ML20028G874

Forwards Supplemental Info to Support 900523 Application for Amends to Licenses DPR-39 & DPR-48 Re Generic Ltr 83-37

ML20028G874
Person / Time
Site:	Zion  File:ZionSolutions icon.png
Issue date:	08/26/1990
From:	Chrzanowski R COMMONWEALTH EDISON CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
GL-83-37, NUDOCS 9009050280
Download: ML20028G874 (65)

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Commonwealth Edison l

__ J Do:;ners Crove, Illinois 60515 1400 Opus Place

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%d August 26, 1990 U.S. Nuclear Regulatory Commission Document Control Desk l

Washington, DC 20555 l

Subject:

Zion Station Unita 1 and 2 Supplemental Information to Support an Application for Amendment to Facility Operating Licenses DRP-39 and DPR-48 HRC_Dacle1 30s. 50-295 and 50-304

Reference:

May 23, 1990, letter from R.A. Chrzanowski to T.E. Murley i

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Dear Dr. Nurley:

Reference (a) contained a proposed Technical Specification Amendment regarding the incorporation of the Post Accident Monitoring Instrumentation requirements from Generic Letter 83-37 into the Zion Station Technical Specifications. After numerous discussions with the NRC Project Manager, the attachment to this letter is being submitted to clarify the intent of the Core Exit Thermocouples, to clarify the PORV position indicators, and to increase the minimum number of channels and the-required number of channels for the Auxiliary Feedwater Flowrate and the Steam Generator Water Level (Wide Range) indications.

The Attachments to this letter contain supplemental information presented with side bars noted in the margin of the affected pages.. The entire submittal is contained in the attachments for completeness. No significant changes have been made with this submittal.

The-conclusions from the Significant llazards Evaluation in reference (a) remain unaffected.

l The supplemental information has been reviewed and approved by both onsite and offsite review in accordance with 10 CFR 50.92 (c).

Please direct any questions regarding this matter to this office.

I Very truly yours, h

GK R.A. Chrzanowski Nuclear Licensing Administrator cc:

C. Patel - NRR Zion Resident Inspector Region III Office Office of Nuclear Facility Safety - IDMS 0312) 8O3 9009050280 900826 PDR ADOCK 05000295 P

PDC

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ful LIST OF.CHARQES The May 23, 1990, Commonwealth Edison submitted proposed Technical l

Specifications for the Regulatory Guide 1.97 Post Accident Monitoring Instrumentation.

Based on discussions with the NRC Project Manager, j

additional information was requested. Tbs purpose of.this attachment is to j

summarize the changes that have been made from the original submittal to this l

submittal. The page numbers noted below correspond to those in this submittal.

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last paragraph was added to provide additional information pertaining 3

to the control room ventilation system.

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added reference to NRC letter dated October 31, 1980.

deleted reference to NRC memorandum dated October 28, 1985.

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deleted reference to NRC memorandum dated October 28, 1985.

11 revised wording in last paragraph for clarity.

l 12 revised last paragraph Ath line. Original submittal referenced Steam Generator level wide range instrumentation minimum number of channels operable was being eb:n ud from 1 to not applicable. This is s

revised in t'.is submittal changing the minimum number of chat.nels operable from 1 to 2.

i also revised this paragraph to delete the words, "and 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> in the event all 4 instruments failing".

13 the first paragraph has been reworded to support changing the minimum number of channels f rom 1 to 2.

i the second paragraph, third line, initially stated that the Auxilary j

Feed water flow instruments minimum number of channels was being changed from 1 to not applicable. After further discussions, this is changed from 1 to 2 minimum number of channels.

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~I in the second paragraph, 18th line, the words, "which is currently in Zion Technical Specifications" was added for clarity.

in the second paragraph, last line, the words, "and 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> in the event of all 4 instruments failing" was deleted.

14 the first paragraph was added for clarity.

In the second paragraph, 7th line, the words " Zion Technical" were added for clarity.

15 paragraphs 3, 4, and 5 were added to provide additional information on the PORVs.

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Changes (continued) 16 paragraphs 1 and the first part of 2 were added/ revised to discuss the FORVs.

the second paragraph deleted reference to NRC memorandum dated October 28, 1985.

'$ 2 was reworded for clarity.

17 paragr*

in paragraph 5, line 4, the word "was" was changed to "were".

paragraph 3, line 9, was added for clarity.

19 paragraph 1, line 6, deleted the redundant word " performing".

paragraph 2, line 23, date changed to October 31, 1990.

the last paragraph was deleted, that reference NRC memorandum dated October 28, 1985.

20 paragraph 1, ifne 23, deleted NRC memorandum dated October 28, 1985, and added reference to NRC Ictter dated April 14, 1989.

paragraph 2 was revised to provide additional information on Core Exit Thermocouples.

21 paragraph I revised wording for clarity on Core Esit Thermocouples.

22 paragraph 2, line 10 added a comma.

47 paragraph I deleted reference to NRC memorandum dated October 28, 1985:

51 paragraph 2 added word " consequences" to line 8.

deleted initial submittal sentence of "The consequences of evaluated accidents will remain the same".

It made redundant statement to previous amended sentence.

reworded sentence in line 11 to reflect changing the minimum number of channels f rom "I to not applicable" to "1 '.o 2 channels".

52 paragraph 3, line 11, the words "2 or more channels" were added tur clarity.

53 in paragraph 1, line 30 a comma was added.

56 variable 7 Steam Line Pressure deleted PT 514, 524, 534, and 544 instruments.

They are not Type A variables.

58 the last part of the lar: paragraph was added for clarity.

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The following Attachments are included in this submittalt Status of Generic Letter 83-37 NUREG 0737

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Technical Specification Enclosure 1-Items.

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Purpose of the Accident Monitoring System.

History Summary of Proposed Changes Technical Specification Pages

-i Evaluation of Significant Hasard. Consideration f

t Environme tal Assessment Statement

-Attachment 8 Zion Station Type A Variable i

Post Accident Monitoring Instrumentation r 0 Unchanged Variables i

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t ATTACHMENT 1 STATUS OF GENERIC LETTER 83-37 NUREG 0737 TECHNICAL SPECIFICATION ENCLOSURE 1 ITEMS i

The following information provides a status of all Enclosure 1 Iter to Generic Letter 83-37.

(1) Reactor Coolant Svttem Vents (IIB.1)

At least one reactor coolant system vent path (consisting of at least two valves in series which are powered from emergency busses) shall be operable and closed at all times (except for cold shutdown and refueling) at each of the following locations *

a. Reactor Vessel Head

b. Pressurizer stetm space t

c. Reactor coolant system high point STATUS:

This item has been complied with as addressed in Technical Specification 3/4.3.1.G (2) Post-accident Samolina (II.B.3)

Licensees should ensure that their plant has the capability to obtain i

and analyze reactor coolant and containment atmosphere samples under accident condttions. An administrative program should be established, implemented and maintained to en;"re this-capability.

The program should include:

l a) training of personnel l

b) procedures for sampilng and analysis, and i

c) provisions for maintenance of sampilng and analysis equipment.

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l It is acceptable to the Staff, if the licensee elects to reference this program in the administrative controls section of the Technical Specifications and include a detailed description of the program in o

the plant operation manuals.

A copy of the program should be easily available to the operating staff during accident and transient conditions.

STATUS:

The requirement to have Post-Accident Sampling Procedures is j

currently in Zion Station's Technical Specifications as Specification 6.2.1.L.

This Specification is being revised in this submittal to l

address the program to include; the requirements of training l

personnel, procedures for sampling and analysis, and maintenance of sampling and analysis equipment.

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i (3) Lona Term Auxiliary Feedwater System Evaluation (li.E.1.1) l The objective of this item is to improve the reliability and 1

performance of the Auxiliary Feedwater (AFN) system.

Technical l

Specifications depend on the results of the licensee's evaluation and staff review of each plant.

The Limiting Conditions of Operation (LCO) and surveillance requirements for the AFW system should be similar to safety related systems.

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STATUS:

This item has been complied with as addressed in Technical Specification 3/4.7.2.

(4)

Noble Gas _ Ef fluent f4oni tors (II.T.I.1)

Noble gas effluent monttors provide 11 formation, during and following an accident, which are considered helpful to the operator in accessing the plant condition.

It is duired that these monitors be operable at all times during plant operation, but they are not required for safe shutdown of the plant.

In case of f>.11ure of the monitor, appropriate actions should be taken to restor i its operational capability in a reasonable period of time.

Considering the importance of the availability of the equipment and possible delays involved in administrative controls, 7 days is considered to be the appropriate time period to restore the operability of the monitor.

An alternate inethod for monitoring the effluent should be initiated as soon as practical, but no later than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after the identification of the failure of the monitor.

If the monitor is not L

restored to operable conditions within 7 days after the failure a special report should be submitted to the NRC within 14 days following the event, outlining the cause of inoperability, actions taken and the planned schedule for restoring the system to operable status.

STATUS 1 These monitors are currently contained in Zion Station Technical Specification Table 3.12-1.

A new action statement is being proposed to incorporate the requirements of Generic Letter 83-37 in this submittal.

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(5) S unlina and Analvtis of Plant Effluen.h (II.F.1.2) i Each operating nuclear power reactor should have the capability to collect and analyze or measure representative samples of radioactive lodines and particulates in plant gaseous effluents during and following an accident.

An administrative program should be established,-implemented and maintained to ensure this capability.

The program should include:

a) training of personnel b) procedures for sampling and analysts, and c) provisions for maintenance of sampling and an61ysis equipment It is acceptable to the staff, if the licensee eltets to reference this program in the administrative controls section of the Technical Specifications and include a detailed descriptiors of the program in the plant operation manuals.

A copy of the program should be rnadily available to the operating staff during accident and transient conditions.

STATUS:

This item is being proposed for incorporation into Technical Specification 6.2.1.L. as addressed in Generic Letter 83-37 in this submittal.

(6) Containment Hich Ranae Radiation Monitor (II.F.1.3)

A minimum of two in containment radiation IcVel monitors w th a

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maximum range of 108 rad /hr (107 R/hr for photon only) should be operable at all times except for cold shutdown and refueling outages.

In case of failure of the monitor, appropriate actions should be taken to restore its (nerational capability as soon as possible.

If the monitor is not restored to operable condition i

within 7 days after the failure, a special report should be submitted to the NRC within 14 days following the event, outlining the cause of inoperability, actions taken and the planned schedule for restoring the equipment to operrble status.

I STATUS:

These monitors ar1 currently contained in Zion Station Technical Specification Tab e 3.14-1.

A new action statement is being proposed l

to incorporate the requirements of Generic Letter 83-37 in this submittal.

19140 (a)

(7) Containment Pressure Monitor (II.F.1.4)

Containment pressure should be continuously indicated in the control room of each operating reactor during Power Operation Start-up and Hot Standby modes of operation.

Two channels should be operable at all times when the reactor is operating in any of the above mentioned modes.

Technical Specifications for these monitors should be included with other accident monitoring instrumentation in the present Technical Specifications.

Limiting conditions for operation (including the required Actions) for the containment pressure monitor should be similar to other accident monitoring instrumentation included in the present Technical Specifications.

STATUS:

This item is being proposed for incorporation into the Post Accident Monitoring Instrumentation Table 3.8.9-1.

Reference'to the Narrow Range Containment Pressure instrumentation will be changed to Hide Range to be consistent with the instrumentation approved in Zion Station's degulatory Guide 1.97 submittal.

This proposed change is t,elng made in accordance with the pidance specified in Generic Letter 83-37.

(8) Containment Water level Monitor (II.F.1.5),

A continuous indication of containment water level should be provided in the control room of each reactor during Pow 9r Operation. Start-up and Hot Standby modes of operation. At least one channel for narrow range and two channels for wide range instruments should be operable at all times when the reactor is operating in any of the above modes.

Narrow range instruments should cover the range from the i

bottom to the top of tne containment sump.

Nide range instruments should cover the range from the bottom of the containment to the elevation equivalent to a 600,000 gallon (or lets if justified) capacity.

Technical Specifications for containment water level monitors should be included with other accident monitoring instrumentation in the present Technical Specifications.

LCOs (including the required Actions) for wide range monitors should be similar to other accident monitoring instrumentation included in the L

present Technical Specifications.

LCOs for narrow range monitor i~

should include the requirement that the inoperable channel stil be l

restored to operable status within 30 days or the plant will be brought to Hot Shutdown condition as required for other accident monitoring 1.istrumentation.

STATUS:

This item is being proposed for incorporation into Accident L

Monitoring Instrumentation Table 3.8.9-1 in this submittal.

This change is in accordance with the guidance specified in Generic Letter 83-37.

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1 (9) Containment Hydronen Monitor (!!.F.I 6)

Two independent containment hydrogen monitors should be operable at i

all times when the reactor is operating in Power Operation or Start-up modes.

LC0 for these monitors should include the requirement that with one hydrogen monitor inoperable, the monitor should be restored to operable status within 30 days or the plant should be brought to at least a hot standby condition within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

If both monitors are inoperable, at least one monitor should be restored to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or the plant 4

should be brought to at least hot standby condition within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

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$_TATUS:

This item has been complied with as addressed in Technical Specification 3/4.8.8.B.

(10) Inrtrumentation for Detection of Inadeaunte Core Coolina (II.F.2)

Subcooling margin monitors, core exit thermocouples, and a reactor coolant inventory tracking system (e.g., differential pressure measurement system designed by Westinghouse. Heated Junction Thermocouple System designed by Combustion Engineering, etc.) may be used to provide indication of the approach to, existence of, and recovery from inadequate core cooling (ICC).

These Instrumentations should be operable during Power Operation, Start-up, and Hot Shutdown i

modes of operation for each reactor.

Subcooling margin monitors should have already been included in the present Technical Specifications.

Technical Specifications for core exit thermocouples and the reactor coolant inventory tracking system should be included with other accident monitoring instrumentation in the present Technical Specifications.

Four core exit thermocouples in each core quadrant and two channels in the reactor coolant tracking system are required to be operable when the reactor is operating in any of the above mentioned modes.

Minimum of two core exit thermocouples in each quadrant and one channel in the reactor coolant tracking system should be operable at all times when the reactor is operating in any of the above mentioned modes.

l STATUS:

These items are being proposed for incorporation into the Post-Accident Monitoring Instrumentation Table 3.8.9-1 in this L

submittal.

These changes are in' accordance with the guidance specified in Generic Letter 83-37.

