SNRC-1013, Forwards Response to Generic Ltr 83-28, Required Actions Based on Generic Implications of Salem ATWS Events. Encl Addresses Each NRC Position & Describes Actions Taken by Util in Response

Forwards Response to Generic Ltr 83-28, Required Actions Based on Generic Implications of Salem ATWS Events. Encl Addresses Each NRC Position & Describes Actions Taken by Util in Response

ML20087H275
Person / Time
Site:	Shoreham File:Long Island Lighting Company icon.png
Issue date:	03/09/1984
From:	Mccaffrey B LONG ISLAND LIGHTING CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
References
GL-83-28, SNRC-1013, NUDOCS 8403200302
Download: ML20087H275 (50)

Text

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! LONG ISLAND LIGHTING COMPANY

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• WADING RIVER, N.Y.11792 March 9, 1984 SNFC-1013 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Cmmission Washington, D.C. 20555 Response to Generic Letter 83-28 Shoreham Nuclear Power Station - Unit 1 Docket No. 50-322 Reference 1: LIICO letter, SNBC 960 (J. L. Smith) to the hPC (H. R. Denton), dated September 9, 1983.

Dear Mr. Denton:

The attac.hed report is in response to Generic Ietter 83-28, " Required Actions Based on Generic Implications of Salem ATWS Events". '1his report addresses each NRC position and provides a description of the actions taken by LIICO in response. hhere a program is under developtent an estimated catuitment date has been provided.

If any additional information is required, please contact this office.

Very truly yours, f /("7/'f 49 B. R. McCaf Manager, Nuclear Canpliance and Safety PJT:ck ec: C. Petrone All Parties Listed in Attachment I Attachment l

1 8403200302 840309 .

PDR ADOCK 05000322

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ATTACHMH7T I Bernard M. Bordenick, Esq. Herbert H. Brown, Esq.

David A. Repka, Esq. Lawrence Coe Ianpher, Esq.

U.S. Nuclear Regulatory Karla J. Ietsche, Esq.

Ccxtmission Kirkpatrick, Iockhart, Hill Washington, D.C. 20555 Christopher & Phillips 8th Floor Mr. James Dougherty 1900 M Street, N.W.

3045 Ibrter Street Washington, D.C. 20008 Matthew J. Kelly, Esq.

State of New York Ben Wiles, Esq. Department of Public Service Assistant Counsel to the Governor Three 9tpire State Plaza Executive Chamber Albany, NY 12223 State Capitol Albany, NY 12224 Gerald C. Crotty, Esq.

Counsel to the Governor Mr. Marc W. Goldsmith Executive Chamber Energy Research Group State Capitol 4001 Totten Pond Road Albany, NY 12224 Waltham, Massachusetts 02154 .

MHB Technical Associates Martin Bradley Ashare 1723 Hamilton Avenue Suffolk County Attorney Suite K H. Lee Dennison Building San Jose, CA_ 95125 Veterans Memorial Highway ,

Hauppauge, NY 11788 Stephen B. Latham, Esq.

hxney, Latham & Shea Ralph Shapiro, Esq. 33 West Second Street Canmer and Shapiro, P.C. P.O. Box 398 9 East 40th Street Riverhead, NY 11901 New York, NY 10016 I

LIICO Response to Generic letter 83-28 Iten 1.1 POST-TRIP REVIEW (PROGRAM DESCRIPTION AND PROGDURE)

NBC Position - Licensees and applicants shall describe their program for ensuring that unscheduled reactor shutd w ns are analyzed and that a detennination is made that the plant can be restarted safely.

LIICO Response LIICO has developed a syste:natic post-trip review program that will be implenented prior to the operation of the shoreham Nuclear Power Station (SNPS) . The program ensures that any unscheduled reactor shutdown is analyzed to determine if the plant can be restarted safely. The controlling procedure for this program is Station Procedure 21.003.01, " Operating Reports." 'Ihis procedure is attached as Appendix A.

'Ihe following is an itanized description of the SNPS post-trip review program in response to NRC Generic Ietter 83-28.

Item 1.1.1 NRC Recuest - Describe the criteria for determining the acceptability of restart.

LIICO Response LIICO is participating in the BWR Owner's Group (BWROG) Salen A'1HS, Generic Issues ccmnittee effort to develop a generic description of the criteria for determining the acceptability of plant restart after an unscheduled reactor shutdown. Although LIICO expects to endorse the ccmnittee's position, we will subnit a formal response to the NRC within sixty (60) days after the receipt of the finalized BWBOG response-to allow time for a Shoreham-specific evaluation.

Item 1.1.2 NRC Request - Describe the responsibilities and authorities of personnel who will perfona the review and analysis of these events (unscheduled reactor shutdowns) .

LIICO Response The Watch Engineer is responsible to ensure that the plant is operated safely and in accordance with the requirenents of the facility Operating License, Technical Specifications, and approved operating procedures. 'Ihe Watch Engineer has the responsibility and authority to direct a shutdown of the' reactor whenever he determines that the safe operation of the plant is in innediate jeopardy or when operating parameters exceed reactor protection set points and automatic action does not occur. The Watch Engineer reports to the Operating Engineer and is responsible to ensure data is collected and an analysis of suci. data performed to detennine the cause of any unscheduled shutdown. He analyses items such as equignent malfunctions, procedure inadequacies and operating errors to determine the cause of the scram. He does not reuams-d a reactor restart unless the cause of the scram is fully understood.

2 SNRC-1013 The Shift Technical Advisor is responsible for advising the Watch Engineer regarding reactor core damage prevention or mitigation, during plant accident or transient conditions. He is responsible to ensure the required data is collected and an analysis is done of the data to determine the cause of the event.

The Operating Engineer is responsible for directing day to day operation of the Shoreham Nuclear Power Station unit including startup, operation, and shutdcwn of all equignent in accordance with approved operating procedures, Technical Specifications and regulatory requirments. In addition, the Operating Engineer has the authority to order the shutdown of the reactor whenever he determines that the safe operation of the plant might be jeopardized or if it appears that operating parameters will exceed the reactor protection setpoints. The Operating Engineer ensures that the unscheduled shutdown has been analyzed and the cause determined and corrected prior to authorizing a restart of the reactor.

Item 1.1.3 NRC Request - Describe the necessary qualifications and training for the responsible personnel.

LIILO Response LIICO is participating in the BWR Owner's Group (BWROG) Salem AWS Generic Issues Cmmittee effort to develop a generic description of the necessary qualifications and training for the responsible personnel. Although LIICO expects to endorse the cxmmittee's position, we will subnit a formal responm to the NBC within sixty (60) days after the receipt of the finalized BWIOG response to allow time for a Shoreham-specific evaluation.

Item 1.1.4 NRC Request - Describe the sources of plant information necessary to conduct the review and analysis. The sources of information should include the measures and equignent that provide the necessary detail and type of information to reconstruct the event accurately and in sufficient detail for proper understanding. (See Item 1.2) .

LIIOD Response The primary sources of information used to evaluate unscheduled reactor shutdowns are the plant process cmputer and various chart recorders used to record specific plant parameters.

The process certputer has the capability of measuring ard sorting values of analog variables at various time intervals to provide a post trip log of historical data.

This data is autmatically printed following a reactor scram. The plant process

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cmputer and additional sources of plant information are discussed in the response to It s 1,2, " Post Trip Review - Data and Information Capability".

Item 1.1.5 NRC Request - Describe the methods and criteria for comparing the event information with known or expected plant behavior (e.g., that ;

safety-related equipnent operates as required by the Technical

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Specifications or other performance specifications related to the safety function).

