MODE the limiting conditions for the checked as indicated in T&ble 4.2.F.

instrumentation that provides post-accident monitoring are given in Table 3.2.F.

/

PNPS TABLE 3.2.F SURVEILLANCE INSTRUMENTATION

1. of Notes Type Indication and Range Operable Instrvment Instrument # Parameter Channels 2 -29A & B Reactor Water Level Indicator 0-60" (1) (2) (3)

I Indicator 0-1200 psig (1) (2) (3) 2 640-25A & Reactor Pressure 2 TRU-9044 I ell Pressure Recorder 0-80 psia (1) (2) (3)

TRU-9045 Recorder, Indicator 0-400°F (1) (2) (3) 2 TRU-9044 Drywell Temper tur TI-9019 Recorder, Indicator 0-400°F (1) (2) (3)

- 2 TRU-9045 Suppression Chaml ier Air TI-9018 Temperature Recorder - 0+7 nches (1) (2) (3) 2 LR-5038 Suppression Chamt )er Water LR-5049 Level (2) (3)(4)

I SRM. IRM. LPRM 1 NA Neutron Monitoring S.

. ... ! J 0 to 100% power Revision 9 3/4.2-25 Amendment No. 31, 48,983, t86

INSERT 3/4.2-25A PNPS TABLE 3.2.F Post-Accident Monitoring Instrumentation Minimum # of Operable Function Instrument Channels Notes

1. Reactor Water Level 2 (1) (2)

2. Reactor Pressure 2 (1) (2)

3. Drywell Temperature 2 (1) (2)

4. Neutron Monitoring 2 (1) (2)

5. Suppression Chamber Water Temperature 2 (1) (2)

6. Suppression Chamber Bottom Pressure 2 (1) (2) 7 Suppression Chamber Water Level 2 (1) (2)

8. Drywell Pressure (Wide Range) 2 (1) (2)

9. Drywell Pressure (Narrow Range) 2 (1) (2)

10. Drywell High Radiation 2 (1) (3)

11. Containment H2 Analyzer 2 (1) (4)

INSERM 3/4.2-25A Continued Notes for Table 3.2.F (1) With less than the minimum number of instrument channels, restorg the anooerable channel(s) within 30 days. F the inoperable channel(s) are rict restoled, prepare and submit a special report to the Commission within 14 days of the event outlining '.c' action taken, the cause of the inoperability and the plans and schedule for restoring the char.. els to operable status.

(2) With the instrument channel(s) providing no indication to the control room, restore the indication to the control room within 7 days. If the indication is not restored, an orderly shutdown shall be initiated and the reactor shall be in the Hot Shutdown Condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

(3) With the instrument channel(s) providing no indication to the control room, restore the indication to the control room within 7 days. If the indication is not restored, prepare and submit a special report to the Commission within 14 days of the event outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the channels to operable status.

(4) With the instrument channel providing no indicat;on to the control room, restore the indication to the control room within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. If the indication is not restored, an orderly shutdown shall be initiated and the reactor shall be in the Hot Shutdown Condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

PNPS "ITDLE3.2.F (Cont) .TnsexT 312. 9 SURVEILLANCE INSTRUMENTATION Minimum # of Operable Instrument Parameter Type Indication Notes annels Instrument # and Range TI-5021-2A Suppression Chamber Indicator/

TRU-5021-1A Water Temperature Multipoint Recorder (1) (2) (3) (4) 30-230°F (Bulk)

T1- 12-23BSuppression Chamber Indicator/

rTR U-0 Water Temperature Multipoint Recorder (1) (2) (3) (4) 3Q0-230°F (Bulk) 1PID-5021 Drywell/Torus Diff.

,,Pressure Indicator -.25 - +3.0 psig (1) (2) (3) (4)

I PID-5067A D Pressure Indicator -.25 - +3.0 psig (1) (2) (3) (4)

PID-5067B Torus Pr sure Indicator - 1.0 - +2.0 psig 1Nalve (a)Primary Safety/Relief Va (a) Acoustic monitor (5) or Position (b) Thermocouple (b) Backup INalve (a)Primary Safety Valve Position (a) ustic monitor (5) or Indicator (b) The couple (b) Backup INalve See Note (6) Tail Pipe Temperature Thermocouple Indication LI-1001-604A Torus Water Level Indicator/Multipoint (1) 3) (4) 2 LR-1001-604A (Wide Range) Recorder 0-300" H2 0 LI-1001-604B LR-1001-604B Torus Water Level Indicator/Multipoint (1) (2) (3) (4)

(Wide Range) Recorder 0-300" H2 0 Revision 196 Amendment No. 56,83, 172 3/4.2-26

INSERT 3/4.2-25A Continued Pages 3/4.2-27 and 3/4.2-28 have been deleted.