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i (11) control naam Wahltability Renuirements (!!!.0.3.4) i Licensees should assure that control room operators will be adequately protected against the effects of the accidental release of toxic and/or radioactive gases and that the nuclear power plant can i

be safely operated or shutdown under design basis accident conditions.. If the results of the analyses of postulated accidental release of toxic gases (at or near the plant) indicate any need for installing the toxic gas detection system, it should be included in the Technical Specifications._ All detection systems should be included in the Technical Specifications.

In addition to the above f

requirements, other aspects of the control room habitability i

requirements should be included in the Technical Specifications for the control room emergency air cleanup system.

Two independent control room emergency air cleanup systems should be operable continuously during all modes of plant operation and capable of meeting design requirements.

STATUS:

Technical Specifications to address these concerns will be submitted at a later date.

Several design and licensing issues must be reviewed to assure that the proposed specifications are properly written to address all of the appropriate concerns.

CEto is currently completing several modifications to the Control Room Ventilation System.

The NRC was informed of these modifications in Ceco letter from G.E. Trzyna to A.B. Davis (NRC) dated May 12, 1989.

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s ATTACHMENT 2 PURPOSE OF THE ACCIDENT MONITORING SYSTEM The operability of the accident monitoring instrumentation assures that there is sufficient information available on selected plant parameters to perform manual actions, and to monitor and assess plant status and behavior following an accident.

Regulatory Guide 1.97 specifies the types, design, and qualification criteria that must be met for these instruments.

These instruments are categorized into several-types; Type A, are those that supply information necessary for personnel to perform manual actions during an accident that are essential for direct accomplishment of specified safety functions for which no automatic control is provided.

Type A variables do not include variables associated with contingency actions.

Type B, variables supply information to indicate whether plant safety functions are being accon$11shed.

Type C, variables supply information indicating the potential for or the actual breach of fission product barriers.

Type D, varinoles supply information to indicate the operation of individual safety related or important to safety systems.

Type E, variables supply information used in determining and continuously assessing the magnitude of radioactive releases. All Type A variables for Zion Station were identified in Ziin Stations Regulatory Guide 1.97 revision 2 submittal dated August 1, 1986. As a minimum, these variables are included into the Technical Specifications to assure the availability of information required to perform manual actions following an accident.

These variables are listed in Attachment 8 " Zion Station Type A Variables" of this submittal, identifying which Technical Specification they are incorporated into.

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i ATTACHMENT 3 HISTORY Following is a listing cf correspondence related to this submittal.

This is included to provide a brief history regarding selected variables referenr.ed in this submittal.

October 31, 1980, Post THI Re;utrements from the NRC to All Utilities.

April 20, 1981, Zion Station was issued Amendments 63 and 60 to facility Operating License No,'s OPR-39 and DPR-48.

These amendments incorporated the requirements for implementation of the THI-2 Lessons Learned Category "A"

1tems.

November 1, 1983, Generic Lettrr 83-37 regarding NUREG 0737 Technical Specifications was issued.

August 1, 1986. Commonwealth Edison transmitted to H.R. Denton a review describing how the requirements of Regulatory Guide 1.97 have or will be met at Zion Station. At that time, Commonwealth Edison identified several instruments that required modifications to meet Regulatory Guide 1.97 t

design and qualification criteria.

The following instruments related to this submittal were taentifled:

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- Refueling Water Storage Tank Level, l

Core Exit Thermocouples,

- RCS Subcooling,

- Condensate Storage Tank Level.

i February 2, 1987, Commonwealth Edison re-enforced commitments made to l

upgrade the Core Exit Thermocouples and RCS Subcooling Monitors to meet t

l Regulatory Guide 1.97 revision 2 qualifications for Unit 2 by July 1, r

1987, and Unit I during the Fall 1987 refueling outage.

In addition, commitments were made to address the Refueling Hater Storage Tank Level indicators, and Condensate Storage Tank indicators during the first scheduled refuel outages for Units 1 and 2, nine months following NRC approval of Zion Station's Regulatory Guide 1.97 revision 2 submittal.

May 15, 1987, the NRC issued an interim Technical Evaluation Report regarding Commonwealth Edison's August 1, 1986 Regulatory Guide 1.97 submittal.

This report concluded that the exceptions to Rcgulatory Guide i

1.97 Revision 2 items provided by Commonwealth Edison were acceptable, with the exception of 22 items. Commonwealth Edison was requested to review the interim report and provide a response to these items.

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August 24, 1987, Commonwealth Edison responded to all but 2 of the items from the NRC interim Technical Evaluation Report dated May 15, 1987.

At this time, Commonwealth Edison committed to upgrade the Pressurizer Safety Valve Temperature elements to comply with Regulatory Guide 1.97 revision 2 environmental qualification recommendations.

These upgrades were to be i

k scheduled pending final resolution of all Regulatory Guide 1.97 revision 2 1

i issues, and evaluation of environmental qualification requirements.

December 8, 1987, Commonwealth Edison responded to the two remaining exceptions identified in the May 15, 1987 letter referenced above.

l April 14, 1989, the NRC transmitted Zion Station's Regulatory Guide 1.97 revision 2 Safety Evaluation Report.

The results of this review concluded that Zion Station was acceptable in respect to conformance with Regulatory Guide 1.97 revision 2 in all areas except for; accumulator level and pressure, neutron flux monitors, and containment isolation valve position indication.

These three issues will be responded to separately from this submittal.

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I ATTACHMENT 4 SLHMARY OF PROPOSED CHANGES j

i The Technical Specification changes in this letter are being sulmitted in accordance with the guidance given in NRC Generic Letter 83-37, regarding NUREG 0737 Technical Specifications.

Zion Station's j

Regulatory Guide 1.97 revision 2 submittal was utilized 'n the selection of all appropriate instrumentation.

This submittal tncludes all of the Type A variables identified for Zion Station (Attachment 8).

The instrumentation specified in this submittal currently meet or will be upgraded to meet the design and qualification criteria specified in Regulatory Guide 1.97 revision 2.

The following is a brief description of the proposed changes being submitted and their associated,Jstifications.

The alphabetic designators annotated in the left hand margin, designate the corresponding No Significant Hazards Consideration that the change was evaluated under.

A)

The Table of Contents page lit was revised to change the following items:

the appropriate page numbers were added to the bases sections where they were previously omitted, changed the title for Technical Specification 3.8.9 to " Accident Monitoring Instrumentation", to be consistent with the proposed amendment.

changed the title for Technical Specification 3.12 to " Gaseous Effluent", to match the title of the Specification, changed Technical Specification 3.10.4 title from "End Anchorages and Concrete" to "End Anchorage and Concrete", to match the title of the Technical Specification.

A)

The Table of Contents page v was revised to change the following items:

correct the page references for items 6.2, 6.4, and 6.5 to accurately depict the appropriate pages, and changed the title of Technical Specification 6.3 to " Action to be Taken in the Event of a Reportable Event in Plant Operation" to match the title of the Specification.

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i A) The List of Tables page x was revised to change the following items:

l The page number for item 4.8.9-1 from 192b to 192c due to an additional page associated with this amendment, and the addition of table 4.8.9-2 titled "AccP W Monitoring Instrumentation Numbers".

This table was auc.; to provide a cross reference of the accident parameter to the co,*esponding instrument.

E)

The Mode of Applicability for Specification 3.8.9, Accident Monitoring Instrumentation has been revised to include Mode 7.

Generic Letter 83-37 was written to address the modes of applicability as doftned in the Standardized Technical Specifications.

Mode 7 in the Zian Station Technical Specifications is defined as; less than or equal to 5% power with reactivity and temperature stated per the specific test.

This mode is synonymous with Mode 2, incorporating specified tests from the Special Test Exemptions in the Standard Technical Specificat'ons.

This change will require the Accidert Monitoring Instruments to be operable in Modes 1, 2, 3, and 7.

This change is consistent with the intended Modes for the Post Acciderc Monitoring Instruments in Generic Letter 83-37.

A) Action statements a and b associated with Technical Specification 3.8.9 have been changed to reference Mode 4 as the shutdown requirement instead of HOT SHUTDOWN mode with Tava less than 350 degrees F.

Mode 4 is currently specified in the parenthesis following this statement to provide clarification in regards to the appropriate condition.

This change will eliminate the ambiguity that exists in l

this action, because of the Technical Specifications containing 2 different Hot Shutdown modes, one at greater than 350 degrees F, and the other at greater than 200 degrees F and less ttan or equal to 350 degrees F.

MODE 4 appropriately defines the required condition for the plant to be placed into.

As such, the required mode for the plant to be pla ad into has not been changed.

A)

Surveillance Requirement 3.8.9 has been renumbered to 4.8.9 to be consistent with the 3.0/4.0 numbering and usage rules.

" INSTRUMENT CHANNEL CHECK AND CHANNEL CALIBRATION" have beeft revised to " CHANNEL CHECK and Instrument CHANNEL CALIBRATION", to be in conformance with the wording specified in the definitions section items 1.7, and 1.8.

These changes are correcting inconsistencies in terminology.

The intent of these items has remained the saine.

The asterisk associated with action statement a and the asterisk note on the bottom of page 184 have been deleted.

An asterisked note on the bottom of Table 3.8.9-1, to t'he PZR PORV, and to the PORV Block Valve indicators has been added to replace the asterisks deleted on page 184.

These changes are addressed.'n detail, with the changes associated to the t'?R PORV and PORV Block Valve indicators.

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A) The titles associated with tables 3.8.9-1 and 4.8.9-1 have been relocated to the bottom of the page.

This change was made to place 3

the titles where they would be readily seen when in use.

i B) Containment Pressure (Wide Range), Tables 3.8.9-1 and 4.8.9-1 item I has been revised,' changing reference from narrow range to wide range.

In addition, the total number of instrument channels in Table 3.8.9-1 has been changed from 4 to 2.

These instruments are Type A variables, as identified in Zion Station's Regulatory Guide 1.97 revision 2 submittal.

As such, the instrumentation specified in this table must meet Category 1 design and qualification criteria to assure availability of the parameter under accident conditions.

Zion Station has a total of 2 containment pressure instruments that meet this criteria.

Reference to the Containment Narrow Range Pressure instruments was deleted since these instruments do not meet the design and qualification criteria specified in Regulatory Guide 1.97.

The action statements associated with these changes have remained' unchanged.

With less than the required number of channels operable (2), but not less than the minimum operable number (1), operations may continue for a period of time not to exceed 7 days. With less than the minimum operable channels (1), operations may continue for a period of time not to exceed 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

If, the minimum operable channels or the required number of channels are not restored within these time frames, the Unit will be placed in Mode 4 within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

The survelliances :pecified for these instruments in Table 4.8.9-1 have remained unchanged.

The surveillances specified in Table 4.8.9-1 are appropriate for these instruments to assure operability l

through the performance of 18 month channel calibrations and monthly channel checks.

These changes have been made to reference the appropriate instruments specified in Zion Station Regulatory Guide 1.97 submittal.

D) Steam Generator Water Level (Hide Range) Table 3.8.9-1 1 tem 8 has been revised, changing the Required Number of Channels from a total of 2 to I per Steam Generator, and changing the Minimum Number of Channels from 1 to 2 channels.

Zion Station has a total of I Hide Range Steam Generator Hater Level Instrument per Steam Generator. One channel per Steam Generator was justified in Zion Station's Regulatory Guide 1.97 submittal based on, " additional instrumentation which indicates the effectiveness and adequacy of the secondary heat sink".

The additional instrumentation listed were; the Steam Generator Narrow Range Level, with backup from the Reactor Coolant System Temperature i

and Reactor Coolant Pump Flow.

The Steam Generator Narrow Range Level Instruments cover over 257, of the Steam Generators Wide Range Level u

l Instruments range.

By changing the required number to 1 per Steam Generator, a total of 4 channels will now be required instead of a total of 2.

Based on the current Technical Specification's, operations could continue for a period of 7 days based on 3 Steam Generator Hide Range Level Instruments failing.

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The proposed changes will allow operations to be continued for a period of 7 days based on a single instrument failure for any Steam Generator.

Seven days is considered to be an acceptable period of time for a single instrument failing.

If, the Required Number of Channels are not restored within this time frame, the unit will be placed in Mode 4 (HOT SHUTDONN) within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

By specifying a total of 2 channels as the Minimum N.1ber of Channels, action statement b of Specification 3.8.) requiring restoration in 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.is still applicable.to this item.

The current Zion design has 1 channel of Hide Range Water Level Instrumentation per Steam Generator.

By requiring a total of 2 chdonels to be operable, any combination of 2 Steam Generator Wide Range Indicators would fulfill this requirement.

The surveillances specified for these instruments i

in Table 4.8.9-1 have remained unchanged.

The surveillances specified in Table 4.8.9-1 are appropriate for these instruments to assure operability through the performance of 16 month Channel Calibrations and Monthly Channel Checks.

Based on thf above information it can be concluded that the overall changes being proposed for the Steam Generator Hide Range Level Instruments a'e more conservative than the current Specifications.

In addition, tre change is justifiable based on the availability of diverse instrumeatation (Steam Generator Narrow Range level Instruments) that provides the same information, thus assuring that a complete loss of indication for the parameter being monitored will not occur.

I D) Auxiliary Feedwater Flow Rate Table 3.8.9-1_ item 10 has been revised, changing the Required Number of Channels from a total of 2 to 1 per Steam Generator, and the Minimum Numbe of Channels from 1 to 2.

Zion Station has a total of 1 Auxillary bedwater Flow Rate Instrument

.j per Steam Generator, and 3 Auxiliary Feedwater Pump Systems.

Each of the 3 pump systems are capable of providing flow to all 4 Steam Generators.

The instrumentation provided indicates the combined flow rate to each Steam Generator from all 3 Auxillary Feedwater Pump Systems.

The 4 Auxiliary Feedwater Flow Rate instruments are powered from redundant instrument busses, 2 from instrument bus 111/211 and L.

the remaining 2 from instrument bus 112/212, Backup indication is available from the Steam Generator Narrow Range Instruments, which is a direct indication of heat sink statu..

These instruments are Type A i

Category I variables. One channel per Steam Generator was reviewed and accepted in Zion Station's Regulatory Guide 1.97 submittal based on the above information.

By changing the Required Number of Operable Channnels to 1 per Steam Generator, a total of 4 channels will now be required instead of a total of 2 which is currently in the Zion Technical Specifications.