I 3 SNRC-1013 j LIICO Response LIIf0 is participating in the BWROG Salem ATWS Generic Issues Camittee effort to develop generic methods and criteria to aid in the ocmparison of the event generated information with known or expected plant behavior. Although LIICO expects to endorse the cmmittee's position, we will submit a formal response to the NRC within

, sixty (60) days after the receipt of the finalized BWROG response to allow time for a Shoreham-specific evaluation.

Item 1.1.6 NRC Request - Describe the criteria for determining the need for independent assessment of an event (e.g., a case in which the cause of the event cannot be positively identified, a cmpetent group such as the Plant Operations Review Cmmittee, will be consulted prior to authorizing re-start) and guidelines on the preservation of physical evidence (both hardware and software) to support independent analysis of the event.

LIICO Response The Operating Engineer will implement SP 21.003.01 and request that the Review of Operations Cmmittee (ROC) be convened when the cause of the event cannot be positively identified. 'Ihe ROC will use the same information as the Watch Engineer, Shift 'Ibchnical Mvisor and Operations Engineer to make their determination. In addition, members of the Shoreham Plant Staff can enlist the expertise of other personnel in the Office of Nuclear or an outside organization to help in determining the cause of the event.

Item 1.1.7 NRC Request - Items 1.1.1 through 1.1.6 are considered to be the basis for the establishment of a syst m atic method to assess unscheduled reactor shutdowns. 'Ihe systmatic safety assessment procedures ocmpiled frm the above items, which are to be used -in-conducting the evaluation, should be in the report.

LIICO Response Station Procedure 21.003.01, " Operations Reports", contains the Scram Report and Scram Evaluation procedures. 'Ihese procedures provide a systematic method to ' assess any unscheduled reactor shutdown.

Item 1.2 IOST-TRIP REVIEW - DA'IA AND INFORMATICN CAPABILITY NRC Position - Licensees and applicants shall have or have planned a capability to record, recall and display data and information to -

permit diagnosing the causes of unscheduled reactor shutdowns prior to. restart and for as rtaining the proper functioning of safety-related equipnent.

Mequate data and information shall be provided to ' correctly. diagnose the cauce of unscheduled reactor shutdowns and the proper functioning of safety-related equipnent during these events using systematic ~

safety assessment procedures (Action 1.1) . The data and information-

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shall be displayed in aiform that permits ease of assimilation and analysis by persons trained in the use of systm atic safety assess-

. ment procedures.

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4 SNRC-1013 LILOO Response At SNPS, plant data and infonnation that may be used to diagnose the cause of an unscheduled reactor shutdown is contained in the plant process cmputer, the Dnergency Response Facility (ERF) computers, radiation monitoring cmputers-and on strip chart recorders. Due to the long engineering lead time required for the ERP camputer system, the function will be incorporated in two phases. Phase I will utilize an upgraded plant process cmputer to provide a limited Safety Parameter Display System SPDS and a special data logging function. Phase I will be available at fuel load. Phase II will be the permanent system. Phase II will be fully operational within six (6) nonths after the first refueling outage.

Item 1.2.1 Capability for Assessing Sequence of Events (On-Off Indications)

Its 1.2.1.1 NRC Request - Provide a brief description of the equipment.

LILOO Response The capability for assessing sequence of events (on-off indications) at Shoreham is provided mainly by the balance-of-plant (BOP) ard nuclear steam supply systs (NSSS) sequence of events logs in the plant process cmputer and supplemented by the digital parameters which provide inputs to the ERF Phase II cmputer system. 'Ihe BOP sequence of events log nonitors those digital points which lead directly to or directly cause a turbine trip. 'Ihe NSSS sequence of events log performs a similar function for digital points which directly lead to or cause a reactor scram. The ERP Phase II cmputer systs is briefly suntnarized in Section 1.2.3.

Its 1.2.1.2 NRC Request - Describe the monitored paraneters.

LIIOD Response Appendix B provides a cross-reference relating plant ; mess cmputer sequence of events cmputer points with associated plant annunciator alarm points. The' dis-tinction between NSSS and BOP points can be made by noting.the system number column.

Syst m 1C51 and 1C71 are NSSS points and the remainder are BOP points.

Item 1.2.1.3 NRC Request - Describe the time discrimination between events.

LILOO Resoonse The plant process cmputer sequence of events log provides a tutHnillisecond resolu-tion between events.

Item 1.2.1.4 NRC Request - Describe the fonnat- for displaying data ard .

information.

LIICO Response The plant process carnputer sequence of events logs are displayed on a printer in the main control romt. . Appendix C is a sample of the format of the sequence of events log.

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SNRC-1013 Item 1.2.1.5 NRC Request - Discuss the capability fcr retention of data and information.

LIILO Response Since the plant process emputer system provides hard copy outputs, the data and information can be retained indefinitely.

Item 1.2.1.6 NRC Request - Describe the power source (s) .

LIILO Response The plant process cmputer is normally powered frm non-Class 1E inverters (uninterruptible power) which is backed by diesel generator power.

Item 1.2.2 Capability for Assessing the Tine History of Analog Variables.

Item 1.2.2.1 NRC Request - Provide a brief description of the equignent (e.g.,

plant c:xtputer, dedicated cmputer, strip charts) .

LIIIJO Response The capability for assessing the time history of the analog variables which may be used to determine the cause of an unscheduled reactor shutdown and the functioning cf safety-related equipnent is provided by the analog parameters monitored by the crocess cmputer systs, the ERF Phase I portion of the process computer systs and by analog parameters recorded and displayed on various strip chart recorders located in the main control tran. Additionally, radiological and meteorological parameters are available via the cmputer-based radiological monitoring systs. Analog signals frm Category I (safety related) radiation unitors are also recorded on strip chart recorders in the main control rom.

s Item 1.2.2.2 NRC Request - Describe the parameters monitored, sampling rate, and basis for selecting parameters and sampling rate.

LIIID Response An analog parameter post-trip log is generated autmatically by the plant process cmputer upon detection of a unit trip. The log consists of ten (10) NSSS parame-ters and forty (40) DOP parameters which are accumulated for a period of five (5) minutes before and after a unit trip. Appendix D lists the NSSS and BOP parameters printed in the process cmputer post-trip log. To provide a limited SPDS/ERF capability during initial plant operation (Phase I), the process plant cartputer has been upgraded to include sme SPDS graphic display _ and Technical Support Center (TSC) post-trip logging capability. Appendix E contains a list of points which are available for this purpose. Analog signals sent to strip chart recorders are _. _

recorded continuously. 'Ihe'following analog parameters ara recorded in the control rom:

6 SNRC-1013

.i (a) Reactor level-(b)

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Reactor pressure (c) Suppression pool temperature (d) Signals from QA Category I radiation monitors (e)~ Reactor building differential pressure (f) _ Drywell pressure (g) Suppression pool level (h) Neutron flux- '

Radiolcgical and meteorological parameters are provided by the radiological monitoring system. During ERF Phase I, the data will be resident only in the radiological monitoring system emputers and will be available upon demand.

Appendix F provides a list of the monitored points.

Itsu 1.2.2.3 NRC Request - Describe the duration of the time history (minutes before trip and minutes after trip).

LIICO Response The strip chart recorders listed above record continuously.before and after the trip. The analog post-trip log function of the plant process cmputer provides output for the NSSS parameters at five (5) second intervals and output for the BOP parameters at fifteen (15) second intervals. . Data is provided for five ~(5) minutes before and after a unit trip in a tabular format. We ERF Phase I Cortputer provides three types of TSC logs upon denand. 'Ihey are:

1. 'Iwo hours pre-event data of 112 points at one minute intervals plus five minutes post-event data which includes:

a. 9 points'every second

b. 16 points every 15 seconds

c. 87 points every minute

2. 5 minutes pre-event to 5 minutes post-event data, consisting of:

a. 9 points every second b.- 16 points every-15 seconds

c. 87 points every minute 3.- Continuous TSC log-of 112 points per minute.