PNPS TABLE 3.2.F (Cont)

SURVEILLANCE INSTRUMENTATION Minim # of Operable In rument Type Indication Channel Instrument # Parameter and Range Notes (PI 1001-600A Containment Pressure, Indicator/Multipoint (4) (1) (2) (3)"

PR 1001-600A (High Range) Recorder 0-225 psig 2

(PT 1( -600B Containment Pressure, Indicator/Multipoint (4) (1) (2) (3)

(PR 1001- OB (High Range) Recorder 0-225 psig (PI 1001-601A ontainment Pressure, Indicator/Multipoint (4) (1) (2) (3)

(PR 1001-600A (L Range) Recorder -5 to 5 psig 2(

(PIl 001-601B Containmen ressure, Indicator/Multipoint (4) (1) (2) (3)

(PR 1001-600B (Low Range) Recorder -5 to 5 psig (RIT 1001-606A Containment High Radia n Monitor/Multipoint 2 (RIT 1001-606B (Drywell) Recorder (4) (7)

(RR 1001-606A to 1x10 7 R/hr (RR 1001-606B "1 RI 1001-609 Reactor Building Vent Indicator/ ipoint (4) (7)

RR 1001-608 Recorder 10-1 to 104 R/hr 1 RI 1001-608 Main Stack Vent Indicator/Multipoint (7)

RR 1001-608 Recorder 10-1 to 104 R/hr 1 RI 1001-610 Turbine Building Vent Indicator/Multipoint (4) (7)

RR 1001-608 Recorder 10-1 to 104 R/hr

, Revision 1q+

Amendment No. 56;-g837;4A? 3/'

4.2-27

NOTES FOR TABLE 3.2.F (1) With less than'the minimum number of instrument channels, restore the inoperable channel(s) within 30 days.

) With the instrument channel(s) providing no indication to the control room, restore the indication to the control-room within seven days.

(3) f the requirements of notes (I)-*or (2) cannot be met, an orderly shutdown s 11 be initiated and the reactor shall be in the'Cold Shutdown Condition with'n-24 hours. I (4) These urveillance instruments are considered to be reddffdaint to each other.

(5) At a mini*m, the primary or back-up* parameter indicators shall be operable for each va'. e when the valves are required to be operable. With both primar and backup* in trument channels inoperable either return one (1) channel to operable status ithin 31 days or be in a shutdown mode within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

The following instru ents are associated with the safety/relief and safety valves:

Prima r* Secondary Valve Acoustic Monitor Tail Pipe Temperature Thermocouple 203-3A ZT-203-3A TE6271 TE6272

• ZT-203-3B

• 203-3B 203-3C ZT-203-3C TE6273

• 203-3D ZT-203-3D TE6276

• 203-4A ZT-203-4A TE6274-B 203-4B ZT-203-4B *675-B

• See Note (6)

(6) At a minimum, for thermocouples providing SRV tail pipe tem erature, one of the dual thermocouples will be operable for each SRV when the valves are required to be operable. If a thermocouple becomes inoperable, it shall be returned to an operable condition within 31 days or the reactor all be placed in a shutdown mode within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

(7) With less than the minimum number of operable instrument channels, re tore the inoperable channels to operable status within 7 days or prepare and su it a special report to the Commission within 14 days of the event outlining t e action taken, the cause of the inoperability and the plans and schedule f restoring the channels to operable status.