Based on the current Technical Specification's, operations could continue for a period of 7 days based on 3 Auxiliary Feedwater Flow Rate Instruments failing.

The proposed changes will allow operations to continued for a period bf 7 days based on a sinale instrument failure on any Steam Generator.

Seven days is considered to be an acceptable period of time for a single instrument failing.

If, the Required Number of Channels are not restored within this time frame the unit will be placed in Mode 4 within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

l 19140 ( u )

i By specifying a Minimum Number of Channels of 2. Using I channel per Steam Generator, would require any combination of 2 Steam Generator Auxillary Feedwater flow Rate Channels to be operable.

Action statement b of Specification 3.8.9 requiring restoration in 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> is still applicable for this item. As discussed in the Steam Gentrator Water Level Wide Range Instrumentation section above.

The surveillances spectfled for these instruments in Table 4.8.9-1 have remained ui; changed.

The surveillances specified in Table 4.8.9-1 are appropriate for these instruments to assure operability through I

the performance of 18 month Channel Calibrations and Monthly Channel 4

Checks.

Based on the above information it can be concluded that the i

overall changes being proposed for the Auxillary Feedwater Flow Rate Instruments are more conservative than the current Zion Technical Specifications, and that the change is justifiable based on the availability of diverse instrumentation (Steam Generator Narrow Range i

Instruments) that provides the same information, thus assuring that a complete loss of indication for the parameter being monitored will not occur.

B) Reactor Coolant System Subcooling Margin Table 3.8.9-1 item 11 has been revised, deleting reference made to procedural calculations.

Procedural calculations were an interim measure used until additional instrumentation was made available or upgraded.

The reactor coolant system subcooling monitor consists of two independent monitoring systems.

With less than the required number of channels operable (2),

but not less than the minimum operable number (1), operations may continue for a period of time not to exceed 7 days.

With less than the minimum operable channels (1), operations may continue for a period of time not to exceed 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

If, the minimum operable l

channels or the required number of channels are not restored within l

these time frames the unit will be placed in Mode 4 within the next 12 i

hours.

The surveillances specified for these instruments in Table 4.8.9-1 have remained unchanged.

The surveillances specified in Table 4.8.9-1 are appropriate for these instruments to assure operability through the performance of 18 month channel calibrations and monthly channel checks.

These instruments are Type A Category i variables.

In Zion Station's Regulatory Guide 1.97 submittal, a commitment was made to either environmentally qualify or replace unqualified components associated with these instrurrnts with safety related components.

These modifications have bu n completed.

As such, the Reactor Coolant System Subcooling Margin Instruments being referred to by this change now meet Category 1 design and qualification criteria.

This change is being made to place this item into conformance with the 1

guidance given in Generic Letter 83-37.

19140 ( u)

3 PZR PORV Position Indicator Table 3.8.9-1 item 12 has been revised.

changing the total and required number of channels from 2 to 1, and deleting reference to the double asterisk statement specifying the primary and backup PORY position indicators.

In addition, an asterisked note has been added to the bottom of Table 3.8.9-1, and asterisks added to the PZR PORV, its Required Number of Channels, and Minimum Number of Channels.

These items have been added to take the i

place of the asterisk note deleted from the bottom of page 184.

The proposed charges regarding the total and required number of channels, and deletion of the double asterisk statement are summarized as follows:

The Zion Station Technical Specifications currently require 2 position indicators per PORV.

These indicators are designated as: 1) stem mounted limit switch, and 2) acoustical monitors which provides indication of flow through the PORV ta11 pipes.

The current Zion design consists of 1) stem mounted limited switches, 2) temperature indication, and 3) acoustical nonitors on the Pressurizer Safety Relief Valves all of which provides the control room with PORV Position Indication.

The stem mounted limit switches are the primary position indication for the PORVs.

These limit switches not only provide positive valve indication, they also are safety grade, powered from an smergency power supply, seismically and environmently qualified with their displays and wntrols located in the control room as required by NUREG 0737.

CECO has provided this information in response to NUREG 0737.

In NRC letter dated February 29, 1980, section 11.0.2 it was stated

~

that the stem mounted limit switches are in compilance with the short term Lessons Learned requirements.

CECO has enhanced the stem mounted limit switches by upgrading them to meet safety related Category 2 instrumentation criteria as required by Regulatory Guide 1.97.

Iion also has temperature indicators on the PORV tall pipes downstream from the PORVs.

This irdication eti be used, in addition to the stem mounted limit switches, is provide 80RV position indication.

If the PORVs were to open, these temperatore indicators would sense the increase in temperature thus indicating flow through the piping.

This temperature increase would be seen in the control room by the operators and appropriate actions taken.

In addition, if the PORVs l.

would happen to open, thus allowing flow through the valve, it would go directly to the Pressurizer Relief Tank (PRT).

The control room would see this increase in PRT level through the PRT level instrumentation in the control room as well as receive various alarms-The-Zion E0Ps contain specific guidance to the operators on what actions to take and how to determine if the PORVs do open, during e

an accident.

19140 (1s)

r A third instrument available to further assist the operator in determining if a PORY opens is the Tcoustical monitors that currently ea!G on the Pressurizer Safety Relief Valves.

These monitors are in clo e pretmity to the PORVs.

In the event that a PORV does actuate, the opening would be picked up by the acoustical monitors.

Since the control roco hat direct indication of these acoustical monitors, the operators would be aware of the PORVs' change in position and take appropriate action.

These acoustical monitors are physically located close to the PORVs and provide additional indication to the control room that the valves have actuated.

Two PORV position indicators per valve (stem mounted limit switches and temperature indication) were reviewed and accepted by the NRC in l'

Zion Station's Regulatory Guide 1.97 submittal.

NRC has stated that the only basis for a Technical Specification on backup indication of safet" or relief valve position is when a non-safety grade channel is provin g as the primary indication of valve position.

Since Z1'on now has safety grade PORV stem mounted limit switches, there is no need to put the other backup systems in the Technical Specifications.

The proposed change allows for the deletion of the acoustical monitoring j

backup indicators from the Technical Specifications.

Based on the current Technical Specifications, operations could continue for a period of 7 days based on the inoperability of either a stem mounted limit switch, or acoustical monitors.

If both of these indicators were inoperable, operations could continue for a pertold of time not i

to exceed 4R hours.

In the event that the required number of channels or minimum operable channels are not restored within these time frames, the unit will be placed in Mode 4 within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

?

The proposed changes will allow operations to continue for a period of l

time not to exceed 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> b6 sed on the inoperability of the stem mounted limit switch alone.

If the required number of channels (same as the minimum operable number) are not restored within this time frame, the unit will be placed in mode 4 within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

As such, if the acoustical monitors or the temperaturo indication is assumed to be inoperable, the action required would remain the same.

The surveillances specified for these instruments in Table 4.8.9-1 have remained unchanged.

The proposed change still requires periodic surveillances to be performed on the stem mounted limit switches.

The surveillances specified in Table 4,8.9-1 are appropriate for this instrument to assure operability through the performance of the 18 month channel calibrations (indicator testing) ano monthly channel checks.

A)

The proposed changes regarding the addition of an asterisked note to the bottom of Table 3.8.9-9, and asterisks added to the PZR PORV, its required number of channels and minimum operable channels, are summarized as follows:

The note being deleted on the bottom of page 184 provided an exemption to the shutdown requirements associated with an inoperable PZR PORV or PORV Block Valve Position Indicator.

This exemption was contingent on l

the position of the associated PORV Block Valve being known to be closed or verified closed within 7 days.

19140 ( u)

i e

Action statements a and b associated with specification 3.3.1.F requires the PORV Block Valve associated with either an inoperable PORV or PORV Block Valve to be closed and de-energizett anytime either of these valves is found inoperable in modes 1, 2, or 3.

These actions must be taken within one hour, or a unit shutdown is required.

The position indication for the PORV Block Valve is powered from the same source required to be de-energized in action statements a and b of Specification 3.3.1.F.

This removes power from the PORV Block Valve and PORV Block Valve Indicator making both the valve and indicator inoperable. Specification 3.8.9 currently allows indefinite operation without PORV position indication or PORV Block Valve e

position indication, when the position of the associated PORV Block j

Valve is known to be closed or verified closed within 7 days, Scecification 3.3.1.F actions a and b require the PORV Block Valve to be closed and de-energized within I hour.

The proposed change deletes the requirement to maintain the PORV or PORV Block Valve associated with an isolated flowpath operable, when the flowpath has been isolated in accordance with Specification 3.3.1.F.

As such, these changes result in no net change to the existing requirements.

PZR PORV Block Valve Position Indicator Table 3.8.9-1 item 13 has been revised.

The Minimum Number of Channels is changing from 0 to 1, an asterisked note has been added to the bottom of Table 3.8.9-1 and to the PZR PORV Block Valve Indicator, its Required Number of Channels, and Minimum Number of Channels.

These items have been added to take the place of the asterisk note deleted from the bottom of page 184 as addressed above for the PORV Position Indicators.

The proposed change regarding the Minimum Number of Channels is summarized as follows:

7 D)

Zion P atton has a total of 1 valve limit switch position indicator per PORV Block Valve.

The PORV Block Valve Position Indicators are Safety Related Type D Category 2 variables.

Two i

position indicators per PORV Block Valve were reviewed and accepted in Zion Station's Regulatory Guide 1.97 submittal.

Backup indication of safety or relief valve position is only required if a non-safety grade channel is provided as the primary I

indication of valve position.

This logic has been applied to the PORV Block Valves as weII.

The current Zion design provides safety grade PORV indication poored from an emergency power supply.

As such, the Required dumber of Channels has remained unchanged.

The Minimum Number of Channels has been revised to 1 per valve, this is consistent with the guidance given in Generic Letter 83-37.

Based on the current Technical Specifications, operations could continue for a period of 7 days based on the inoperability of the PORV Block Valve Position Indicator.

19140 ( n)

. ~.

a 1

If, the Required Number of Channels is not restored within this time frame, the unit will be placed in Mode 4 within the next 12

)

hours.

The proposed changes will allow operations to continue for a period of time not to exceed 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> based on the inoperability of the PORV Block Valve Position Indicator.

If, the Required Number of Channels (the same as the Minimum Number of Channels)'are not restored within this time frame, the unit will be placed in Mode 4 within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

The surveillances specified for this parameter in Table 4.8.9-1 have remained unchanged.

The proposed change will still require 1

periodic surveillances to be performed on the PORV Block Valve Position Indicators.

The surveillances specified in Table 4.8.9-1 are appropriate for this instrument to assure operability through the performance of 18 month Channel Calibrations (Indicator testing) and Monthly Channel Checks.

A)

The proposed changes regarding the addition of an asterisked note to the bottom of Table 3.8.9-1, and asterisks added to the ?7R PORV Block Valve Position Indicator, Required Number of Channe h and Minimum Number of Channels are summarized as follows:

The note being deleted on the bottom of page 184 provided an exemption to the shutdown requirements associated with an i

inoperable PZR PORV or PORV Block Valve Position Indicator.

This exemption is contingent on the position of the associated PORV Block Valve being known to be closed or verified closed within 7 days. Action statements a and b associated with specification 3.3.1.F requires the PORV Block Valve associated wtth either an inoperable PORV or PORV Block Valve to be closed and de-energized I

anytime either of these valves is found inoperable in modes 1, 2, l'

or 3.

These actions must be taken within one hour, or a unit i

shutdown is required.

The position indication for the PORV Block Valve is powered from the same source required to be de-energized in action statements a and b of Specification 3.3.1.F.

This removes power from the PORV Block Valve and PORV Block Valve Indicator making both the valve and indicator inoperable.

Specification 3.8.9 currently allows indefinite operation without FORV position indication or PORV Block Valve position indication when the position of the associated PORV Block Valve is known to be closed or verified closed within 7 days.

Specification 3.3.1.F actions a and b require the PORV Block Valve to be closed I

and de-energized within I hour.

The proposed change deletes the l

requirement to maintain the PORV or PORV Block Valve associated L

with an isolated flowpath operable, when the flowpath has been isolated in accordance with Specification 3.3.1.F.

As such, these changes result in no change to the existing requirements.

A) An asterisk note has been added to the bottom of Table 4.8.9-1 and asterisks have been added to PZR PORV and PORV Block Valve Position l

Indicators.

These changes are being proposed in conjunction with the above changes proposed on the PORV and PORV Block Valve Indicators.

I 19140 (is)

1 The PORV and PORV Block Valve Position Indicators are not required to be operable, when their associated flowpath is isolated as previously addressed.

These changes will exempt the requirement to perform periodic surveillances on these indicators when the PORV Block Valve i

is closed and de-energized in accordance with Specification 3.3.1.F.

The exemption from performing periodic surveillance on the PORV and 4

PORV Block Valves currently exists under Specification 4.0.3.

Specification 4.0.3 does not require surveillances to be performed on i

inoperable equipment.

As such, this change results in no change to the existing requirements.

B)

PZR Safety Valve Position Indicator (Primary Detector) Tables 3.8.9-1 anti 4.8.9-1 item 14 have been revised, referencing this item as the F2R Safety Valve Position Indicator, primary indicator.

In addition, the required number of channels has been changed from a total.of 2 to 1 per valve, the minimum operable channels from a total of I to 1 per valve, and reference to a double asteri k note stating the origin of this item has been added.

These chang ' re being proposed in l

conjunction with proposed item number on Tables 3.8.9-1 and 4.8.9-1, requiring the operability of a backup indicator.

Zion Station has I thermocouple located in the discharge line of each Safety Valve.

These monitors provide an indication of safety valve position in the control room.

Temperature was selected as the primary I

detector because, through the use of temperature indication it is l

possible to detect both large and small flows past the valve.

These instruments are non-safety related Type O Category 3 variables.

A commitment was made to upgrade these instruments to Category 2 qualifications in the supplemental report to Zion Station's Regulatory Guide 1.97 submittal dated August 24, 1987. The use of these detectors as a means of determining Pressurizer Safety Valve position was reviewed and approved in Zion Station's Regulatory Guide 1.97 submittal. The date for modificatir

' these instruments to Category 2 l

criterie has not been determined yet.

e dates will be submitted in Zion Station's Regulatory Guide 1.97 t=t ~1ance response. Zion Station's Regulatory Guide 1.97 compliancc response will be submitted I

by October 31, 1990. Based on the current Technical Specifications.

l operations could continue for a period of 7 days based on 2 PZR Safety i

Valve Position Indicators falling, and 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> in the event of all 3 instruments failing.