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1 7 SNBC-1013 The radiological nonitoring cmputer can provide historical data upon demand as follows:

1. Sixty (one second intervals) listings of instantaneous values of all points in one minute data blocks.

2. Sixty one-hour averages for each point based upon data gathered in 1 above.

3. Sixty forty-eight hour averages for each point based upon data gathered in 2 above.

1.2.2.4 NRC Request - Describe the format for displaying data including scale (readability) of time histories.

LIIfD Response Analog parameters monitored on one two, or three pen strip chart recorders having 4 inch chart paper (minimum) are displayed on paper which is scaled for the parameter being recorded. Additionally, a separate scale and pointer on the recorder shows the current value of the parameter being recorded.

Analog parameters nonitored by the ERF Phase I cmputer system can be displayed on CRT consoles as shown in Appendix G. The format for the ERF Phase I Special log is shown in Appendix H.

1.2.2.5 NRC Request - Describe the capability for retention of data, information and physical evidence (both hardware and software) .

LIICO Response Strip chart recorder output is available indefinitely as long as the chart rolls are stored in an accessible location. This is true also for printed outputs frm the plant process cmputer system (including ERF Phase I) . 'Ihe ERF Phase I cmputer systm utilizes periodic data transfers to magnetic tape drives for long-term data storage. 'Iherefore, the data is expected to be available indefinitely.

1.2.2.6 NRC Request - Describe the power source (s) .

LIICO Response The plant process cmputer system, including ERF Phase I, is powered by non-Class 1E inverters (uninterruptible power) and is backed by diesel generator power. 'Ihe radiation monitoring systm cmputers are also powered frm diesel backed, non-Class 1E power. Recorders for parameters a, b, c, d, e, f, and g listed in Section 1.2.2.2 above are powered frm 1E power. Recorders for parameter h are powered frm non-Class 1E sources. In the case of parameter c, loop power for the 1 foot tmper-ature sensors is frm RPS power, while the loop power for the 2 foot temperature sensors is frm Class 1E (Divisions I and II) power sources.

8' SNRC-1013 Item 1.2.3 NRC Request - Describe other data and information provided to assess the cause of unscheduled reactor shutdown.

LIIOD Response The ERF Phase II emputer system will replace the Phase I systs within six (6) months after the return frm the first refueling outage. The Phase II syst s is essentially independent of the Phase I system and the SPDS and special data logging functions of Phase 1 will not be affected during the startup of the Phase II systs.

The infonnation provided is preluninary and is being used for developrent of the ERF Phase II Systs per SNRC-863 (4/4/83) . '

The plant process cmputer will remain tha major source of detailed sequence of events information since most of these points will not be nonitored by the Phase II system. Contact (digital) inputs to the ERF Phase II ccuputer systs are scanned with a one-second resolution.

The ERF Phase II cmputer system digital point status may be displayed on cathode ray tube (CRT) terminals and printers in the technical support center (TSC) and emergency operations facility (EOF) and by CRT only in the control rom. The ERF post-trip log concept under developnent will consist of the entire data base taken at various frequencies and intervals. Printed outputs can be retained indefinitely, he ERF Phase II systm will be powered frm Non-Class 1E inverters (uninterruptible power).

We ERF Phase II system will implement full SPDS and analog trend information as described below. In addition, a data link will transfer certain data frm the radiation nonitoring system to the ERF computer to allow display of certain meteorological and radiological parameters.

These points in the input list were mlected using NUREG-0696 guidelines with Shoreham specific Regulatory Guide 1.97 BWR parameters e m prising the minimum data set. The scan class was selected based on the rate at which an' individual parameter could change so as not to lose information during transient conditions. We scan periods vary from 0.1 to 60 seconds.

The ERF Phase II cmputer system records continuously on a circular file and stores data for two hours prior to any reactor scram (manual or autmatic) and for up to two weeks following reactor scram. This feature is effective. for both analog and digital inputs. ne capability to record up to two weeks of additional post-event data is provided by utilizing disk drives for long-terr data storage. Data recording beyond two weeks is possible by transferring the data on disk to magnetic tape.

Further detailed information on the ERF Phase II Ccnputer System is outlined in the Shoreham FSAR,Section III.A.l.2. It should be noted that there are additional digital and analog parameters which are monitored by the plant process cmputer and additional analog recorders in the control rom which do not specifically provide sequence of events information but which do give information which could be useful in assisting in the determi:mtion of the cause of a reactor trip.

LIICO may modify the number of points monitored, listed setpoints, scan groups, formats, etc., based upon the results of'startup and operations experience, to inprove the usefulness of the runitored data. -

9. SNRC-1013 Its 1.2.4 NBC Request - Provide the schedule for any planned changes to existing data and information capability.

LIICO Response The plant process cmputer and ERF Phase I will be operational at fuel load. ERF Phase II will be fully operational within six (6) months after the first refueling outage. It should be noted that the Phase I and II systems are cmpletely indepen-dent but use sme equipnent in parallel.

Item 2.1 EQUIPMENT CLASSIFICATION AND VENDOR INTERFACE (REACIOR TRIP SYSTEM COMPONENTS).

Item 2.1.1 Equipnent Classification (Reactor Trip System Components) .

NRC Position - Licensees and applicants shall confirm that all cmponents whose functioning is required to trip the reactor are identified as safety-related on documents, procedures and information handling systems used in the plant to control safety-related activ-ities, including maintenance, work orders, and parts replacement.

LIICO Response LIICO has identified the Reactor Trip System (RTS) components that should be classified as safety related for SNPS. The resulting list of active cmponents of existing plant systems that function to implement a reactor scram is in the process of being incorporated into the SNPS Cmposite Cmponent List (CCL) . 'Ihe CCL is a listing of safety-related cxxuponents that is part of the plant information handling system. It is further discussed in the response to item 2.2.1.

In addition, LILOO is participating in the BWROG Salem A'IWS Generic Issues utmittee effort to develop a generic Reactor Trip Function List and provide several safety classification examples for cmponents with various cmplexity levels. LIICO expects to endorse the Ccxtmittee's position and we will subnit a formal response to the NRC within sixty (60) days after the receipt of the finalized BWROG response.

'Ihis response will indicate the status of any CCL modifications deemed necessary as a result of the Shoreham-specific evaluation of the BWBOG generic response.

Item 2.1.1 Vendor Interface (Reactor Trip System Ccxrponents) .

NBC Position - For these cmponents, licensees and applicants shall establish, implement and maintain a continuing program to ensure that vendor information is emplete, current and controlled throughout the life of the plant, and appropriately referenced or incorporated in plant instructions and procedures. Vendors of these cmponents should be contacted and an interface established. Where vendors can not be identified, have gone out of business, or will not supply the information, the licensee or applicant shall assure that sufficient i attention is paid to equipnent maintenance, replacment, and repair, to cmpensate for the lack of vendor backup, to assure reactor trip system reliability. The vendor interface program shall include periodic ccmuunication with "cndors to assure that all applicable information has been received. The program rhould use a system of

' 10 positive feedback with vendors for mailings containing technical l l information. This could be accmplished by licensee acknowledgement  !

j' for receipt of technical mailings. The program shall also define the i interface and Ovision of responsibilities among the licensees and

' the nuclear and non-nuclear divisions of their vendors that provide  !

service on reactor trip systen caponents to assure that requisite control of and applicable instructions for maintenance work are provided.

LIIID Response T.TTm is instituting a program to receive, control, review and utilize vendor technical information such as GE Service Information IArtters (SILS) and Service

j. Advisory Letters (SALS) as per the requirements of Corporate Procedure NED 2.07,

" Review of Vendor Technical Bulletins". The procedure _is expected to be inplemented by April,.1984. GE also has further responsibilities, purstant to-10 CFR 21, which require it to identify safety pr@ lens.