Revision 177 Amendment No. 48;-83;-103 3/4 2-28

-I - --

S~ TABLE 4.2.F

, INIMýUM TEST AýND CALIBRAýTION ýFREIQUENCY FOR SURVEILLANCE INSTRUMENTATION Instrument C nnel Calibration Frequency Instrument Check

1) Reactor Water Level Each Refueling Outage Each Shift

2) Reactor Pressure Each Refueling Outage Each Shift'

3) Drywell Pressure Each Refueling Outage Each Shift

4) Drywell Temperature Once/6 Months Each Shift

5) Suppression Chamber Temperature Once!6Months Each Shift

6) Suppression Chamber Water Level Once/6 Mon Each Shift

7) NA

8) Neutron Monitoring (2) Each Shift

9) Drywell/Torus Differential Pressure Once/6 Months Each Shift

10) Drywell Pressure Once/6 Months Each Shift Torus Pressure Once/6 Months

11) Safety/Relief Valve Position Indicator Each refueling outage Onc ach day (Prim ary/Secondary)

12) Safety Valve Position Indicator Each refueling outage Once each da (Primary/ Secondary) _ _, - 37 Revision 22 Amendment No. +86- 3/4.2-37 I3-,48-97,

INSERI 3/4.2-37A PNPS TABLE 4.2.F MINIMUM TEST AND CALIBRATION FREQUENCY FOR Post-Accident Monitoring Instrumentation Calibration Instrui.vint Function Frequency Cheok

1. Reactor Water Level Once/24 Months Once/31 Days

2. Reactor Pressure Once/24 Months Once/31 Days

3. Drywell Temperature Once/6 Months Once/31 Days

4. Neutron Monitoring Once/24 Monihs Or.c6/31 Dayv, 5 Suppression Chamber Water Temperature Once/I Months Once/31 Days

6. Suppression Chamber Bottom Pressure Once/6 Months Once/31 Days

7. Suppression Chamber Water Level Once/24 Months Once/31 Days

8. Drywell Pressure (Wide Range) Once/24 Months Once/31 Days

9. Drywell Pressure (Narrow Range) Once/24 Months Once/31 Days

10. Drywell High Radiation Once/24 Months Once/31 Days

11. Containment H2 Analyzer Once/6 Months Once/31 Days

INSERI' 3/4.2-37A Continued Page 3/4.2-37 has been deleted.

PNpS TABLE 4.2.F (Cont)

MINIMUM TEST AND CALIBRATION FREQUENCY FOR SURVEILLANCE INSTRUMENTATION annel Calibration Frequency Instrument Check

13) Torus Water Level (Wide e Each refueling outage Once every 30 days

14) Containment Pressure refueling outage Once every 30 days

15) Containment High Radiation Once/Op Cycle Once every 30 days

16) Reactor Building Vent Radiation Monitor Once/Operating Cycle Once every 30 days

17) Main Stack Vent Radiation Monitor Once/Operating Cycle Onc 30 days

18) Turbine Building Vent Radiatibn Monitor Once/Operating Cycle Once every 30 days RevisionQ46y Amendment No. 42,83, 4689- 3/4.2-38

NOTES FOR TABLES 4.2.AN.THROUGH 4.2.G

1. Initially once per month until exposure hours (M as defined on Figure 4.1.1) is 2.0 x 105; thereafter, according to Figure 4.1.1 with an interval not less than one month nor more than three months.

SFunctional tests, calibrations and instrument checks are not required when these are not required to be operable or are tripped. Functional tests shall be performe -dbfor-e-e h startup with a required frequency not to exceed once per week. Calibrations o d SRMs shall be performed during each startup or during controlled shtdowns a required frequency not to exceed once per week. Instrument checks shall be per d at least once per day during those periods when the instruments are required t operable.

3. This instrumentation is excepted from the functional test definition. The functional test will consist of injecting a simulated electrical signal into the measurement channel.

4. Simulated automatic actuation shall be performed once each operating cycle.

Where possible, all logic system functional tests will be performed using the test jacks.

5. Reactor low water level and high drywell pressure are not included on Table 4.2.A since they are tested on Tables 4.1.1 and 4.1.2.

6. The logic system functional tests shall include a calibration of time delay 2 relays and timers necessary for proper functioning of the trip systems.

7. Calibration of analog trip units will be performed concurrent with functional testing. The functional test will consist of injecting a simulated electrical signal into the measurement channel. Calibration of associated analog transmitters will be performed each refueling outage.

Revision - Amendment No. 42;-99;-I1Q-147;-1-54 3/4.2-41

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.7 CONTAINMENT SYSTEMS (Cont) 4.7 CONTAINMENT SYSTEMS (Cont)

A. Primary Containment (Cont) A. Primary Containment (Cont)

7. Containment Atmosphere Dilution 7. Containment Atmosphere Dilution

a. Within the 24-hour period a. The post-LOCA containment after placing the reactor in atmosphere dilution system the Run Mode the Post-LOCA shall be functionally tested Containment Atmosphere once per operating cycle.