The proposed changes will allow operations to continue for a period of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> based on a sinale failure of a PZR Safety Valve Position Indicator.

If the required number of channels (sam as the minimum operable channels) are not restored within this time frame the unit will be placed in Mode 4 within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The surveillances specified for these instruments in Table 4.8.9-1 have remained unchanged.

The survalliances specified in Table 4.8.9-1 are appropriate for these instruments to assure operability through the performance of 18 month channel calibrations and monthly channel checks.

The double asterisk statement added to the PZR Safety l

Valve Position Indicator and to the bottom of this table provide l

reference to the origin of this requirement being NUREG 0578 Item 2.1.3.a.

l 19140 (1,)

1 B) PZR Safety Valve Position Indicator (8ackup Detector) has been added l

to Accident Monitoring Instrumentation Tatles 3.8.9-1 and 4.8.9-1 as

)

item 15, specifying a total of 3 (1/ valve) channels, 1/ valve as the

?

required number, and Not applicable as the minimum operable number.

This change is being proposed in conjunction with item 14 above.

Zion Station has I acoustic monitor per Safety Valve.

These monitors provide indication of safety valve position in the control room.

These instruments are type D Category 3 variables.

The use of these instruments as a backup means of determining Pressurizer Safety Valve position was reviewed and approved in Zion Station's Regulatory Guide I.97 submittal.

With less than the required number of channels operable (1/ valve), operations may continue for a period of time not to exceed 7 days.

If, the required number of channels are not restored within this time-frame the unit will be placed in Mode 4 within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

By specifying Not Applicable as the minimum operable rutber, actl%i statement b of Specification 3.8.9 requiring restoratlin ir,si i.ours is not applicable to this item.

This item has been added to Table 4.8.9-1 to assure operability of this par ameter through tts performance of periodic surveillance tests.

The surveillances specified in Table 4.8.9-1 are appropriate for these instruments to assure opi ability through the performance of 18 month channel calibrations and monthly channel checks.

This change is consistent with the NRC requirements of NUREG 0737 and guidance provided in the NRC letter concerning Regulatory Guide 1.97 compliance (NRC letter dated April 14, 1989).

P B) Core Exit Thermocouples have been added to Accident Monitoring Instrumentation Tables 3.8.9-1 and 4.8.9-1 as item 16.

The Zion design has two channels of Core Exit Thermocouples. One channel has 32 thermocouples and the other has 33.

The Required Number of Channels would consist of 2 channels.

At least 2 Core Exit Thermocouples per channel per quadrant are required in Table 3.8.9-1 and 4.8.9-1.

With less than the Required Number of Channels Operable.

l operations may continue for a period not to exceed 7 days.

In the I

case of the Minumum Channels Operable, there would be I channel operable consisting of at least 2 Thermocouples per quadrant.

With less than the Minimum Operable Channels, operations may continue for a period of time not to exceed 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

This change is consistent with L

the guidance provided by the NRC in letter dated November 1, 1983.

That letter provided Technical Specificattons that would be acceptable for the utility.

For the In Core Thermocouples (T/Cs) it stated that I

there should be a minimum of 4 T/Cs per quadrant as the Required Number of Channels.

In addition, the NRC proposed Technical Specifications have a Minimun Number of Operable Channels as 2 per quadrant.

The Zlon proposed Technical Specifications are in agreement however, the Zion Technical Specifications provide more r.larification specifically showing that there are two trains to T/Cs to be considered.

l l

19140 (ro)

1 If, the minimum operable channels or the required number of channels are not restored within these time frames, the unit will be placed in Mode 4 within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

This item has been added to Table 4.8.9-1 to assure operability of this parameter through the performance of periodic survelliance tests.

Requiring a minimum of I channel to be operable ensures that at least 2 T/Cs are operable in j

each quadrant of the reactor core.

This provides sufficient monitoring capability.

The surveillances specified in Table 4.8.9-1 are appropriate for these instruments to assure operability through the performance of 18 month channel calibrations and monthly channel i

checks.

These instruments are Type A Category 1 variables.

The l

modifications to environmentally qualify and replace safety related components, stated in Zion Station's Regulatory Guide 1.97 submittal have been completed.

As such, the Core Exit Thermocouples referred to in this submittal now moet Category 1 design and qualification criteria.

These changes are being made consistent with the guidance j

specified in Generic Letter 83-37.

B) Containment Water Level (Narrow Range) has been added to Accident Monitoring Instrumentation Tables 3.8.9-1 and 4.8.9-1 as item 17.

There are a total of 2 Narrow Range Containment Water Level Instruments, with I specified for the required number and I for the minimum number.

The required and minimum operabia number of channels have a pound sign notation associated with their action that provides for an allowable outage time of 30 days.

This note has been added to i

the bottom of page 192b. As such, with less than the required number of channels operable or the minimum operable number, operations may continue for a period of timt not to exceed 30 days.

If, the minimum operable channels (same as the required number of channels) are not restored within this time frame the unit will be placed in Mode 4 within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

This item has been added te Table 4.8.9-1 to assure operability of this parameter through the performance of periodic survelliance tests.

The surveillances specified in Table 4.8.9-1 are appropriate for these instruments to assure operability through the performance of 18 month channel calibrations and monthly channel checks.

These changes are being made consistent with the guidance specified in Generic Letter 83-37.

B) Containment Water Level (Hide Range) has been added to Accident Monitoring Instrumentation Tables 3.8.9-1 and 4.8.9-1 as item 18.

There are a total of 2 Wide Range Containment Water Leve! Instruments, with 2 specified for the required number, and 1 for the minitum operable number.

With less than the required number of channefs operable (2), but not less than the minimum operable number (1),

operations may continue for a period of time not to exceed 7 days.

l l

Hith less than the minimum operable channels (1), operations may continue for a period of time not to exceed 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

If, the minimum operable channels or the required number of channels are not restored within these time frames the unit will be placed in Mode 4 within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

This item has been added to Table 4.8.9-1 to assure operability of this parameter through the performance of periodic surveillance tests.

The surveillances specified in Table 4.8.9-1 are appropriate for these instruments to assure operability through the performa.nce of 18 month channel calibrations and monthly channel checks.

Thesc changes are being made consistent with the guidance specified in Generic letter 83-37.

19140 (n)

i

/

B) The Reactor Coolant Inventory Tracking System has been added to i

Accident Monitoring Instrumentation Tables 3.8.9-1 and 4.8.9-1 as item 19 and 20.

The Reactor Coolant Inventory Tracking System has been i

divided into 2 separate line items.

The reason for this change is to address the two different operating modes for this system.

The Reactor Coolant Inventory Tracking System at Zion Station works on a differential pressure principle.

This system consists of 4 channels, two of which are calibrated for Reactor Coolant Pumps in operation and the other two for pumps off.

Reactor Vessel Water Level (Wide Range at least one RCP running)

Table 3.8.9-1 item 19 has a total of 2 instruments, with 2 specified for the required number, and 1 for the minimum number.

With less than the reqdired number of channels operable (2), but not less than the minimum operable number (1), operations may continue for a period of time not to exceed 7 days.

With less than the minimum operable channels (1), operations may continue for a period of time not to exceed 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

If, the minimum operable channels or the required number of channels are not j

restored within these time frames, the unit will be placed in i

Mode 4 within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

This item has been added to Table 4.8.9-1 to assure operability of this parameter through the performance of periodic surveillance tests.

The surveillances specified in Table 4.8.9-1 are appropriate for these instruments l

to assure operability through the performance of 18 month channel calibrations and monthly channel checks.

These changes are being made consistent with the guidance specified in Generic Letter 83-37.

Reactor Vessel Hater Level (Narrow Range, all RCPs stopped) Table 3.8.9-1 item 20 has a total of 2 instruments, with 2 specified for the required number, and 1 for the minimum number.

With less than the required number of channels operable (2), but not less than the minimum operaole number (1), operations may continue for a period of time not to exceed 7 days.

With less than the minimum operable channels (1), operations may continue for a period of time not to exceed 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

If, the minimum operable L

channels or the required number of channels are not restored l

.ithin these time frames the unit will be 71 aced in Mode 4 within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

This item has been adL J to Table 4.8.9-1 to i

l assure operability of this parameter through the performance of periodic surveillance tests.

The turveillances specified in Table 4.8.9-1 are appropriate for these instruments to assure operability through the performance of 18 month channel calibrations and monthly channel checks.

These changes are being made consistent with the guidance specified in Generic Letter 83-37.

l 19140 (22)

~ -.

r

8) Condensate Storage Tank Level has been added to Accident Monitoring Instrumentation Tables 3.8.9-1 and 4.8.9-1 as item 21.

There are a total of 2 Condensate Storage Tank Level indicators, with 2 specified as the required number, and I for the minimum operable r.?mber.

These instruments are Type A Category 1 variables.

Neither of the 2 Condensate Storage Tank Level instruments currently meet the design or qualification criteria of Regulatory Guide 1.97.

The modifications to upgrade these instruments will be completed for 1) nit I and 2 during their cycle 13 Refueling Outages. After completion of these modifications, the Condensate Storage Tank Level Indicators will meet Category-1 design and qualification criteria.

With less than the required number of channels operable (2), but not.less than the minimum operable number (1), operations may continue for a period of time not to exceed 7 days.

With less than the minimum operable channels (1), operations may continue for a period of time not to exceed 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

If, the minimum operable channels or the required number of channels are not restored within these time frames, the unit will be placed in Mode 4 within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

This item has been added to Table 4.8.9-1 to assure operability of this parameter through the performance of periodic surveillance tests.

The surveillances specified in Table 4.8.9-1 are appropriate for these instruments to assure operability through the performance of 18 month channel calibrations and monthly channel checks.

These changes are being made consistent with the guidance specified in Generic Letter 83-37.

A) H: Monthly, and R: Refueling have been deleted from the bottom of page 192b.

This item is not necessary since the frequencies associated with these two items are appropriately defined in the Definitions Section of the Technical Specifications. Table 1.2.

A)

Table 4.8.9-2 " Accident Monitoring Instrument Numbers" has been developed to provide a cross reference between the accident parameter and the instrument used.

These instruments are consistent with those submitted in Zion Station's Regulatory Guide 1.97 submittal.

This table is intended to be a reference aid for personnel.

No actions or requirements are imposed or altered as a result of this table.

A)

Bases Page 195 was revised, to change the following items:

page format to be consistent with that used for other bases pages after revision, added reference tu NRC Generic Letter 83-37 into the bases for the accident menitoring instruments, and placing the revision number and date into the Bases to include which revision of Regulatory Guide 1.97 Zion Station's submittal was approved to.

19140 (n)

L' P

C) Action Statement 4 has been added to Table 3.12-1.

In addition, the actions associated with the Steam Generator Atmospheric Relief and Safety Valve Radiation Monitors, and the. Vent Stack' Noble Gas Effluent Radiation Monitors channels 7, and 9 have been revised to reference action 4.

Action Statement 4 was chosen because it is not currently used in the Technical Specification.

This action statement has been modeled after action 30 of Generic Letter 83-37.

Only the actions associated with the Steam Generator Atmospheric Relief and Safety Valves, and the Vent Stack Noble Gas Effluent Radiation Monitors channels 7, and 9 have been revised to reference action 4.

These are the only effluent path radiation monitors that must be considered accident monitors for the following reasons:

The Vent Stack Noble Gas Monitor Channels 7. and 9 are located on the Unit I and 2 Ventilation Stacks downstream of all inputs into the stacks.

The Ventilation Stacks are a common release point for all identified release pathways for gaseous radioactive materials with the exception of the Steam Generator Atmospheric Relief and Safety Valve discharges paths.

L Steam Generator Atmosphoric Relief and Safety Valve discharges paths are the only exception to the above statement.

As such this pathway has also been included for the purpose of accident monitoring.

This position was reviewed and accepted in Zion Station's Regulatory Guide 1.97 submittal, l

The Vent Stack Noble Gas Monitor Channels 7 and 9 and the Steam Generator Atmospheric Relief and Safety Valves Radiation Monitors have been changed to reference action 4.

Based on the current Technical Specifications with these monitors inoperable, the inoperable monitor must be restored to operable status within 30 days, or an alternate means of monitoring the parameter must be estabitshed.

The proposed changes will require an alternate method of monitoring the appropriate parameter to be established within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, and to restore the inoperable monitor to operable status within 7 days of the event, or to prepare and l

submit a report to the commission within 14 days following the l

event outlining the actions taken, the cause of inoperability, l.

and the plans and schedule for restoring the monitor to operable status.

The surveillances specified for these instruments in l

Table 4.12-2 have remained unchanged.

These surveillances are l.

appropriate for these instruments to assure operability.

These I

changes are consistent with the requirements of Generic. Letter i

83-37.

L A)

Item number 7.C.4, on page 236b has been changed from "1R-AR04B" to L

"2R-AR04B".

This change has been made to correct a typographical error.

This monitor should have been referenced as a Unit 2 Monitor.

1 19140 (n)

C) Action Statement 31 has been added to Table 3.14-1.

In addition, the actions associated with the Containment Area High Range Monitors have i

been revised to reference action 31.

Action Statement 31 was chosen becauie it is not currently used in the echnical Specification.

This action statement has been modeled after action 30 of Generic Letter 83-37.

Based on the current Technical Specifications, with these i

monitors inoperable, an alternate method of monitoring the appropriate parameter (if feasitle) must be hittated within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, and the j

inepsrable monitor must be restored to operable status within 7 days of the event, or a Station Review will be conducted within 14 days to 1

determine a plan of action to restore the monitor to operable status.

The proposed changes will require an alternate method of monitoring the appropriate parameter to be established within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, and to restore the inoperable monitor to operable status within 7 days of the event, or to prepare and submit a report to the commission within 14 days following the event outlining the actions taken, the cause of inoperability, and the plans and schedule for restoring the monitor to operable status.

The surveillances specified for these instruments in Table 4.14-1 have remained unchanged.

These surveillances are anpropriate far these instruments to assure operability.

These l-changes are consistent with the requirements of Generic Letter 83-37.

C) Section 6.2.L. has been revised to include the requirement to have Post Accident sampling programs capable of obtaining and analyzing reactor coolant and containment atmosphere, and collect and analyze or measure radioactive lodines and particulates in plant gaseous

[

effluents samples under accident conditions.

This program includes; the training of personnel, procedures for sampling and analysis, and provisions for maintenance of sampling and analysis equipment.

These changes are consistent with the guidance given in Generic Letter 83-37.