, LIIID also reviews relevant industry information and experience through programs involving participation in the Nuclear Plant Reliability Data System (NPRDS) and the Significant Event Evaluation and Information Network (SEE-IN), both of whicit are IMnaged in INPO.

LILOO is participating in the BWROG' Salem AINS Generic Issues Cmmittee effort to develop a program to assure that RTS Vendor information is cmplete and current.

Although LIICO expects to endorse the Cmmittee's position, we will subnit a formal response to the NRC approximately sixty (60) days after the receipt of the finalized

{ BWROG response to allow time for a Shoreham-specific evaluation.. '

In addition, LIITO is a member of the Nuclear Utility Task Action Cmmittee (NUTAC) which has beut formed on vendor interface. - An approved program is expected at the.

end of the first quarter of 1984. . The NUPAC program is further discussed in the response to item 2.2.2.

Iten 2.2 EQUIPMENT CIASSIFICATION AND VD3 DOR INTERFACE (PROGRAMS EUR ALL -

SAFETY-RETATED 001PONENIS)

Itan 2.2.1. Equipnent Classification (Programs For All Safety-Related Cwwwnts)

NRC POSITION - For equipnent classification, licensees and applicants shall describe their program for_' ensuring that all e :-wrts of safety-related systems necessary for.accxmplishing required safety-functions are identified as safety-related on h=mts, procedures, and information handling systems used in the plant to control' . .

safety-related activities, including maintenance, work orders and

. replacement parts.-

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Titan 2.2.1.l NRC Request-- Describe the' criteria for identifying u.=Wients as

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3-safety-related within ~ systems currently classified.asisafety-related.

i This shall not be . interpreted to ;requiire changes-in safety classi- 1

- fic tion at the systems level. '

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i Item 2.2.1.2 NRC Request - Provide a description of the infomation handhng system used to identify safety-related emponents (e.g., cmputerized equipnent list) and the methods used for its development and validation.

LIIOD Response (Itans 2.2.1.1 and 2.2.1.2 )

Caponents within structures or systes classified as safety related are designated as " safety.related" if they are necessary to assure

1) '1he integrity of the reactor coolant pressure boundary,

2) - 'Ihe capability to shutdown the reactor and maintain it in a safe. shutdown 3 condition, or

. 3) '1he capability to. prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to the guideline exposures of 10 CFR Part 100, Appendix A.

i Accordingly, two broad functional classifications; safety-related and non safety-related, have been established. 0A Category I is assigned to safety-related omponents. CA Category II'is assigned to nonsafety-related ccmponents.

Inherent in the designation of a cmponent's classification is a-recognition of the function performed by the cmponent. '1he General Design Criteria, Appendix A to 10 CFR Part 50, establishes the mininum requirments for the design of a nuclear power plant. '1he American Nuclear Society Standard ANS 22, " Nuclear-Safety Criteria for .

the Design of Stationary' Boiling Water Reactor Plants", was used in establishing the ;

classification of structures, systems and conponents for SNPS. An evaluation of the design bases of the Shoreham Nuclear Power Station as measured against the General' Design Criteria, is presented in Section 3.1 of the Shoreham FSAR. A sumary of he

, classification of the Shoreham structures, systems and umpents is presented in :

Section 3.2 of the Shorehm FSAR.

LIICO has a emprehensive program which ensures _ that' all safety-related cmponents .

are identified as' safety-related on documents,' procedures and information handling used in the plant to control. safety-related activities. ' '1he program reflects the

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' design' criteria presented in the Shoreham FSAR,. the T. Tim Quality Assurance Program in effect during construction and operation, and the LIIOD procedures controlling:

all engineering and operational activities at'Shoreham. T.TTm has performed a Plant'.

Configuration Review to ensure that the as-built configuration'of safety-related

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systems ocmform to. the camitments in the FSAR and licensing h=mts.- _'1he inple-mentation and the results of the SNPS Plant Configuration Review Program were recently reviewed and found acceptable by the NRC Office of-Inspection and Ehforce-ment, Region I, as am-mted in Inspection Report 50-322/83-38. In addition, T.TTm i ):a 'a program in place -to ensure that by the time of fuel load, or shortly . '

L thereafter, the configuration'of the plant will be accurately reflected by drawings?

[ that will be used by the station operations staff.1 j .

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-12 SNRC-1013 Drawings which cddress safety-related cmponents are identified as such and labeled QA Cat 1.

In addition to the engineering assurance and design control efforts implemented for Shoreham, LIICO has maintained a quality assurance program which meets the requirements of 10 CFR Part 50, Appendix B. W e quality assurance program will be applied to structures, systems and cmponents throughout operations, maintenance, station modifications and appropriate repair activities at Shoreham. We operational QA program will apply to all organizations performing design, design review and/or design audit activities. Section 3 of the LIICO QA manual describes the QA program requirements established to provide this control. Similarly, Section 4 of the LIICO QA manual describes the QA program requirements established to control procurement of safety-related material, equignent and services, while Section 5 of the LIICO QA manual assures that activities affecting the quality of safety-related structures, syst s and components during operations are controlled according to instructions, procedures and drawings.

Finally, LIICO has developed a ocxnprehensive set of station procedures to ensure that the design controls and criteria are maintained throughout the operational life of Shoreham. These procedures address the procurement of spare or replacement parts, the installation, inspection and testing of all cmponents and the perfor-mance of all maintenance activities.

The identification system, including unique part or mark numbers, developed during the design and construction phases, is maintained current during the operational phase. Identification is provided on specifications, drawings, purchase orders or other documents, maintaining traceability to manufacturing and inspection documents, heat numbers, and all test reports. The cmponent identification is either on the item or on record, directly and readily traceable to the it s . Physical identification is used to the maxiuum extent possible, and applied so it does not affect the item's function. The identification is verified throughout fabrication, assembly, shipping, and installation. We operating plant staff will maintain the identification system used during the design and construction ~ phase.

To faci]itate operations, LIICO developed a Caposite Cmponent List (CCL) . We CCL will suppleent existing engineering and operations controls. We CCL contains the unique cmponent mark number, the applicable procurement or construction specifica-tion reference and the records management file references. his list was developed based on the engineering, construction and preoperational test activities performed for-Shoreham. In order to ensure the accuracy of the list, LIICO has performed an engineering review of the list and developed procedures to control inputs and changes to the list.

13 SNRC-1013 Item 2.2.1.3 NRC Request - Provide a dascription of the process by which station personnel use this information handling system to determine that an activity is safety-related and what procedures for maintenance, surveillance, parts replacement and other activities defined in the introduction to 10CFR50, Appendix B, apply to safety-related emponents.

LIILO Response LIILO uses approved procedures for ensuring that safety-related systems, structures, and cmponents are identified, and are treated as such for various plant activities such as design modifications, maintenance, surveillance, parts replacement, repair handling, inspection, etc. When for any activity, a uniquely identifiable cmponent safety classification must be verified, the Nuclear Engineering Department controlled Camposite Component List is consulted to determine if the cmponent is safety or nonsafety-related.

Appropriate guidance is given in SP 12.013.01, " Maintenance Work Requests" for maintenance activities performed on safety-related syst ms, structures, and cmponents. The Scheduled Activity Worksheets, (SAWS), used for implenentation of surveillance and preventive maintenance programs, also indicate whether the activity is to be performed on safety or nonsafety-related components for utilization of applicable Quality Assurance controls.