Dilution System must be operable and capable of b. The level in the liquid supplying nitrogen to the N2 storage tank shall be containment for atmosphere recorded weekly.

dilution. If this specification cannot be met, the system must be restored to an operable condition within 30 days or the reactor must be at least in Hot Shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. d. Once per month each manual or power operated valve in

b. Within the 24-hour period the CAD system flow path not after placing the reactor in locked, sealed or otherwise the Run Mode, the Nitrogen secured in position shall be Storage Tank shall contain a observed and recorded to be minimum of 1500 gallons of in its correct position.

liquid N2 . If this specification cannot be met the minimum volume will be restored within 30 days or the reactor must be in at least Hot Shutdown within 12 S~hours.

~~~c. There are 2 H2-----ana yzers *

~available to serve the

~drywell.

'th only 1 H2 analyzer op able, reactor operation is al ed for up to 7 days.

If the in erable analyzer is not made ly operable within 7 days, t reactor shall be in a least Shutdown within the next hours.

Revision -W--;

Amendment No.

3/4.7-10

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS

1) 3.7 CONTAINMENT SYSTEMS (Cont.) 4.7 CONTAINMENT SYSTEMS (Cont.)

\*'*/itb*no H2 analyzer operable, reactor operatibr.s allowed for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. If fully operable withn b the reactor S\shall be in at least Hot Shutdo6VWrrmth4Lh I next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />._.)--- ".

B. Standby Gas Treatment System and B. Standby Gas Treatment System and Control Room Hiqh Efficiency Air Control Room High Efficiency Air Filtration Filtration System System

1. Standby Gas Treatment System 1. Standby Gas Treatment System

a. Except as specified in 3.7.B.1 .c or a. 1. At least once per operating cycle, 3o7.B.1.e below, both trains of the it shall be demonstrated that standby gas treatment shall be pressure drop across the operable when in the Run, Startup, combined high efficiency filters and and Hot Shutdown MODES, during charcoal adsorber banks is less movement of irradiated fuel than 8 inches of water at 4000 cfm.

assemblies in the secondary 2. At least once per operating cycle, containment, and during movement demonstrate that the inlet heaters of new fuel over the spent fuel pool, on each train are operable and are and during CORE ALTERATIONS, capable of an output of at least 14

) and during operations with a kW.

potential for draining the reactor vessel (OPDRVs), 3. The tests and analysis of Specification 3.7.B.1 .b. shall be performed at least once per or operating cycle or following painting, fire or chemical release in the reactor shall be in cold any ventilation zone shutdown within the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. communicating with the system while the system is operating that

b. 1. The results of the in-place cold could contaminate the HEPA filters DOP tests on HEPA filters shall or charcoal adsorbers.

show >99% DOP removal. The results of halogenated 4. At least once per operating cycle, hydrocarbon tests on charcoal automatic initiation of adsorber banks shall show

>99.9% halogenated hydrocarbon removal.

Revision P1,3 Amendment No. 15, 42,50,51, 52,4 12, 114,151, 46, 170,487- 3/4.7-11

BASES:

3.2 PROTECTIVE INSTRUMENTATION (Cont)

Trip settings of < 100 mr/hr for the monitors in the refueling area ventilation exhaust ducts are based upon initiating normal ventilation isolation and standby gas treatment system operation so that none of the activity released during the refueling accident leaves the Reactor Building via the normal ventilation path but rather all the activity rocessed by the standby gas treatment system. ( - )

For~qost parameters monitored, as listed in Table 3.2;F, there-are two (2) channels of instrumentation. By comparing readings between these two (2) channels, a near continuous surveillance of instrument performance is available. Meaningful deviation in comparative readings of these instruments will initiate an early recalibration, thereby maintaining the quality of the instrument readings.

e Safety - Safety/Relief Valve position indication instrumentation provides t he operator w aideon-on-se/p npex~s tn mnnoizr*-ud.