As a result of adding this item, Specification 6.2.2 has been moved to page 310, and page 310 has been included as a revised page.

l A)

Item r has been added to Special Report Table 6.6-1 to address the 14 day written report that must be submitted any time a Post-Accident Radiation Monitor has been inoperable in excess of 7 days. Addition of this item reflects the addition of this reporting requirement to the Technical Specifications.

I Boric Acid Tank Solution Level item number 10 of Generic Letter 83-37 was not incorporated into this submittal.

The Boric Acid Tank Solution Level is not a Type A variable at Zion Station.

Regulatory Guide 1.97 does not contain the Boric Acid Tank Level under any variable type.

The safety function of the Boric Acid Storage Tanks is to maintain sufficient boric acid quantity, to borate the Reactor Coolant System to cold shutdown concentration anytime during the core cycle consistent with the Technical Specifications shutdown margin requirements.

There are no automatic or manual actions required for Design Basis Accident mitigation associated with the Boric Acid Storage Tank.

19140 (as)

ATTACHMENT 5

1. Copy of the current Technical Specifications, marked-up to denote the changes being proposed.

Pages: til v

x 184 192a

'192b

.192c 195 236a 236b 237 251 252a 309 310 324

2. Copy of the typed change being proposed.

Pages: til y

X 184 192a 192b 192c 192d 192e 195 236a 236b 237 251 252a i

309 i

310 l

324 1

1 19140 (as) i i

g 2

- SURVER LANCE LlHITING CONDITION FOR OPERATION.

REOUIGMENT

' P_A_fd 3.7 Steam Generator Emergency Heat Removal

- 4.7 156 3.7.1 Steam Generator Safety Valves _

4.7.1 156' 3.7.2 Auxillary Feedwater Pump System 4.7.2

. 158 3.7.3 Auxillary Feedsater Supply System 4.7.3 159a l

Bases 162 3.8 Emergency Core Coo 1 W, md Core Cooling Support 4.8 164 3.8.1 Centrifugal Charging Pug *vstem' 4.8.1 164 3.8.2 Safety Injection Pump System 4.8.2 168 3.8.3 Residual Heat Removal Pump System 4.8.3 170 3.8.4 System Testing of Centrifugal Charging, Safety 4.8.4 173 F

Injection, and Residual Heat Removal Pump Systems 3.8.5 Accumulator System

. 4.8.5 174 3.8.6 Component Cooling System 4.8.6 175 3.8.7 Service Mater System 4.8.7

' 178 3.8.8 Hydrogen Control Systems 4.8.8 180

'3.8.2 Accident Monitoring Instrumentation 4.8.9 184' Bases 193 3.9 Containment Isolation Systems 4.9 197 3.9.1 Isolation Valve Seal Water System 4.9.1 197 3.9.2 Penetration Pressurization Systems 4.9.2 198 3.9.3 Containment Isolation Valdes 4.9.3 199a 3.9.4 Main Steam Isolation Valves and Bypasses 4.9.4 200 3.9.5 Containment Integrity 4.9.5 201 i

Bases 209 3.10 Containment Structural Integrity 4.10 212 3.10.1 Containment Leakage Rate' Testing 4.10.1 212 3.10.2 Containment Air Locks 4.10.2 214a 3.10.3 Containment Tendons 4.10.3 215 3.10.4 End Anchorage and Concrete 4.10.4 217

').10.5 Containment Pressure 3.10.5 219 3.10.6 Containment Temperature w.10.6 219 I

Bases 220 3.11 Radioactive Liquids 4.11 222 Bases 229 l

3.12 Gaseous Effluents 4.12 230 l

Bases 242 TABLE OF CONTENTS (Continued) 19140 Ii1 TSC 90-02

.,mo-

-mswJ-

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+y

--vw.c-w---

,.p

'r 1

--w,,v-ww

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=,

. u..

~n gj

" ?

5.0 Design Features 296'-

1-a 5.1 Site 296 5.2 Reactor Coolant System

'296-5.3 Reactor Core 296 5.4 Containment System............

296; 5.5 Fuel Storage 298 5.6 Seismic Design..........................

299 6.0 AQtiINISTRATIVE CONTROLS.................................

300 6.1 Organization, Review, Investigation and Audit....................

300 l

6.2 Plant Oper7 ting Procedures 309 l

6.3 Action to be Taken.in the event of a Reportable Event in Plant Operation 310 l

6.4 Action to be Taken in the Event A Safety Limit is Exceeded 311 l

l 6.5 Plan' Operating Records...............................

311 l

6.6 Reporting Requirements.... --......................-.....

312 6.7 Offsite Dose Calculation Manual (ODCM).......................

325 6.8 Flooding Protection 326 Table of Contents (Continued) v.

19140 TSC 90-02.

.... ~

.=

~

..... i-Iahle Eage 4.8-1 Centrifugal Charging 0 mp System 185' 4.8-2 Safety Injection Pump System.

186 4.8-3 Residual Heat Removal Pump System 187 4.8-4 Accumulator Tanks 188 4.8-5 Component Cooling Pump System 189 4.8-6 Service Water Pump System 190 4.8-7 Hydrogen Control System 192 4.8.9 1 Accident Monitoring Instrumentation Surveillance Requirements 192c 4.8.9-2 Accident Monitoring Instrument Numbers 192d 4.9-1 Isolation Seal Water System 203 4.9-2 Penetration Pressurization System 204 4.11-1 Radioactive Liquid Effluent Sampling and Analysis 227 Surveillance 4.11-2 Radioactive Liquid Effluent Monitoring Instrumentation 228b Surveillance 4.12-1 Radioactive Gaseous Effluent Sampling and Analysis Program 238 4.12-2 Radioactive Gaseous Effluent Monitoring Instrumentation 240 Surveillance 4.14-1 Plant Radiation Monitoring Instrumentation Surveillance 253 Requirements 4.15-1 4160-Volt Engineered Safeguard Bus Main, Reserve, and 270 Standby Feeds 4.16-1 Maximum Values for the Lower Limits of Detection (LLD) 280 List of Tables (Continued) a.

-u

3 m

.y

~

y -. ~

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT ~

l ACCIDENT MONITORING INSTRUMENTATION ACCIDENT MONITORING INSTRUhENTATION l

3.8.9 The acr.ident monitoring instrumentatien 4.8.9 Each accident monitoring instrumentation channels shown in Table 3.8.9-1 shall be channel shall be demonstrated OPERABLE by OPERABLE.

performance of the CHANNEL CHECK and Instrument -

CHANNEL CALIBRATION operations'at the l

APPLICABILU1:

MODES 1, 2, 3 and 7-frequencies shown in Table 4.8.9-1.

ACIIDH:

a.

With the number of OPERABLE accident monitoring instrument channels less than the Required Number of Channels shown in Table 3.8.9-1, (Col. 2), restore the Inoperable channel (s) to OPERABLE status within 7 days, or be in at least MODE 4 tet thin the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

b.

With the number of OPERABLE accident moattoring instrumentation channels less-than the Minimum Operable Channels requirements of. Table.3.8.9-1, (Col. 3),. restore the Inoperable channel (s) to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, or be~in at least MODE 4 within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

c.

The provisions of Specifications 3.0.4 are not appilcable.

19140 184 TSC 90-02 c.

_.a

=

r f

J 1.

'2.

'3.~.

TOTAL NO.-

REQUIRCD NO._

JMINIMUM OPERA 8LE" '

~

. INSTRUMENT (PARAMETER) 0F CHANNELS OF CHANNELS

' CHANNELS

~ ~ ~

l

1. Containment Pressure (Hide Range) 2

.7' 1

2. Reactor Coolant Outlet Temperature - THOT (Hide Range) 4-(1/ loop) 2 1

3. Heactor Coolant Inlet Temperature - TCOLD (Hide Range) 4-(1/ loop) 2 1

4. Reactor Coolant Pressure (Hide Range) 2 2

1

5. Steam Line Pressur9 12-(3/SG) 2/SG 1/SG

6. Pressurizer Water Level 3

2-1

7. Steam Generator Water Level (Narrow Range) 12-(3/SG) 2/SG 1/SG 4

l

8. Steam Generator Mater Level (Hide Range) 4-(1/SG) 1/SG 2-l

9. Refueling Water Storage Tank Level 2

2 1

l

10. Auxillary Feedwater Flow Rate 4-(1/SG) 1/SG 2

l l

11 Reactor Coolant System Subcooling Margin 2

2 1

r Accident Monitoring Instrumentation Table 3.8.9- ?

19140 192a TSC 90-02' t-

~.

e 1.

.2.

?3h TOTAL NO.'

REQUIRED NO.

MINIMUM 0PERA8tEG

~

INSTRUMENT (PARAMETER)

QF CHANNELS OF CHANNELS CHADELS

12. PZR PORV Position Indicator

• 1/ valve 1/ valve *

~ / valve *'

1

13. PZR PORV Block Valve Position Indicator

• 1/ valve 1/ valve

• 1/ valve *

14. PZR Safety Valve Posttion Indicator **(Primary: Temperature 1/valvc 1/ valve Detectors)-

3-(1/ valve)

15. PZR Safety Valve Position Indicator (Backup:

Acoustic Monitors) 3-(1/ valve) 1/ valve N/A-

16. Core Exit Thermocouples 2****

2***

1***

17. Containment Water Level (Narrow-Range) 2 1#-

1#

18. Containment Water Level (Hide Range) 2 2

1

19. Reactor Vessel Hater Level (Hide Range)

[At least one RCP running]

2 2

1

20. Reactor Vessel Water Level (Narrow Range)

[All RCPs stopped]

2 2

1

21. Condensate Storage Tank Level 2

2 1

• Not required if the PZR PORV Block Valve is closed and de-energized in accordance with specification 3.3.1.F.

~

• • Direct indication of PZR Safety Valve Position - NUREG 0578, Ite a.

• • • An OPERABLE channel consists of at least 2 core exit thermocoupi-are quadrant.

• • • • 1 Channel consists of 32 thermocouples and the other consists of

. hermocouples.

1. Operation may continue up to 30 days with less than the REQUIRED NO. OF CHANNELS or.less than the MINIMUM OPERABLE CHANNELS.

Accident Monitorina Instr - ntation (Continued)

Table 3.8.9-1 (Continued) 1914')/11880 192b TSC 90-02~

- a = --

+

.m

.w CHANNEL

_ CHANNEL

~

INSTRUMENT (PARAMETER)

_QiEQ_.

WhREQH.

[

l

'I. Containment Pressure (Wide Range)

M

R

2. Reactor Coolant Outlet Temperature ~- THOT (Hide Range)

M

-R w

3. Reactor Coolant Inlet Temperature

'TCOLD (Hide Range)

M R

1

4. Reactor Coolant Pressure (Hide Range).

M R-

5. Steam Line Pressure M

.R'

~

6. Pressurizer Water Level M

R-

7. Steam Generator Mater Level (Narrow Range)

M R

8. Steam Generator Water Level (Mide Range)

M R

9. Refueling Hater Storage Tank Level M

R

10. Auxillary Feedwater Flow Rate M

R 11 Reactor Coolant System Subcooling Margin M

R

12. PZR PORV Position Indicator

• M R

13. PZR PORV Block Valve Position Indicator

• M R

14. PZR Safety Valve Position Indicator **(Primary: Temperature Detectors)

M R

15. PZR Safety Valve Position Indicator **(Backup: Acoustic Monitors)

M R

16. Core Extt Thermocouples M

R

17. Containment Water Level (Narrow Range)

M R

18. Containment Water Level (Hide Range)

M R

19. Reactor Vessel Water Level (Wide Range)

M-R

20. Reactor Vessel Water Level (Narrow Range)

M R.

21. Condensate Storage Tank Level M

R Not required if the PZR PORV Block Valve is closed and de-energized in accordance with Specification 3.3.1;F.

l Direct Indication of PZR Safety Valve Position - NUREG 0578, Item 2.1.3.a.

Accident Monitorina Instrumentation Surveillance Reautrements Table 4.8.9-1 19140 192c-TSC 90-02, s

.~

6CCIDENT PARAMETEP, SENSOR INSTRUMEM1_NUMSib

1. Containment Pressure (Hide' Range)

PT-CS50, PT-CS51

2. Reactor Coolant Outlet Temperature - THOT (Hide Range)

TE-413A-, TE-423A, TE-433A, TE-443A l

3. Reactor Coolant Inlet Temperature - TCOLD (Hide Range)

TE-4138 TE-4238. TE-4338, TE-4438

4. Restttr Coolant Pressure (Hide Range)

PT-403, PT-405 g

5. Steam Line Pressure PT-514, FT-515, PT-516, PT-524, PT-525, l

PT-526. PT-534, PT-535', PT-536, PT-544, PT-535, PT-546

6. Fressurizcr Water Level LT-459, LT-460, LT-461

7. Steam Generitor Water Level (Narrow Range)

LT-517, LT-518, LT-519,.LT-527, LT-528, LT-529, LT-537, Lt-538,.LT-539, LT-547, LT-548, LT-549

8. Steam Generator Water Level (Hide Range)

LT-501, LT-502, LT-503 LT-504-

9. Refueling Hater Storage Tank Level LT-920, LT-921

10. Auxiliary Feedwater Flow Rate FT-FH02, FT-FH03, FT-FH04, FT-FH25

11. Reactor Coolant System Subcooling Margin RCC01A, RCC01B Pressure Inputs: PT-403, PT-405 Temperature Inputs:

T-0001 thru T-0065

12. PZR PORV Position Indicator Valve Limit Switches

13. PZR PORV Block Valve Position Indicator Valve Limit Switches

14. PZR Safety Valve Position Indicator (Primary: Temperature Detectors)

TE-464 TE-465 TE-466 Accident Mont toring Instrument Numtutri Table 4.8.9-2 19140 192d-TSC 90-02

'w

.=

-i 3w ACCIDENT PARAMEIER-SENSOR INSTRUMENT NUMBER

~

J

15. PZR Safety Valve Position Indicator (Backup: Acoustic Monitors)

Acoustic Monitors LLO25, LLO26, LLO27

16. Core Exit Thermocouples T-0001'thru T-0065

17. Containment Water Level-(Narrow Range)

LT-CS48, LT-CS49

18. Containment Water Level (Hide Range)

LT-CS46, LT-CS47

19. Reactor Vessel Hater level (Hide Range)

LT-1321, LT-1322

20. Reactor Vesse! Water Leve? (Narrow Range)

LT-1311, LT-1312 j

g

21. Condensate Storage Tank Level LT-CD94, LT-CD139 i

i n

Accident Monitoring Instrument N W rs Table 4.8.9-2-(contlnued) 19140 192e TSC 90 4 a, _..