Safety related replacement c mponents and parts have initially been evaluated to be campatible with the originally supplied camponents and parts per the requirements of the applicable Safety Related Purchase Specifications. New conponents (added as a result of plant modifications will receive an independent engineering and Quality Assurance evaluation for determining their safety classification, safety function, and failure effect on other safety related cmponents as described in the LIICO -

Program Description PD-NE-01, " Nuclear Organization Interim Management Control Program for Station Modifications". If a component is determined to perform a safety function, it will also receive a detailed review of the technical, packaging, shipping, tagging, traceability / documentation, vendor qualification, etc.

requirements. The above engineering and Quality Assurance evaluation and technical review shall also be performed before an existing plant cmponent can be upgraded or downgraded frm its original safety classification.

Approved procedures are used for receiving inspection, storage, and issue of safety related spare parts, materials, and omponents.

Item 2.2.1.4 NRC Request - Describe the managment controls utilized to verify that the procedures for the preparation, validation, and routine utilization of the information handling syst s have been followed.

LIICO Response

'Ihe LIICO QA Program requires that activities affecting safety are accmplished and controlled in accordance with documented instructions and procedures. .To cmply with this requirment the LIIOD Nuclear Operations Support and Nuclear Engineering Departments have prepared procedures. 'Ihese procedures describe the preparation, revision and control of the conponent classification designation system and provide

14 SNRC-1013 a description of the criteria used to classify structures, systems and cmponents.

W e procedures also require classifications to be applied to all structures, systes and cmponents that may be added due to modifications, repairs and additions to ensure that the quality of the plant is not degraded during its operating life.

We classification of the itms mentioned above are listed in the Cmposite Cmponent List (CCL) . Procedures have been prepared to describe the method used to revise the CCL. Wese procedures require any changes to be approved and documented.

In addition, the QA Audit Program described in Section 18 and Appendix A of the QA Manual identifies auditable functions of the Nuclear Operations Support and the Nuclear Engineering Department who prepare, review and control the CCL. All audits are performed in accordance with OA Procedures.

Item 2.2.1.5 NRC Request - Demonstrate that appropriate design verification and qualification testing is specified for procurement of safety-related cmponents. h e specifications shall include qualification testing for expected safety service conditions, and provide support for the licensee's receipt of testing documentation to support the limits of life recmmended by the supplier.

LIIOD Response The demonstration that adequate design verification and qualification testing has been in place for safety related conponents for Shoreham was achieved thrcugh rigorous equipnent qualification programs. Wese programs include 1) Seismic and Dynamic Qualification of Safety-Related Electrical and Mechanical Bquignent, and 2)

Environmental Qualification of Safety-Related Electrical Equignent. Both of these SNPS programs have received extensive review by the NER Staff and their consultants.

The extent of these reviews, which consisted of programatic reviews and also audits of staff selected program documentation, is identified in the Shoreham Safety Eval-uation Report (SER), NUREG-0420 and its supplements. _These specific programs are addressed in Sections 3.10 and 3.11 of the SER. .These programs demonstrated the adequate procurement, testing and documentation of these requiremnts. Required service conditions were identified for each cmponent along with verification that the qualification documentation adequately handled the service conditions required.

Documntation packages have been assembled for all cmponents requiring seismic and/

or environmental qualification the documentation package contains the following:

Environmental Qualification

1) Index, which fully identifies the content of the package'_by name, document number, forms, dates, revision number, titles and/or sumaries.

2) Required Service condition form (s) applicable to the qualification package.

3) Environmental Qualification Sumary Sheet sumarizing the' qualification docu-mentation review.

'4) Environmental Qualification Report Evaluation form; a checklist which guides l the document review assessment of the qualification documentation.

i

15

5) Stone & Webster Engineering Corporation Review of Supplier's Technical Document form 5040.51B, where applicable.

6) Test Report

7) Applicable supplanental analysis.

8) Pertinent backup correspondence (eg vendor letters, S&W letters, LILOO letters), and supplernental documentation.

Dynamic Qualification

1) Index of included and referenced documents.

2) Pariewers sumary of qualification basis.

3) Ccmpleted IGC form entitled " Qualification Sumary of Equipnent," including required acceleration level or response spectra, as appropriate.

4) Qualification report (s) .

The maintenance of qualified life of these safety-related cmponents is ensured through the Shoreham surveillance and maintenance programs. The Shoreham surveil-lance and maintenance programs include documented program plans and procedures, to assure that the safety-related equignent is maintained in a state of readiness and operability so that it will perform its intended safety function upcn demand.

The results of tha Environmental Qualification Program are directly input to the Shoreham maintenance and spare parts programs to ensure timely device or parts replacanent as needed due to identified qualified life limitations and to ensure timely perfor.ance of any other identified maintenance activities required to preserve the applicability of the qualification.

The Shoreham surveillance and maintenance program includes information supplied by the equipment manufacturers and veadors regarding required equipnent maintenance actions and their frequency.

The equipnent envirorrnental qualification packages are reviewed by the Station Technical Support Staff and sub-packages of relevant information assembled and distributed to station Maintenance and Instrument & Controls sections as appropriate. Upon receipt of these packages, the responsible groups arrange for any new information to beccme part of the ocmputerized program for plant maintenance which will alert them to upccming maintenance requirements on a timely basis.

Reporting of equipnent failures will be performed by the LIICO Nuc' lear' Engineering Department (NED) using the Nuclear Plant Reliability Data System . (NPRDS) program ackninistered by the Institute of Nuclear Power Operations (INPO) . During 1981, plans were cmpleted within LIICO to initiate the data collection and reporting effort. These plans have been canpleted and are now being updated in a manner consistent with the Plant Modification Program. Procedures will be prepared and approved outlining the responsibilities for the accurate and uninterrupted flow of

' data frcm the Shoreham Nuclear Power Station to the NPRDS.

16 The Shoreham RIuipnent Qualification Programs are on-going and continue to be in place as equipment is rodified or additional equignent added to the plant design.

Documentation packages are either revised to reflect the changes or new documenta-tion packages are created. Wese packages receive extensive review to ensure addition of any applicable new requirements to the Station Surveillance and Mainte-nance Programs.

Procurement LILOO has currently in place a Corporate Policy to 03ntrol procurement of material, equipment or services for Shoreham. %is Policy, NOC 4 (Nuclear Operations Corporate Policy) entitled, " Corporate Procurenent Document Control" identifies programatic requirements which include the requirenent that all procurement activity. The technical review ensures that appropriate technical requirenents are specified including seismic or environmental qualifications.

Each LILOO department with procurenent responsibility has specific departmental implenenting procedures in place to ensure ccnpliance with the Corporate Policy.

For procurement activities parformed by S&W as LILOO's agent, S&W Project Procedures are in place to ensure appropriate technical requirenents are specified. We S&W prccurement documents are further reviewed by LIIID in accordance with other LIICO procedures to ensure the doc nnents are technically adequate. --

Maintenance of Design Adequacy and Qualification To ensure adequate control of future design modifications at Shoreham, a Design Control Program was developed and is in place. Wis Program, entitled "LIICO Nuclear Organization Interim Managenent Control Program for Station Modifications" controls all modifications to the Shoreham plant fran initiation through implenenta-tion and final closecut. It requires that all associated documents and procedures are updated to reflect specific plant modifications. h e NBC Office of Inspection and Enforcanent, Region I, has reviewed this program in detail and reported favorably in Inspection Report 50-322/83-29.

Canponent Classification As new canponents are added or existing canponents nodified (via the Design Modi- '

fication Program) their classification is identified (i.e, safety related, nonsafety-related) and they are added to the Carposite Canponent List (OCL) . The -

CCL is a canputerized index of uniquely identified canponents for the Shoreham facility. Additions, deletions or changes to this list are procedurally controlled by the LIICO Nuclear Engineering Department.

This list ensures that procurenent and maintenance activities required for safety related canponents are followed for items designated as safety related. %e on-going RIuignent Qualification Programs are maintained to identify additions, deletions or changes and documentation packages are either revised or prepared to

-1 17 ,

4 l

! reflect these new requirements. . 'Ihis ensures that all required documentation was received, reviewed and is adequate to represent proper qualification.