ssess these variables during and following an accident" In response to NUREG-0737, modifications were made to the ADS logic to extend automatic ADS operation to a class of transients that involve slowly uncovering the core without depressurizing the vessel or pressurizing the drywell. These transients were analyzed assuming no high pressure injection systems (feedwater, HPCI or RCIC) are available. Only ADS is considered available to depressurize the vessel, permitting operation of LPCI. The transients generally involve pipe breaks outside containment. Automatic ADS would not occur on low water level because high drywell pressure would not be present and ADS logic has a high drywell pressure permissive. The modification added a timer to the ADS logic which bypasses the high drywell pressure permissive, and a manual inhibit switch which allows the operator to inhibit automatic ADS initiation for events where automatic initiation is not desirable.

An analysis'was performed to determine an upper time limit on the bypass timer. The goal was to ensure ADS is automatically initiated in time to prevent peak clad temperature (PCT) from exceeding 1500°F for a limiting break, which was determined to be a Reactor Water Cleanup line break. The analysis concluded that there are 18 minutes between the low water level initiation of the timer and the heatup of the cladding to the limit. Since the logic includes a 2 minute delay already, the bypass timer upper limit can not be more than 16 minutes, which provides a conservative margin for PCT and allows sufficient time for operator intervention if required. A minimum time delay is incorporated to allow RPV water level to recover, resetting the timer and preventing depressurization. The choice of a timer setting of 11 minutes places the setting in the middle and provides maximum tolerance from either limit. (

Reference:

GE Report "Bypass Timer Calculation for the ADS/ECCS Modification for Pilgrim Station" December 16, 1986).

Revision -+/--7 Amendment No. 89;-133;---9* B3/4.2-5

INSERT B3/4.2-5A The. following are the instruments which meet thequality requirements of Regulatory Guide 1.97 and can be credited to meet the operabiiity requirements of Specificatioi6'3/4.2.F. Two channels of indication in the control room from the below list must be available to meet the Specification requirements.

Regulatory Guide 1.97 Instrurrncr.t.,

Specification 314.2.F Function

1. Reactor Water Level LI 263-1O0A, LI 263-106A, LR 1001-604A LI 263-100B, LI 263-106B, LR 10'i-60-B LI 1001-650A, LI 1001-65CB

2. Reactor Pressure PI 263-49A, PR-1 001-600A PI 263-49B, PR-1001-600B

3. Drywell Temperature TRU-9044, TI-9019

4. Neutron Monitoring APRM, SRM

5. Suppression Chamber Water Temperature TI 5021-02A, TRU-5021-01A TI 5022-02B, TRU-5022-01 B

6. Suppression Chamber Bottom Pressure PI 1001-69A, PI 1001-69B PR 1001-69

7. Suppression Chamber Water Level LI 1001 604A, LR 1001-604A LI 1001 604B, LR 1001-604B

8. Drywell Pressure (Wide Range) PI-1001-600A, PR-1001-600A PI-1001-600B, PR-1 001 -600B

9. Drywell Pressure (Narrow Range) PI-1001-601A, PR-1001-600A PI-1001-601 B, PR- 001 -600B

10. Drywell High Radiation RIT-1001-606A, RR 1001-606A RIT-1001-606B, RR 1001-606B

11. Containment H2 Analyzer AI-1-5184A, AR 1001-612A AI-1-5184B, AR 1001-612B

BASES:

3/4.7 CONTAINMENT SYSTEMS (Cont)

A. Primary Containment (Cont) wetwell pressure differential to keep the suppression chamber downcomer legs clear of water significantly reduced suppression chamber post LOCA hydrodynamic loads. A pressure of 1.17 psid is required to sufficiently clear the water legs of the downcomers without bubbling nitrogen into the suppression chamber at the 3.00 ft. downcomer submergence which corresponds to approximately 84,000 ft. 3 of water. Maximum downcomer submergence is 3.25 ft.

at operating suppression chamber water level. The above pressure differential and submergence number are used in the Pilgrim I Plant Unique Analysis.

Post LOCA Atmosphere Dilution In order to ensure that the containment atmosphere remains inerted, i.e. the oxygen-hydrogen mixture below the flammable limit, the capability to inject nitrogen into the containment after a LOCA is provided. A minimum of 1500 gallons of liquid N 2 in the storage tank assures that a three-day supply of N2 for post-LOCA containment inerting is available. Since the inerting makeup system is continually functioning, no periodic testing of the system is required.

The Post-LOCA Containment Atmospheric Dilution (CAD)

System is designed to meet the requirements of AEC Regulatory Guides 1.3, 1.7 and 1.29, ASME Section III, Class 2 (except for code stamping) and seismic Class I as defined in the PNPS FSAR.

In summary, the limiting criteria are:

1. Maintain hydrogen concentration in the containment during post-LOCA conditions to less than 4%.

2. Limit the buildup in the containment pressure due to nitrogen addition to less than 28 psig.

3. To limit the offsite dose due to containment venting (for pressure control) to less than 300 Rem to the thyroid.

By maintaining at least a 3-day supply of N 2 on site there will be sufficient time after the occurrence of a LOCA for obtaining additional nitrogen supply from local commercial sources.(') The system design contains sufficient redundancy to ensure its reliability. Thus, it is sufficient o.e.t.of to test the the whole system once per operating cycle. T analyzer, ide reunancy for the drywe i.., ner are two H2 analyzers for the Unit. By pe ctor operation for 7 days with one of the two H2 ion_11be analyzers inoperable, aintained*

reddundancy of ana yzing c ill be maintained while not imposin an immediate interruption in lant operation.

(1) As listed in Pilgrim Nuclear Power Station Procedure No. 5.4.6 "Post Accident Venting".

Revision -77 Amendment No. §§;-4 VI A -

L.J/ +, I -0

BASES:

3/4.7 CONTAINMENT SYSTEMS (Cont) testing of the analyzer~s using H2 will be adequate-'ensure to" the system's readiness ecause deT'tti., of the design.

to us.I Since rL~to-~e~t~erpine the analyzers are normally not in operation there will be little H2 concentration, thee analyzers must be warmed up 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> prior to putting into service. fT'si ý " týable for accident conditions be causea 1%Hý level will not.be-re, __hed in th rwel until 16 hour1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> w*.

he accident. ,Due to nitrogen addition, the pressure in the containment a ter a LOCA will increase wit time. Under the worst expected conditions the containment pressure will reach 28 psig in approximately 45 days. If and when that pressure is reached, venting from the containment shall be manually initiated per the requirements of 10CFR50.44.

The venting path will be through the Standby Gas Treatment system in order to minimize the off site dose.

B.1 Standby Gas Treatment System The Standby Gas Treatment System is designed to filter and exhaust the reactor building atmosphere to the stack during secondary containment isolation conditions. Upon containment isolation, both standby gas treatment fans are designed to start to bring the reactor building pressure negative so that all leakage should be in leakage. After a preset time delay, the standby fan automatically shuts down so the reactor building pressure is maintained approximately 1/4 inch of water negative. Should one system fail to start, the redundant system is designed to start automatically. Each of the two trains has 100%

capacity.

High Efficiency Particulate Air (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 0.1% bypass leakage for the charcoal adsorbers and a HEPA filter efficiency of at least 99% removal of cold DOP particulates.

The laboratory carbon sample test results should indicate a methyl iodide removal efficiency of at least 97.5% for expected accident conditions. The specified efficiencies for the charcoal and particulate filters is sufficient to preclude exceeding 10CFR100 guidelines for the accidents analyzed. The analysis of the loss of coolant accident assumed a charcoal adsorber efficiency of 95% and TID 14844 fission product source terms, hence, installing two banks of adsorbers and filters in each train provides adequate margin. A 14 kW heater maintains relative humidity below 70% in order to ensure the efficient removal of methyl iodide on the impregnated charcoal adsorbers. Considering the relative simplicity of the heating circuit, the test frequency of once/operating cycle is adequate to demonstrate operability.

Air flow through the filters and charcoal adsorbers for 15 minutes each month assures operability of the system. Since the system heaters are automatically controlled, the air flowing through the filters and adsorbers will be <70% relative humidity and will have the desired drying effect.

Revision t~T Amendment No.42,4- .4 , 4,4.6 B3/4.7-9

ATTACHMENT 2 LIST OF REGULATORY COMMITMENTS

List of Regulatory Commitments The following table identified those action,, committed to by Pilgrim in this document. Any other statements in this submittal are provided for information purposes and are not considered to be regulatory commitments.

REGULATORY COMMITMENT DUE DATE Implement changes to the Technical 60 days following approval of the Specifications within 60 days of approval, amendment.