,, m.

~

vr :

r Bases 3.8 and 4.8 (Continued)

The OPERABILITY of the Service Water System ensures that sufficient cooling capacity is available for continued operation of safety related equipment during normal and' accident conditions.

In the spilt discharge' header OPERATING mcde (Unit 1 & 2 with separate discharge headers), two 0PERATING pumps and one standby are. required for-

~

normal operation.

In the cross-tied discharge header OPERATING mode (Unit 1 & 2 with common discharge. header),

three pumps have sufficient capacity for normal operation, but five pumps being OPERABLE with only one pump from.

either unit satisfying this requirement for the other' unit, are required to provide sufficient redundancy. 'In an' accident er shutdown mode, only one pump per unit is required.

For the Service Water System, the OPERABILITY requirements include consideration of the standby AC & DC power supplies so that a single failure of the "0" diesel will not cause a common mode failure In the Service Mater System; "0" diesel powers the emergency busses for Service Mater pumps 1A and 2A.

A Hydrogen Recombiner System is installed to remove the hydrogen and oxygen gases that accumulate in the containment atmosphere following a noss-of-coolant accident.

(7) The containment Hydrogen Monitoring System is used to determine the effectiveness of this system.

The CHANNEL CALIBRATION of the hydrogen monitors requires disassembly and electronic testing, adjustment and j

reassembly of the instrument, therefore the REFUELING CYCLE frequency is deemed adequate.

The instrumentation, equipment, and procedures for the tests which are required on the ventilation filter system will generally conform to the recommendations of ANSI M510-1975.

The OPERABILITY of the accident monitoring instrumencition ensures that sufficient information is available on selected plant parameters to monitor and assess these variables during and following an accident. This capability

[

is consistent with the recommendations of Regulatory Guide 1.97 Rev. 2, " Instrumentation for Light-Nater-Cooled Nuclear Power Plants to Assess. Plant Conditions During and-Following an Accident," December 1980, NUREG-0578..

l "THI-2' Lessons Learned Task Force Status Report and Short-Term Recommendations", July 1979 and NRC Generic Letter 83-37 "NUREG-0737 Technical Specifications," November 1, 1983.

i l

l (1) FSAR Chapter 9 (2) FSAR Section 6.2 (3) FSAR Section 6.2.3 (4) FSAR Section 14.3 l

(5) FSAR Section 9.3 (6) FSAR Section 9.6 & FSAR answer to question 9.1 (7) FSAR Section 6.8 19140 195 TSC 90-02 l

~

i Minimum Channels.

Applicable-.

Instrument Operable Acilon tlRdes 4.

Auxillary Building Ventilation and Miscellaneous Ventilation Stack A.

Gas Activity Monitor 1.

OR-0014 or 1

6 All 2.

1R-PR25 and 2R-PR25 1

6 All 3.

OR-PR188 Gas 1-

1 All-4.

1R-PR49E (Channel 5) 1 6

/11-5.

2R-PR49E (Channel 5) 1 6

All B.

Iodine Monitor 1.

1R-PR49C (Channel 3) 1 8

All 2.

2R-PR49C (Channel 3) 1 8

All C.

Particulate Monitor 1.

OR-PR18A Particulate' 1

6 All 1.

1R-PR49A (Channel 1) 1 8

All 2.

2R-PR49A (Channel 1) 1 8

All D.

Flow Rate Monitor 1.

ILP-084 1

9 All 2.

2LP-084 I

9 All 5.

Service _flulldino Ventilation A.

Gas Activity Monitor 1.

OR-PR22 1

8 All 8.

Particulate / Iodine Monitor 1.

OR-PR36 1

8 Ali 6.

Steam Generator Atmossneric Relief and Safety Valves 1.

1R-PR58 1'

4 1,2,3,7 2.

2R-PR58 1

4 1,2,3,7 3.

IR-PR59 1

4 1,2,3,7 4.

2R-PR59 1

4 1,2,3,7 5.

1R-PR60 1

4 1,2,3,7-6.

?R-PR60 1

4-1,2,3,7 7.

1R-PR61 1

4 1,2,3,7 8.

2R-PR61 1

4 1,2,3,7 Radioactive Gaseous Effluent Monitar Instrumentation (Continued)

Table 3.12-1 (Continued)

TSC 90 --'

3 M

's w

~

n n-

q Minimum 4-

. Channels Appilcable Instrument

'QDatAhlt Milan Modes 7.

Accident Monitoring A.

Containment 1.

1R-PR40G (Channel 7) 1

-10 1,2,3,4,7' 2.

2R-PR40G (Channel 7) 1-10 1.2,3,4,7.

3.

1R-PR40I (Channel 9) 1 10 1,2,3,4,7 4.

2R-PR40I (Channel 9)

'1 10 1,2,3,4,7 8.

Miscellaneous Vent Stack 1.

IR-PR49G (Channel 7) 1 4

1,2,3,4,7 2.

2R-PR49G (Channel 7) 1 4

-1,2,3,4,7 3.

1R-PR49I (Channel 9) 1 4

1,2,3,4,7 4.

2R-PR491 (Channel 9) 1 4

1,2,3,4,7 C.

Containment fuel Handling Area Moni?or 1.

IR-AR04A 1

11 6

When purging during-2.

1R-AR048 1

11 6

fuel handling

~

3.

2R-AR04A 1

11 6

operations l

4.

2R-AR048 1

11 6

i Radioactive Gaseous Effluent Monitor Instrumentation (Continued)

Table 3.12-1(Continued)

-TSC 90-02 19140 236b-

I

ACTION 4 - Hith the number of 0PERABLE channels less'than required by the Minimum Channels OPERABLE requirements, initiate an alternate method of monitoring the appropriate parameter (s) within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, and:

1.

either restore the Inoperable Channel (s) to OPERABLE status within 7 days of the event, or 2.

prepare and submit a Special Report to the Commission pursuant to Specification 6.6.3.8 within 14 days-following the event outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.

ACTION 5 - With the number of channels OPERABLE less than the minimum number required, the cuntents of the tank may be released to the environment provided that prior to initiating the release:

1.

At least two independent samples of the tank's content are analy::ed, and 2.

At least two technically quallfled members of the facility staff Independeutly verify the release rate calculations and discharge flow' path valving; otherwise suspend release of radioactive effluents via this pathway.

ACTION 6 - Hith the number of channels OPERABLE less than the minimum number required, effluent releases via this pathway may continue for up to 30 days provided grab samples are taken at least once per shift and these samples are analyzed for gross activity within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

ACTION 7 - Hith the number of channels OPERABLE less than the minimum number required, and no redundant monitor OPERABLE in this flow path, immediately suspend PURGING of radioactive effluents via this pathway.

ACTION 8 - Hith the number of channels OPERABLE less than the minimum number required, effluent releases via this pathway may continue for up to 30 days provided samples are continuously collected with auxillary sampilng equipment as required in Table 4.12.1.

ACTION 9 - Hith the number of OPERABLE channels less than the minimum number required, effluent releases via this pathway may continue provided the flow rate is estimated at least once per shift while release is in progress.

ACTION 10 - Hith the number of OPERABLE channels less than the minimum number required, restore the inoperable monitor i

to OPERABLE status within 30 days or establish an alternate means of monitorliq the parameter.

ACTION 11 - With the number of OPERABLE channels less than the minimum number required, suspend vent and purge operations and close each vent and purge valve providing dire'ct access from the containment atmosphere to the outside atmosphere or suspend the movement of nuclear fuel and reactor components in the vicinity of R

the reactor, refueling cavity, and transfer canal (containment side).

ACTION 12 - With the number of OPERABLE thannels less than the minimum number required, effluent releases via this pathway may continue provided the effluent flow is being accounted for in the total plant effluent.

Radioactive Gaseous Effluent Monitor Instrumentation (Continued) l Table Notation l

Table 3.12-1 (Continued) 19140 237 TSC 90-02

~,

Ninimum Channels l Applicable-Instrument Operattle :

Actton #

Modes-7 1.

Area Monitors A.

' Fuel Storage Pool. Area i

1.

OR-0005 24

.All 2.

OR-AR03 1

21 During Fuel Handling ~

Operations or Crane Operation In or near SFP.

3.

OR-AR13 1

21 During Fuel Handling Operations-B.

Containment Purge Isolation 1.

IR-AR04A 1

27.

6-When purging during i

2.

1R-AR04B 1

22 6

fuel handling 4

3.

2R-AR04A 1

22 6

operations 4.

2R-AR04B 1

22 6

l l

C.

Containment Area (High Range) 1.

IR-AR02 1

31 1,2,3,4,7 1

2.

2R-AR02

-1 31 1,2,3,4,7 3.

IR-AR03 1-31 1,2,3,4,7 4.

2R-AR03 1-31 1,2,3,4,7 D.

Control Room 1.

OR-0001 1

24 All E.

Technical Support Center 1.

Portable Area Monitor 1

24 All RADIATION MONITORING INSTRUMENTATION

' TABLE 3.14-1

.TSC 90-02 19140 251 m.-

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Action 27:

Hith the number of-channels OPERABLE"less than.the minimum number required. effluent-vla this' pathway may continue provided the.' gross radioactivity level (beta / gamma or isotopic)'is: determined H

at'least once per day.

If the inoperable channel is not~ returned to OPERABLE status within 30 days L -

conduct a Station Review to determine a plan'of action to restore the channel to~ operability.

Action 28i Hith the number of channels OPERABLE less than the minimum number required, comply with'the survelliance requirements 4.3.3.A.'2 and 4.3.3.B.

Actton'30:

W1th the number of channels OPERABLE 1ess'than the minimum number required,'inttlate an alternate method (if feasible) of monitoring.the appropriate parameter (s) within;72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />,1 and :-

1)

Elther restore the inoperable Channel (s) to OPERABLE status within 7 days of the avent. or 2)

Conduct a Station Review within 14 days to determine a plan of action to restore the channes to OPERABLE status.

ACTION 31:

With the number of OPERABLE Channels less than required by the Minimum Channels OPERABLE'

. requirements, initiate an alternate method of monitoring the appropriate parameter (s) within 72:

hours, and:.

1.

either restore the inoperable Channel (s) to OPERABLE status within 7 days'of-the event..or 2.

prepare and submit a Special Report to the Commission pursuant to Specification 6.6.3.C @

14 days following the event outlining the action taken, the cause of the inoperability &ad the plans-and schedule for restoring the system to OPERABLE' status.

t TABLE' NOTATION-(Continued)'

TABLE 3.14-1 (Continued)

TSC 90-02 1914G 252a

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6.2.

Plant Oogratino Procedurti 1.

Written procedures including applicable checkoff lists covering items listed below shall be prepared, implemented, and maintained:

A.

Normal startup, operation, and shutdown of the reactor and othe

ye'. ems and components involving nuclear safety of the facility.

B.

Refueling operations.

[

C.

Actions to be taken to correct specific and foreseen potential malfunctions of systems or components including responses to alarms, suspected primary system leaks, and abnormal reactivity changes.

D.

Emergency conditions involving potential or actual release of radioactivity

" Generating' Stations Emergency Plan" and station emergency and abnormal procedures.

E.

Instrumentation operation which could have an effect on the safety of the facility.

F.

Preventive and corrective maintenance operations which could hew an effect on the safety of the facility.

G.

Surveillance and testing requirements.

H.

Tests and experiments.

I.

Procedures to ensure safe shutdown of the plant.

J.

Station Security Plan and implementing proceoures.

K.

Fire Protection Program implementation.

L.

Post Accident Sampling Program which will ensure the capability to: obtain and analyze reactor. coolant and containment atmosphere samples, and collect and analyze or measure radioactive lodine and particulates in plant gaseous effluents under_ accident conditions. The program shaII include the following:

(1) Training of personnel, (11) Procedures for sampling and analysis.

(111) Provisions for maintenance of sampling and analysis equipment.

M.

Working hours of the Shift Engineer, Shift Control Room Engineer, Shift Foreman, and Nuclear Station Operator such that the heavy use of overtime is not routinely required.

19140 309 TSC 90-02

6.2 (Continued) 2.

Radiation control procedures shall be prepared, implemented and maintained.. These procedures shall specify.

I permissible radiation exposure limits and shall be consistent with the requirements of 10CFR 20.

The radiation protection program shall meet the requirements of 10CFR 20.

3.

Procedures for items identified in Specification 6.2.1 and any changes to such procedures shall be reviewed and approved by.the Operating Engineer and the Technical Staff Supervisor in the areas of operation and fuel' handling, and by the Maintenance Assistant Superintendent and Technical Staff Supervisor in the areas of plant maintenance, instrument maintenance, and plant inspection. Procedures for items identifled in Specification 6.2.2 and any changes to such procedures shall be reviewed and approved by the Technical Staff.

Supervisor and the Health Physics Supervisor / Chemistry Supervisor or designees. At least one person approving each of the above procedures shall hold a valid Senior Reactor Operator's license.

In addition.

l these procedures and changes thereto must have the authorization of the Station Manager or designee before being implemented.

l Hork and instruction type procedures which implement approved maintenance or modification procedures shall be approved and authorized by the Production Superintendent. The " Maintenance /Modifttation Procedure" utilized for safety related work shall be so approved only if procedures referenced in the " Maintenance / Modification Procedure" have been approved as required by 6.2.1.

Procedures which do not fall within the requirements of 6.2.1 or 6.2.2 may be approved by the Department Heads.

4.

Temporary changes to procedures identified in Specifications 6.2.1 and 6.2.2 above may be made provided:

A.

The intent of the original procedure is not altered.

j B.

The change is approved by.two members of the plant management staff, at least one of whom holds a Senior Reactor Operator's License on the unit affected.

C.

The change is documented, reviewed by the Onsite Review and Investigative Function and approved by the Station Manager or designee within 14 days of implementation.

5.

Drills of the emergency procedures described in Specification 6.2.1.D shall be conducted at the frequency specified in the Generating Station Emergency Plan. These drills will be plarined so that during the course of the year, communication links are tested and outside agencies are contacted.

l 6.3 action to be Taken in the Event of a Reportable Event in Plant Operation:

Any Reportable Event shall be promptly reported to th( 'Ilce President PHR Operations or his designated alternate.