! ' Item 2.2.1.6 NRC Rquest - Licensees and applicants need only to subnit for staff review the equipnent classification progrm for safety-related

ccmponents. Although not required to be subnitted for staff review, your equipment classification _ program should also include the broader 1

class of structures, systems, and vapnts inportant to safety required by GDC-1. (defined in 10CFR Part 50, Appendix A, " General Design Criteria, Introduction") .

LILOO Response

'Ihe Shoreham Nuclear Power Station program for the classification of safety related i cxmponents is described in Sections 2.2.1.1 through 2.2.1.5'of this subnittal.. With regard to the equipnent classification program in use at Shorelum for structures, systems and couponents inportant to. safety, we are participating in the Utility Safety Classification Group and are seeking a generic resolution' to the staff's-concern in this_ regard through the efforts of this Group. We do not agree that plant structures and ccuponents inportant to safety constitute a broader class.than the safety related set. As previously stated 'in SNRC-844 (3/2/83) and SNRC-854 (3/8/83) LILOO believes that non safety related plant structures, systems and

ocuponents have been designed, and are maintained, in a manner ccanensurate with '

] their inportance to the safety and operation of the plant.

5' Item 2.2.2 NRC Position - For vendor interface, licensees and applicants shall

establish, inplement and maintain a continuing program to ensure that a

vendor'information for safety-related ocuponents is-ocmplete, current and controlled throughout the life of their plants, and appropriately'-

referenced or incorporated in plant instructions and. procedures.

~

i i Vendors of safety-related equipnent should be contacted and an 4

-interface established. Where vendors cannot be identified, have gone

out of busimss, or will not supply information, the licensee or applicant shall-assure that sufficient attention is. paid to equipment

, maintenance, replacement, and repair, to ocupenaate for the lack of vendor backup, to' assure reliability ocmanensurate with its safety 1

~

function' (GDC-1) . . 'Ihe program shall be closely coupled with action .

2.2.1 above : (equipnent qualification) . 'Ihe program shall' include-periodic ccumunication with vendors;to assure that all applicable informatiorihas been received. 'Ihe program should use a system of-

, positive feedback with vendors for mailings containing technical. -

~ information. 'Ihis.could be acocuplished by liceinse acknowledgment

' for receipt .of technical mailings. - It shallialso~ define the inter-

, face and division of responsibilities among the' licensee.and the i

. nuclear and nonnuclear divisions _of f eir vendors that provide

. service (on safety-related 'equignent to cassure that requisite. control .

.of;andl applicable ~;instructionsifor maintenance work on safety-related -

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equipnent are provided. *N

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Nuclear Utilities Task Action' OtstnitteeF(NUDC)ito s 1p, #. -

address the NRC concerns.raissdi

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18- ShK -1013 in this item. 'Ib date LIICO has fully participated in this cmmittee activity which has resulted in a Vendor aquipnent Technical Information Program (VETIP) to be managed by INPO. Several VLTIP drafts have already been issued by INPO. A utility /INPO approved VETIP is expected at .the end of the first quarter of 1984.

Although LIICO fully expects to endorse the INPO issued VETIP, LIICO intends to withhold formal endorsement for 60 days in order to allow time for a Shoreham-specific evaluation Its 3.1 POST-MAINNCE TESTING (REAC'IOR TRIP SYSTD4 COMPONENTS)

It s 3.1.1 NPC Request - Licensees and applicants shall sutznit the results of their review of test and maintenance procedures and Technical Speci-fications to assure that post-maintenance procedures and Technical Specifications to assure that post-maintenance operability _ testing of safety-related cmponents in the reactor trip systs is _ required to be conducted and that the testing demonstrates that the equipnent is capable of performing its safety functions before being returned to service.

LIICO Response LIICO is ommitted to the requirements of the Technical Specifications to operate the Shoreham Nuclear Power Plant. The Technical Specifications outline the accept-able operating parameters, Limiting Conditions for Operation (IID), and surveillance reauirements for the safety related systems, -instrumentation channels and cmponents of the plant. 'Ihis includes the instrunentation channels and cmponents of the Reactor Protection System. In addition.to the regular intervals specified in the Technical Specifications, surveillance testing (functional operability test) is required to be performed on a system, instrumentation channel or'ccmponent, each time the cmponent is subject to a maintenance activity prior to returning the cmponent or.systs to service. Statiui procedure SP 12.013.01, " Maintenance Work Requests", (WR) outlines the requirements for post . maintenance operability tests

("Postwork Tests") following 'a cmponent maintenance or repair. ' The required -

post-maintenance operability test is ' determined by_ the. Operations Section and documented on the MWR form. In most cases, the dmonstration of the post-maintenance cperability of a cmponent will be satisfied by the performance of a Technical specifications surveillance-test (i.e. functional = operability test) .

The test may be performed by the Maintenance, I&C Operations or other sections (depending on the department jurisdiction over the' test),- however, the verification

.of inpimentation is the Operations Section's responsibility.

An Operations Section procedure SP 21.001.02, _" Return of Safety Related Cmponents to an Operable Status", suppleenting SP 12.013.01, " Maintenance Work Requests", was; also established to improve the method by which MNRs will be handled for retests.

'Ihis procedure extends the post-maintenance operability. test requirements for. all, safety related cmponents even if there is.no specific-surveillance testing

specified by the Technical Specifications or the Inservice.' Inspection Program. 'Ihe

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procedure also gives guidelines for determining the extent of post-maintenance operability tes_ ting for the various plant cmponents.

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SERC-1013 Itcm 3.1.2 NRC Request - Limnsees and applicants shall sulnit the results of their chect: of vgudor and engineering-reendations to ensure that any appropriate test guidance is inbluded in the tcst and maintenance procedures.or the Technical Spe,:ificationVwhere required.

LIILO Response ,

Appropriate vendor and engineeri:q recmmendations-have ben or are being incor-porated in the test vad mainteriance procedures and' Technical Specifications.

Depart 2nent level procedures req 6ffe the use of engineering documents and vendor manuals for the generation of tha various test and maintccice procedures. These source materials are inc1hded in the referenco section ?f the procedures.

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'Ihe LIICO program callsiior ryriodic review and updating of the station operating, test and maintenance ptoo~dures to assure that thf laterh vendor and enginwring reccanendations are ir.corpop.p3 and industry practices as followed. Station Procedure SP 12.007.01/ 7'Ibcnnical Correspondenr.= nnagButletins," assures that all inccning correspondenbe bulletinu, and vendor infonution are properly tracked and assigned for action within the Shoreham Plant Staff 'and the necessary updates to existing procedures and progtw a c.re made, as appropriate.

f,, , < e Item 3.1.3 N9C RequeE$ - Licensees and applicant's shall identify, if applicable, ~

a5y post-maintemnce test requirements in existing Technical Speci-fications which can be demonstrated to ' degrade rather than enhance safety. ,Appropria'Ee chances to these' test, requirements, with sup-porting juptification, chall be submiM;ed' for staff approval. (Note that actice 4.5 discusses on-line syttem function testing.)

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LIICO Response O j

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< 1 The Shoreham Technical S@ficat/on.s are currently undergoiry review and approval.

If, during our review aid 6se off:he SNPS Technical Specifications any requirements are fourd to degrade rather than enhance safety, the apinopriato changes and associated justification will be sub?.itted to the NBC. '

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Item 3.2 POST-MAINTEllANCT TESTING (N1 OIEER SAFEfY-RELATED CCt@ONENTS) s,-

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Itcm 3.2.1 NBC Ftxp*ste- Li'censees and applicants shall-submit a report documenting the extending of test and mainteriance procedures and Technical Specifications review to assure that post-maintenance operabi).ity testing of all safety-related equipment /is required to be ~

conducted and, that the testing demonstrates that the equipnent'is , y capable of performing its safety functions before being returned to . I service. ' _ , 9 ,,

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20 SNRC-1013 Itan 3.2.2 NBC Request - Licensees and applicants shall subnit the results of their check of vendor and engineering recorrnendations to ensure that any appropriate test guidance is included in the test and maintenance procedures or the Technical specifications were required.