I The incident shall be promptly reviewed pursuant to Sptctfication 6.1.7.B.2.(j) and a separate report for each reportable event shall be prepared in accordance with the requirements of 10CFR 50.73.

i 19140 310 TSC 90-02

6.6.3.B Special Reports (Continued) 7 IQP_LC -

SUBMITTAL DATE l.

Pressurizer PORV or Safety Valve challenggi Document the event in the Annual Report j.

(Future) k.

Steam generator tube inspection and/or Per surveillance requirements 4.3.1.B.5 plugging.

1.

Emergency Core Cooling System (ECCS)

Hithin 30 days as LER - include nozzle usage actuation and injection when RCS temp 2 factor per T.S. 3.3.2.F.3 350*F 1

t m.

Fire detector inoperability Within 30 days per T.S. 3.21.1.C k

n.

Fire pump system inoperability Within 30 days per T.S. 3.21.2.8 o.

Fire suppression system inoperability Nlthin 30 days per T.S. 3.21.2.C or 3.21.2.D L

p.

Sprinkler system inoperability Nithin 30 days per T.S. 3.21.3.C q.

Low pressure CO2 system inoperability Mithin 30 days per T.S. 3.21.4.C r.

Post Accident Radiation monitor Nithin 14 days per T.S. Tables 3.12-1 and 3.14-1 to the Inoperable greater than 7 days.

Regional Administrator of the NRC Regional Office.

SPECIAL REPORTS-Table 6.6-1 (Continued)'

19140 324 TSC 90 - -

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ATTACHMENT 6 EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATIONS For ease of review, the proposed changes have been broken down into 5 distinct categories, based on the changes made.

The No Significant' Hazards. Considerations has been broken down into separate assessments, to reflect these categories, and to appropriately address all the changes made.

The alphabetic designators following the categorical title.

designates the corresponding proposed change that is addressed _in the No Significant Hazards Considerations.

Commonwealth Edison has evaluated this proposed amendment and-determined that it involves no :: nificant' hazards considerations. According to 10 CFR 50.92(c), a proposed amendment to an operating license-involves no significant hazards considerations if operation of the facility in accordance with the proposed amendment would not:

-1.

Involve a_significant increase in the probability or consequences of an accident previously evaluated; or 2.

Create _the possibility of-a new or different kind of accident from any accident previously evaluated; or 3.

Involve a significant reduction in a margin of safety.

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ADMINISTRATIVE CHANGES (A) l The following. changes have been categorized as administrative in nature, tThese changes do not involve a change in intent, requirement, or regulation.

As such, this No Significant Hazards Considerations will address these items:

}

Page number changes and additions, 1

l=

3 Title changes, relocations, and additions, Location and wording of notes associated with the PORVs and PORV Block Valves, 1

-Clarified proper mode'to eliminate ambiguity Changer in terminology, Generation of new table to be used for reference purposes, Addition.of notes clarifying current exemptions to performing periodic surveillances, Changes providing reference to base documents and revision

numbers, Deletion of surveillance frequency notation addressed appropriately elsewhere in the Technical Specifications, Reformating of pages, 1

~

Equipment identification changes that are typographical in nature, and Inclusion'of reporting requirements into the administrative section of the Technical Specifications, required to be performed under an action statement, j

l The proposed changes do not result in a significant increase in the probability or consequences of accidents previously evaluated.

The proposed changes to the Technical Specification listed above have been i

determined to be administrative in nature.

The proposed amendment to the Technical Specifications does not change or alter any current operator actions, or requirements for the mitigation of any_previously evaluated accident.

These changes have no impact on assumed margins l4 or actions'during an evaluated accident.

Plant response to previously b

evaluated accidents will not be altered as a result of these changes.

As such, the probability and consequences associated with evaluated accidents have remained unchanged, t

The proposed changes do not create the possibility of a new o,-

L different kind of accident from any accident previously evaluated.

These proposed changes will not impose or result in plant operation that differ from any existing requirements.

No new equipment is being introduced as a result of these changes. As such, the possibility of a new or different kind of accident from any accident previously evaluated will not occur as a result of these changes.

The proposed changes do not involve a significant reduction in a margin of safety.

These changes do not alter the manner in which equipment required for safe operation of the plant is operated.

There are no setpoint, or operational limitations being altered or changed as a result of these revisions.

As such, these changes have no significant effect on the margin of safety.

19140 (4s)

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h ADDITION OR CHANGES OF INSTRUMENTATION-(B)

The_following changes involve the addition of instrumentation as a result of the guidance given in, Generic Letter 83-37 regarding "NUREG 0737 Technical-Specifications". As such, this No Significant Hazards Considerations will address the following items:

1 Containment Pressure (Hide Range) changing from narrow range to wide range.

In addition, the total number of instrument channels has been changed from 4 to 2 to address the appropriate number of qualified instruments, Changing of the Pressurizer Safety Valve Position Indicator to include both a Technical Specification for the primary and the backup indicators, Revision of the RCS Subcooling Monitor, to delete previously L

approved manual calculations in Iteu of utilization of installed instrumentation, and

+

The addition of the following instrumentation into the Technical Specifications:

L Core Exit Thermocouples, l

Containment Hater Level Hide Range, Containment Hater Level Narrow Range, l-Reactor Coolant Inventory Tracking System, and Condensate Storage Tank Level, The-proposed change does not result in'a significant increase in the probability or consequences of accidents previously evaluated.

The purpose of the Accident Monitoring-Instrumentation is to provide sufficient information to perform required manual actions, and to monitor and assess plant status and behavior during and following an accident.

The instrumentation referenced in the' Technical Specifications to perform these functions must'be capable of providing this_information under assumed accident conditions.

Therefore, there must both be an adequate number of parameters monitored, and these parameters must be monitored by appropriately qualified instrumentation.

The probability for an evaluated accident is independent of the number and type of instruments designated for monitoring purposes.

The probability for an accident is linked to the precursor events leading to an accident.

None of the' accident monitoring instrument changes addressed here are-linked in any fashion to these precursor events.

l i-L 19140 (u)

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s) a As such, the probability for previously evaluated events has remained unchanged.

The consenuences of evaluated events are lessened by assuring that the minimum number of parameters required to perform a*

required manual actions, and monitor and assess plant. status following-an accident are available.

a L

The-proposed changes do not create the possibility of a.new or l:

different kind of accident from any accident previously evaluated..

The proposed changes will not approve or. result in plant operation that:could create a new or different kind of accident.

The addition-of these instruments to'the technical specifications will provide; assurance in regards to their availability, through the establishment l

of Limiting Condicions for. Operation, Remedial Actions, and y

appropriate Surveillance Requirements.

The accident monitoring i

instrumentation is directly tied to accident mitigation, through the accomplishment of manual actions and assessment.

Through proper decision making the possibility of different type of accident or L

accident conditions propagating in a differing manner will be j

L prevented.

i The proposed change does not involve a significant reduction in a

'l margin of safety.

The inclusion of these instruments into the i

Technical-Specification will provide assurance that an adequate number l

of the appropriate. instruments are available to monitor required plant i

parameters in the event of an accident.

By referencing instrumentation that is appropriately qualified for the expectod-L-

condition, assurance is gained that these parameters will be available l

if required.

By reducing the total number of channels for the-Containment Hide Range Pressure Instruments from 4 to 2 the appropriately qualified instruments are referenced.

However, the margin of safety has remained unchanged by virtue of the required number of channels remaining the same.

For the instrumentation being added, these changes will enhance the safety of the plant through the 1

establishment of minimum acceptable levels of performance and-availability for these instruments in the Technical Specifications.

Through the establishment of.these minimum levels, information L

required for appropriate decision-making during and following an 0

accident can be assured.

Surveillance requirements are being l

specified for the purpose of determining operability.

The remedial

-actions associated with the added Specifications will require the U

plant to be placed into a shutdown condition, after providing an appropriate time frame to restore the inoperable instrument (s) to operable status.

As such, these proposed changes to the Technical Specifications will result in an increase in the current margin of safety.

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19140 (4s)

ACTION AND PROGRAMATIC CHANGES (C) t The following assessment involves the revision of the action statements associated with the accident radiation monitoring instruments.

This assessment also addresses the addition of programatic' control to assure these monitoring capabilities.

As such, this No Significant Hazards Considerations will address the following items:

' Addition of action statement 4-regarding the actions associated l

with the Steam Generator Atmospheric Relief and Safety Valve Radiation Monitors, and the Vent Stack Noble Gas Radiation

Monitors, Addition of action statement 31'regar' ding the actions-associated-with'the Containment High Range Radiation Monitors, and

- The addition of the requireh.?nt to have Post Accident sampling-programs. capable of' obtaining and analyzing reactor coolant and containment atmosphere, and to collect and analyze or measure radioactive lodines and particulates in plant gaseous effluents samples under accident conditions, i

The proposed changes do not result in a significant increase in the L

probability or consequences of accidents previously evaluated.

The L

radiation monitor action statement changes and programs being added to the-administrative section of the, Technical Specifications of themselves have no impact on the probability of any events that are u

L inputs to evaluated accidents.

As such, the probability for evaluated I

events has remained the same..The changes made are providing assurance to the availability of these monitors or alternative methods-of providing comparable information'during and following an accident.

Through assuring the availability of these parameters and programs, the consequences of evaluated accidents remain the same.

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

The proposed changes address the revisions to programs and action e

statements.

There is no new equipment being proposed for installation l

that could present the possibility for a new or different type of accident, as a result of these changes.

Plant operations will not be altered in any-fashion that could create the possibility for a new or different type of accident.

Through assuring the availability of l

these parameters during an accident, proper decision making can be I

assured.

As such, these changes will not create the possibility for a new or different type of accident.

In addition, the potential for accident conditions propagating in a differing manner will be l

prevented.

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19140 (u)

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The proposed change does not involve a significant reduction in a margin'of safety.

The proposed actions will result in the establishment of alternate methods of monitoring the effluent release paths for the Steam Generator Atmospheric Relief and: Safety. Valves, and Vent Stack within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> instead of 30 days as addressed in the current Technical ~ Specifications.

The' reduction in these time frames will provide for an increase in the readiness to monitor these parameters in the unlikely event of an accident._ The administrativa changes being added to the Technical Specifications involve the addition of the requirements to have Post Accident. sampling programs capable of obtaining'and analyzing reactor coolant and containme,'t atmosphere, and to collect and analyze or measure. radioactive iod),:0s-and particulates in plant gaseous effluents samples under-accident conditions.

These programs include; the training of personnel, procedures for sampling and analysis, and provisions for maintenance

'of sampling.and analysis equipment.

These changes provide assurance r

that these capabilities and programs will be established and main +sined.

The results of these changes are viewed.as an improvement in t!.; plants overall ability to assess = accident conditions, and the potential for or quantification of releases.

As such, these changes do not result in a significant reduction in-the margin of safety.

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CHANGES TO_ REQUIRED NUMBER OF INSTRUMENTS (D)

The following assessment involves changes to the required number of instruments, and. parameters that must be monitored.

As such, this No Significant Han rds Considerations will address the following items:

1

- Changes.to the Required and Minimum Number of instruments for the Steam Generator Wide Range Level Indicators,

' Changes to the Required and Minimum Number of instruments for the Auxiliary Feedwater Flow Rate Monitoring Instruments, The deletion of the PZR PORV acoustic monitors from the Technical-Specification, and Revision of the PZR PORV Block Valve Position Indicator, increasing the Minimum Number of channels from 0 to 1.

The proposed changes do not result in a significant increase in the probability or consequences of a:cidents previously evaluated. - The above noted instruments provide monitoring capabilities of parameters important for the proper assessment of plant status.

These proposed changes are not linked to increasing the prolability of any accident.

These changes do not alter any of the assumed initial conditions for evaluated accidents at the Zion Station.

As such, the probability and consequences for an evaluated accident has remained unchanged.

In the l

case of the Steam Generator Hide Range level and the Auxiliary Feedwater Flow Monitoring Instruments, the required number of instruments have become more restrictive.

In addition, the PZR PORV Block Valve Position Indicator, the Minimum Number of Channels has also become more restrictive.

The proposed change will increase the Minimum Number of Channels required for the Steam Generator Hide Range Level and Auxiliary Feedwater Flow Monitoring instruments based on the more restricive limits imposed on the Required Number of Channels operable. In addition, the Steam Generator Narrow Range Level instruments provide redundant indication of heat sink status for both the Steam Generator Hide Range level instruments, and the Auxiliary _

Feedwater Flow Rate instruments. As such, there is redundancy available for monitoring these parameters.

The Steam Generator Narrow Range Level instruments have a required number of 2, and a minimum number of 1, and are powered from redundant safety related power supplies.

In this fashion, it can be shown that there is sufficient redundancy available to justify a 7 day action statement alone for the Steam Generator Hide Range Level and Auxiliary Feedwater Flow Rate instruments.

In reference to the PORV Position Indicator, the safety related limit switch position indicators will still be retained in the Technical Specifications.

19140 (si)

.g These indicators on their own are sufficient to provide va;ve status indication in the Control. Room.

In.this fashion, the consequences of previously evaluated accidents will remain unchanged.

Based on the diversity available for monitoring the parameter associated with the Steam Generator Hide Range Level and the Auxiliary Feedwater Flow Rate m

indicator, in addition to the safety related limit switches associated t

with the PZR PORV' Position indication, sufficient monitoring capability will.be maintained.

As such, the consequences of evaluated events will remain the same through the assurance that the minimum-number of parameters required to perform" required manual actions, and monitor and assess plant status following an accident are available.

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

The proposed changes will not impose or result in plant operation that will challenge the integrity of any fission product barriers.

Post accident monitoring instrumentation is directly. tied to accident r

mitigation in regards to assessment and decision making process.

The number and type of instruments specified in the Technical Specifications, of there own will not result in the possibility of a different type of accident.

All of these items as proposed, will

'still have adequate indication available through either safety related or diverse indication.

Through the assurance that the required pcrameters necessary for appropriate decision making during and following an accident are available, it can be determined that the possibility of a different type of accident or an accident conditions i

propagating in a differing manner will be prevented.

L The proposed change does not involve a significant reduction in a i

margin of safety.

These changes will enhance the-safety of the plant through the establishment of more restrictive minimum acceptable levels of performance and availability for the Steam G~ aerator Hide Range Level.and Auxiliary Feedwater Flow Rate instruments in the Technical Specifications.

The current Technical Specifications do not h.

invoke any required remedial actions until'at least 3 instruments l'

fail.

In the event of 3 failures,'a 7 day action statement would be entered.

In the event of all 4 instruments-falling, a 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> action statement time clock is entered.

The proposed changes would invoke a 7 day time clock based on a single failure and a 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> time clock would apply for two or more channel failures.

The Steam Generator Narrow Range Level instruments provide redundant indication of heat sink status for both the Steam Generator Hide Range Level instruments, and the Auxiliary Feedwater Flow Rate instruments.

As-such, there is redundancy available for monitoring these parameters.

The Steam l

Generator Narrow Range level instruments have a required number of 2, i

and a minimum number of 1.

In this fashion, it can be shown that there is sufficient redundancy available to justify a 7 day action statement alone for the Steam Generator Wide Range Level and Auxiliary feedwater Flow Rate instruments.

19140 (sa)

,7-,__.,_

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Based on the diversity 4.va11able for monitoring this parameter, it can be co,'cluded that suffic'.snt monitoring capability will be-1 maintained.- As such,' based on the redundancy in indication, and the l

lower threshold for action statement entry, it has been concluded that' I

the overall margin of safety has'been enhanced.

In regards to the PZR PORV Valve Position Indicators, valve position indication will be t

available fre.ii the safety related limit switches associated with these valves.

Based on the current Technical Specifications, operations

'could continue for a period of 7 days based on the inoperability of either a stem mounted limit switch, or an'acoust.;al monitor.

If both of these indicators were inoperable, operations could continue for a period of time not to exceed 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.. If, the-required number of channels or minimum operable channels'are not restored within these L

time frames, the unit will be placed in Mode 4 within the next 12 l

hours.

The proposed changes will allow operations to continue for a period of time not to exceed 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> based on the inoperability of the stem mounted limit switch alone.. If, the required number of channels (same as the minimum operable number) are not restored within-this time frame, the unit will be placed in-Mode 4 within the next 12 i

hours.

In the event that the acoustical monitor is inoperable, the actions required will remained the same. As sJch, it can be concluded that the restoration time frame for an inoperable limit-switch is the same in the proposed change, is the same as the current Technical Specifications assuming loss of the acoustical monitor.

Regarding the PZR PORV Block Valve Position Indicators, valve position indication will be available from the safety related limit. switches associated l

with these valves.

Based on the current Technical Specifications, operations could continue for a period of 7 days based on the inoperability of a stem mounted limit switch.

If the required number of channels is not restored within this time frame, the unit will be placed in Mode 4 within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

The proposed changes will allow operations to continue for a period of time not to exceed 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> based on the inoperability of the stem mounted limit switch.

If, the required number of channels (same as.the minimum operable L

i number) are not restored within this time frame, the> unit will be placed in Mode 4 within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

In this fashion, the margin of safety wt11 remain the same.

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. MODE OF APPLICABILITY CHANGE (E) l The following No Significant Hazards Considerations addresses the addition of Mode 7 as a required mode for the accident monitoring instrumentation.

The Mode of Applicability for Specification 3.8.9 Accident Monitoring 1

Instrumentation has been revised to include Mode 7.

centric Letter 83 >

was written to address the Modes of Applicability as defined in the Standardi7+d Technical Specifications.

Mode 7 in the Zion Station Technical Specifications is defined as;:less than or equal to 5% power with reactivity and temperature stated per the specific test.

This mode is synonymous with Mode 2, incorporating specified tests from the Special Test Exemptions in the Standard Technical Specifications.

This change will require the Accident Monitoring Instruments to be operable in Modes 1, 2, 3, and 7.

The proposed change does not result in a significant increase in the probability or consequences of accidents previously evaluated. Mode 7 at the Zion Station is Low Power Physics Testing.

Low Power Physic Testing involves operation at pcwer levels not to exceed 5% of rated thermal power.

In order to establish these conditions, the plant must transition Mode 3.

In transitioning Mode 3 the plant is in a condition where the accident monitoring instruments are required to be operable.

The i

probability for a previously evaluated accident has remained unchanged..

The addition of this mode has no effect on an events or conditions that

(

are precursors to any evaluated accidents.

By assuring the operability of these instruments, in all appropriate modes, the consequences of previously evaluated accidents will remain the same.

The proposed ch6 7e does not create the possibility of a new or different kind of accident from any accident veviously evaluated. The proposed change does not approve or resuli uie installation of new equipment.

The proposed ciianges will not inr ne or result in plant operation that will deviate from current pract'. es.

In order to establish Mode 7, the plant must first enter Mode 3.

The Accident Monitoring Instruments are currently required operable in Mde 3.

As such, this change does not result in operations different from current. As such, the proposed change will not create the possibility for a new or different kind of accident 1

from those previously evaluated.

The proposed change does not involve a significant reduction in a margin of safety.

This proposed change will result in the accident monitoring instruments specified in table 3.8.9, being required to be operable in Mode 7.

The inclusion of this mode provides assurance that these instruments would be maintained operable during low power physics testing. As s';cn, the margin of safety will not be significantly reduced.

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ATTACHMENT 7 (hv!RONMENTAL ASSESSMENT STATEMENT Zion Station has evaluated the proposed amendm@nt against the criteria for and identification of Itcensing and regulatory actions requiring environmental assessment in accordance with 10 CFR51.21.

The proposed changes do not individually ce cumulatively have a Significant impact on the human environment.

As such, this change has been evaluated against d.he criteria for a categorical exclusion.

It has been determined that the prooosed change meets the criteria for a categorical exclusion as provided for under 10 CFR51.21(c)(9).

This determination is based on; involving the issuance of an amendment to a license for a reactor issued pursuant to 10 CFR50, which changes both the requirements for and the survel lance requirements associated with comoonents located within the restricted area, do not involve any significant hazards considerations, there is no change in the amount or type of releases made offsite, and there is no increase in individual or cumulative occupational radiation exposure.

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ATTACHMENT 8 ZION STATION TYPE A VARIABLES This attachment provides a listing of the Type A variables identified in Zion Station's Regulatory Gulje 1.97 submittal.

The instrument numbers and Technical Specification locations are provided for cross referencing purposes.

SPECIFICATION VARIABLE INSTRUMENT TABLE 3.8.9-1 ITEM

1) RCS Hot Leg Temperature TE-413A 2

l (Hide Range)

TE-423A 2

TE-433A 2

TE-443A 2

2) RCS Cold Leg Temperature TE-4138 3

(Hide Range)

TE-4238 3

TE-433B 3

TE-4438 3

3) RCS Pressure-PT-403 4

(Hide Range)

PT-405 4

4) Steam Generator Level LT-501 8

(Wide Range)

LT-502 8

c LT-503

-8 l

LT-504 8

5) Steam Generatcr Level LT-517, 518, 519 7

l (Narro.s Range)

LT-527, 528, 529 7

i LT-537, 538, 539 7

l LT-547, 548, 549 7

6) Prettu' irer Level LT-459 6

LT-460 6

LT-461 6

7) Steam Line Pressure PT-515, 516 5

L PT-525, 526 5

PT-535, 536 5

PT-545, 546 5

8) Containment Pressure PT-CS50 1

PT-CS51 1

9) Refueling Water Storage LT-920 9

1 I

Tank Level (Note 1)

LT-921 9

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10) Containment Water Leve' LT-CS46 18 (Mide Range)

LT-CS47 18

11) Auxiliary Feedwater Flow FT-FW02 10 FT-FW03 10 FT-FW04 10 FT-FW25 10

12) Containment Arta Radiation RE-AR02 Note 3 RE-AR03 Note 3

13) Core Exit Temperature T-0001 thru 16 l

T-0065

14) RCS Subcooling PT-403, 405, 456, 11 457, 458, and T-0001 thru T-0065 4

15) Condensate Storage Tank LT-CD94 21 l

Level (Note 2)

LT-CD139 21 Note 1: As addressed in Regulatory Guide 1.97 submittal dated August 1, i

1986, these transmitters uo not met Environmental qualification.

These instruments will be upgraded during the Unit I and 2 Cycle 13 Refueling Outages.

c Note 2: As addressed in Regulatory Guide 1.97 substttal dated August 1, 1986, neither of these level indicators met the criteria for Type A variables.

Instrumentation upgrades will be performed during the Unit 1 and 2 Cycle 13 Refueling Outages.

o Note 3: Containment High Ringe Radiation Monitors addressed in Specification 3.14.1. Table 3.14-1 item 1.C.

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10) intainment Water Level LT-CS46 18 I

(Wide Range)

LT-CS47 18

]

11) Auxiliary Feedwater flow FT-FWO2 10 FT-FWO3 10 FT-FWO4 10 FT-FW25 10

)

12) Containment Area Radiation RE-AR02 Note 3 RE-AR03 Note 3

13) Core Erit Temperature T-0001 thru 16 l

T-0065

14) RCS Subcooling PT-403, 405 and 11 T-0001 thru T-0065

15) Condensate Storage Tank LT-C094 21 level (Note I)

LT-CD139 21 i

Note 1: As addrested in Regulatory Guide 1.97 submit'.tl dated August 1 l

1986, these transmitters do not met Environmental ge411fication.

These instruments will be upgraded during the Unit I and 2 Cycle 13 Refuel 199 Outages.

Note 2: As addressed in Regulatory Guide 1.97 submittal dated August 1, i

1986, neither of these ivvel indicators mit the criteria for Type A variables.

Instrumentation upgrades will be performed during the Unit 1 and 2 Cycle 13 Refueling Outages.

Note 3: Containment High Range Radiation Monitors addressed in Specifitation 3.14.1 Table 3.14-1 item 1.C.

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. o ATTACHMENT 9 POST ACCIDENT MONITORING INSTRUMENTATION

{

This attachment provides a listing of the Type B through E variables referenced in this change.

The instrument numbers and Technical Specification locations are provided for cross referencing purposes.

These instruments are consistent with those specified in Zion Station's Regulatory Guide 1.97 submittal.

TECHNICAL SPECIFICATION TABLE ITEM INSTRUMENT 3.8.9-1 ITEM i

1) PORY Position Indicator Direct Limit Switch 12 I

2) PORV Block Valve Position Direct limit Switch 13

3) Safety Valve Position TE-464 14 (Primary Detector)

TE-405 14 (Note 6)

TE a66 14

4) Safety Valve Position Acoustic Monitor LLO25 15 (Backup Detector)

Acoustic Monitor LLO26 15 Acoustic honitor LLO27 15 f

5) Containment Water Level LT-CS48 17 (Narrow Range)

LT-CS49

6) Reactor Vessel Water Level LT-1311 20 (Narrow Range)

LT-1312

7) Reactor Vessel Water Level LT-1321 19 (Hide Range)

LT-1. 322 19

8) Vent Stack Monitor RIA-PR49 channels 7, and 9 Note 4

9) Steam Generator Atmospheric RIM-PR58 Note 5 Relief and Safety Valve RIM-PR59 Note 5 Monitors RIM-PR60 Note 5 RIM-PR61 Note 5 Note 4: Vent Stack Monitors addressed in Specification 3.12.3, Table 3.12-1 items 4.A.5, and 7.B.

Note 5: Steam Generator Atmospheric Relief and Safety Valve Monitors addressed in Specification 3.12.3, Table 3.12-1 item 6.

Note 6: These transmitters do not meet the criteria for a Category 2 variable.

The date for completing modification of these instruments has not been determined yet.

The design and modification packages are currently being reviewed by the plant.

The schedule for the modifications will be submitted in August, 1990 in Zion Station's L

Regulatory Guide 1.97 compliance response.

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ATTACHMENT 10 UNCHANGED PARANETERS i

This attachment provides a listing of parameters that have remained unchanged.

These items are currently in conformance with the guidance given in Generic Letter 83-37.

A brief statement summarizing compilance to the design and qualification criteria established in Regulatory guide 1.97 is provided for each of these parameters.

These items have not,been changed, as such no justifications have been provided.

Reactor Coolant Outlet Temperature - Tao" (Hide Range) Table 3.8.9-1 item 2 has remained the same.

These inscruments are Type A Category I variables.

The Reactor Coolant Outlet Temperature - THOT (Hide Range) instruments referred to by this table meet Category 1 design and qualification criteria.

Reactor Coolant Inlet Temperature - TCOLD (Hide Range) Table 3.8.9-1 item 3 has remained the same.

These Instruments are Type A Category 1 variables.

The Reactor Coolant Outlet Temperature - TCOLD (Hide Range) instruments referred to by this table meet Category 1 design and qualification criteria.

Reactor Coolant Pressure (Hide Range) Table 3.8.9-1 item 4 has remained the same.

These instruments are Type A Category I variables.

The Reactor Coolant Pressure (Hide Range) instruments referred to by this table meet Category 1 design and qualification criteria.

Steam Line Pressure Table 3.8.9-1 Item 5 has remained the same.

These instruments are Type A Category I variables.

The Steam Line Pressure Instruments referred to by this table meet Category I design and qualification criteria.

Pressurizer Water level Table 3.8.9-1 item 6 has remained the same.

These instruments are Type A Category 1 variables.

The Pressurizer Water Level Instruments referred to by this table meet Category I design and qualification criteria.

Steam Generator Water Level (Narrow Range) Table 3.8.9-1 item 7 has i

remained the same.

This specification deviates from the required l

number of channels specified in Generic Letter 83-37.

In Generic Letter 83-37 the required number of channels is 1, while the current Technical Specifications require 2.

These instruments are Type A Category I variables.

Regulatory Guide 1.9) requires Category 1 instrumentation to be redundant.

In order to assure redundancy is maintained. 2 Steam Generator Narrow Range Water Level Instruments are required.

The Steam Generator Narrow Range Water Level Instruments referred to by this table meet Category I design and qualification criteria.

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'A.+.=

Refueling Nater Storage Tank level Table 3.8.9-1 ties 9 has remained the same.

These instruments are Type A Catepory I variables.

The Refueling Water Storage Tank Level instrume.its referred to by this table meet Category I criteria, with the exception of environmental qualifications.

This exception was discussed in Zion Station's Regulatory Guide 1.97 submittal.

Modifications to upgrade these instruments are being processed at this time.

The commitment made to upgrade these instruments to Regulatory Guide 1.97 revision 2 qualifications was summarized in Attachment 3 of this submittal.

The modifications to upgrade these instruments will be completed for Unit I and 2 during their cycle 13 Refueling Outages. After completion of these modifications, the Refueling Water Storage Tank Level Indicators will meet Category I design and qualification criteria.

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