Itan 3.2.3 NPC Pequest - Licensees and applicants shall identify, if applicable, any post-maintenance test requirenents in existing Technical Speci-fications which are perceived to degrade rather than enhance safety.

Appropriate changes to these test requirenents, with supporting justification, shall be subnitted for staff approval.

LIIOD Response (Itans 3.2.1, 3.2.2, and 3.2.3 _

'Ihe response to Iton 3.1 is also appl 4. cable to all other safety-related components.

Item .t.5 RFJCIOR TRIP SYSTEM RELIABILITY (SYSTEM FUNCTIONAL TESTDJG)

NFC Position - On-line functional testing of the reactor trip system, including independent testing of the diverse trip features, shall be perfonned on all plants.

Item 4.5.1 NRC Request - The diverse trip features to be tested include the scram pilot valve and backup scram valves (including all initiating circuitry) on GE plants.

Itau 4.5.2 NRC Request - Plants not currentiv designed to permit periodic on-line testing shall justify not making modifications to permit such testing. Alternatives to on-line testing proposed by licensees will be considered where special circumstances exist and where the objec-tive of high reliability can be met in another way.

LIIOD Response (Itans 4.5.1 and 4.5.2)

LIIDO is participating in the BWROG Salem A'IWS Catmittee effort to address the NRC concerns raised by these itens. GE has been requested to address these concerns and has issued several draft responses which have received internal camnittee review. A final Catinittee position on these items is expected during the second quarter of 1984. Although LIILO expects to endorse the Cattnittee's position, we will subnit a fcnnal msponse to the NRC within sixty (60) days after the receipt of the finalized BWROG position to allow time for a Shoreham-specific evaluation.

Item 4.5.3 NBC Request - Existing intervals for on-line functional testing required by Technical Specifications shall be reviewed to determine that the intervals are consistent with achieving Itigh reactor trip systen cailability when accounting for considerations such as:

1. uncertainties in canponent failure rates

2. uncertainty in catinon mode failure rates

3. reduced redundancy during testing

4. operator errors during testing 5.- canponent " wear-cut" caused by the testing Limnsees corrently not perfonning periodic on-line testing shall determine appropriate test intervals as described above. Changes to

21 SNRC-1013 existing requircxl intervals for on-line testing as well as the intervals to be determined by licensees currently not performing on-line testing shall be justified by information on the sensitivity of reactor trip system availability to parameturs such as the test intervals, ccuponent failure rates, and conmon mode failure rates.

LIILD Response LIICO is currently evaluating the proposed BWBoG Technical Specification Inprovement Ccmnittee effort with regard to item 4.5.3 in parallel with an internal review of the proof and review copy of the SNPS Technical Specifications. A decision regarding participation in the BWacG Tech Spec Cmmittee will be made within three (3) months.

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s e s t s , y 4 i

l APPENDIX A j I

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• Submitted: _ .,

Rrviewed/0QA Eng.: //* 1 u g h 1

Approved: WJcTQg4 5 (Plant Manage /) /

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MC-1 S? W-der 21.003.01 Revision 1 Date Eff. . I L/ L )f3 TPC TPC TPC OPERATIO!!S REPORTS 1.0 PURPOSE This procedure describes measures which operations personnel will use to inform appropriate levels of management of Reactor Scrams and events which may require reporting to outside agencies.

2.0 RESPONSIBILITY The Operating Engineer shall be responsible for assuring implementation of the 4 requirements of this procedure.

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SR2-1021. 200-6.421 DR%ig;p](

Na NO ? 1 bi)if;;

4

r 3.0 DISCUSSION 3.1 It is the responsibility of all personnel in the Operations Section to report conditions that may have an adverse effect on the plant or its operation to appropriate levels of management. There are several methods operating personnel have to report these conditions such as:

3.1.1 Maintenance Work Requests (HWR) - Reference 11.2 3.1.2 LILOO Deficiency Report (LDR) - Reference 11.10 3.1.3 Scheduled Activity Worksheets (SAWS) - References 11.3 and 11.4 3.1.4 Operating Logs - Reference 11.5 3.1.5 Scram Reports - this procedure 3.1.6 Appendix 12.1 Scram Report - this procedure #,

3.1.7 Appendix 12.2 Scram Evaluation - this procedure 3.2 The selection of one of the above control mechanisms for any particular condition or event will depend on the nature of che event and to some extent the preference of the individual responaible for the documentation and followup action.

3.3 This procedure will discuss the use of the Scram Report and the LILCO deficiency Report.

3.3.1 Scram Report - Appendix 12.1 A report used by operations personnel to report to appropriate levels of management the necessary information regarding all reactor scrams. A reactor scram is defined as a sudden shutdown of the reactor by the rapid insertion of control rods.

3.3.2 Scram Evaluation - Appendix 12.2 An evaluation used by operations personnel to insure that the f) causes for unscheduled reactor shutdowns as well as the response of safety-related equipment are fully understood prior to plant restart.

3.3.3 LILCO Deficiency Report (LDR) - see Reference 11.10 l4f

1. A report used by operations personnel to report conditions or events which may be reportable to outside agencies as defined in Reference 11.1 or

2. Other nonconformances as defined in Reference 11.10.

3.4 The following topics are contained in this procedure:

Pages 8.1 Scram Report 3 8.2 Scram Evaluation 3 8.3 LILCO Deficiency Report 4 4.0 PRECAUTIONS N/A r

5.0 PREREQUISITES N/A SP 21.003.01 Rev.1 Page 2

6.0 LIMITATIONS AND ACTIONS N/A l 7.0 MATERIALS AND/OR TEST EQUIPMENT ,

N/A 8.0 PROCEDURE 8.1 Scram Report'  ;

. 8.1.1 Af ter a scram has occurred, the Watch Engineer or his designee shall prepare a Scram Report.

8.1.2 The Watch Engineer shall review and approve the Scram Report and attachacats and forward the report the the Operating Engineer.

8.1.3 The Shift Technical Advisor shall review the report for technical lAl i

accuracy.

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l 8.1.4 The Operating Engineer shall ensure the Plant Manager, Chief p/

> Operating and Technical Engineers, Station Technical Support Manager and OQAE receive a copy of the report.

8.1.5 The Operating Engineer shall ensure charts, recorders etc. are l2/

reviewed to determine if any infaty limits, trip setpoints, or other limits have been exceeded.  !

8.1.6 The Operations Section shall ensure the " Scram No." on the Scram l4f Report reflects the total number of scrams to date and ensure cyclic and transient limits of Table 5.7-1 of Technical l Specifications are not exceeded.

8.1.7 The Technical Support Manager shall ensure reportability require-ments of Reference 11.1 are met.

8.1.8 The PAC shall maintain records associated with all Reactor Scrams. lt /

8.2 Scram Evaluation-8.2.1 After a scram has occurred, the Watch Engineer or his designee .

shall prepare a scram evaluation.

8.2.2 The Watch Engineer shall review and approve the scram evaluation.

8.2.3 The Shift Technical Advisor shall review the evaluation for technical accuracy. 7 8.2.4 ' The Scram Evaluation shall be reviewed withL the Operations Engineer prior to commencing a reactorEstartup.

1 SP 21.003.01 lRev. 1 Page 3

8.2.5 The Scram Evaluation shall be attached to the Scram Report and forwarded with the report to the Operating Engineer.

8.3 LILCO Deficiency Report g, 8 .3.1 Any station personnel may initiate a LDR in accordance with Reference 11.10.

8.3.2 The initiator should indicate on the LDR whether or not a Technical Specification (TS) or Environmental Technical Specification (ETS) wie violated and reference the specification number on the LDR.

8.3.3 The initiator shall immediately inform'the On-duty Watch Engineer of all violations of TS or ETS.

g, 4

8.3.4 The LDR will be processed in accordance with Reference 11.10. l 9.0 ACCEPTANCE CRITERIA N/A 10.0 FINAL CONDITIONS i

N/A l

11.0 REFERENCES

11.1 SP 12.009.01, Station Reporting Requirements - NRC.

11.2 SP 12.013.01, Maintenance Work Requests.

11.3 SP 12.015.01, Preventive Kantenance Program.

11.4 SP 12.016.01, Surveillance Program.

11.5 SP 21.002.01, Operations Logs and Records.

11.6 Technical Specifications, 9/1/79, Section 6.0 11.7 Regulatory Guide 1.16 Rev. 4, 8/75, Reporting of Operating Information -

Appendix A, Technical Specifications.

11.8 Regulatory Guide 10.1, Rev. 3, 5/77, Compilation of Reporting Requirements for Persons Subject to NRC Regulations.

11.9 Environmental Technical Specifications, Rev. 3,~ Saction 5.6.

11.10 QAP-S-15.1, SJte OQA Nonconformance Control.

12.0 APPENDICES 12.1 SPF 21.003.01-6, Scram Report.

-12.2 ' SPF 21.003.01-2 ,LScram Evalbation.

l4 I i

-SP 21.003.01 Rev. % ,

Page 4 g -- .- -, , , - - - - - -. -e

• Appendix 12.1 Pcg3 I of 2 SCRAM REPORT SCRAM NUMBER TIME OF SCRAM 1.0 Mode Switch Position

2.0 Reactor

Critical Subcritical 3.0 Plant Evolution: Startup Shutting Down Steady Operation 4.0 Plant Conditions: ,

4.1 Reactor Power: MWT 4.2 SRMS: A. cps, B. c ps , C. _ c ps ,

D, cps 4.3 IRM: A.  % range , D.  % range , G  % range B. JE range . E  % range , H  % range C.  % range . F  % range ,

4I

4.4 APRMS

A.  % power, C.  % power, E.  % Power.

B.  % power, D.  % power, p.  % power, 4.5 Generator Load MWE 4.6 Reactor Pressure: pai 4.7 Steam Flow: lb/hr.

4.8 Feedwater Flow: lb/hr.

4.9 Vessel Water Level:

4.10 Recire System: Loop Hanual _, Master Manual 4.11 Recirculation Drive Flow: Loop A Loop B  !

.I CPH CPM  :

l 4.12 Core Flow: _ lb/hr.

SPF 21.003.01- 6 SP 21.003.01 .Rev. 1 Rev. 1 Page 5

Appendix 12.1 Pega 2 of 2 5.0 Relief Valve Operation Sununary Valve Operated idpenings Manual Auto Comments RV-092A RV-0928 RV-092C RV-092D RV-092E RV-092F RV-092C RV-092H RV-092J RV-092K RV-092L 6.0 External Visual Examination of the suppression chamber completed. (only required if there has been a safety / relief valve operation with the suppression chamber average water temperature greater than or equal to 1970 F and reactor coolant system pressure greater than 200 peig.)

7.0 Suppression Chamber drywell vacuum breakers demonstrated operable per SP 24.654.02, l if there has been any discharge of steam to the suppression chamber from the safety relief valves.

8.0 Appendix 12.2, Scram Evauluation Report completed and attached.

9.0 Computer sequence of svents print out attached.

10.0 A copy of the Watch Engineer's and Control Room Log attached.

11.0 Recorders marked to indicate date and time, l

(

Watch Engineer Shift Technical Advisor l

l SPF 21.003.01-6 SP 21.003.01 Rev. 1 C _ Pc~0 6

- Appendix 12.2 Pcgs 1 of 2 SCRAM EVALUATION 1.0 Screw Number ,

2.0 Recorder Charts Reviewed 2.1 Reactor Power 2.2 Reactor Level 2.3 Reactor Pressure 2.4 Core Flow 2.5 Feedwater Flow 2.6 Steam Flow 2.7 Station Vent Gas Monitor t

3.0 Computer Logs Reviewed 3.1 Sequence of Events . Log.

Ll 3.2 Alarm Log.

4.0 Equipment Malfunction (include failures of CRD's to insert) 5.0 Procedure Inadequacies 6.0 Operating Errors 7.0 Cause of the Scram l

l NOTE: R.O.C. must be consulted prior to performing a reactor startup if the cause of the scram can not be positively identified.

I 8.0 Evaluation to Restart the Reactor Has the condition which caused the scram been corrected?

8.1 8.2. A sample of reactor coolant has been taken and activity levels are normal.

8.3 All ECCS equipment operated normally consistent with l

plant conditions.

i i SPF 21.003.01- 2 SP 21.003.01 Rev. 1 Page 7 l Rev.1

, Appendix 12.2 Pcg3 2 of 2 I

8.4 All control rods inserted into the core and reactor power decreased at approximately a-90 second period.

8.5 Reactor water level was maintained within normal range consistent with the expected transient.

8.6 Reactor pressure was control normally consistent with the experienced transient.

8.7 All containment parameters were maintained within cormal limits consistent with the experienced transient.

8.8 There are no abnormal radiation levels in the plant.

8.9 A determination has been made from the sequence of events recorder that the reactor protection system operated normally.

9.0 Corrective Actions 9.1 Is there any equipment out of service which would prevent c the unit from being returned to service?

9.2 Are there any repairs which should be made prior to returning the unit to service?

Report Prepared by: Date Time Watch Engineer Review Date Time Shift Technical Advisor Review Date Time Report reviewed with the Operations Engineer and concurrence received.

Watch Engineer Date Time _

SPF 21.003.01-2 SP 21.003.01 Rev. 1 Rev.1 - Page 8 l

- _____- -_-___- --__-_.__-____-_______ _ _ _ _ _ . _ - - - - .---_______-_---___-_______--__________-___a

APPENDIX B +

SEQU[NCE Of EVENIS COMPUTER P01:1S VS. ASSOCI ATED ALARM P08N15 (CSSS/ BOP) Pag 3 1 cf 4 g July 11, 1933

• SHOREllAM NUCLEAR POWER STATION - UNIT 1 Rev. 2/JKS ,

Annun Annun 4 Comp. Comp. Assoc.

Mid Pt.. Alarm Window Sys. Pat .

.. gumbe r 14. Computer Point Descripijpg Number lofation No. l 29m Alarm Description C1501 D500 Scram Disch Vol Lvl Ch-A 1203 A9-D3 1C71 603 Discharge Vol Hi Wtr 4 vi Trip .

'C1502 D501 Scram Disch Vol Lvl Ch-B 1203 A9-D3 1C71 603 Discharge Vol Hi Wtr Lyl Trip

-C1503 D502- Scram Disch Vol Lvl Ch-C 1187 A10-03 1C71 603 Discharge Vol Hi Wtr Lvl Trip

• C1504 .D503 Scram Disch Vol Lvl Ch-D 1187 'A10-D3 1C71 603 Discharge Vol HI Wtr tvl Trip C1509 D507 MSiv not full open Ch-A 1204 A9-C4 1C71 603 MSIV Not full Open Trip C1510 0508 MSIV not full open Cte-B 1204 A9-C4 1C71 603 MSIV Not Full Open Trip C1511 D509 MSIV not full open Ch-C 1188 A10-C4 1C71 603 MSIV Not Full OpOn Trip C1512 D510 MSIV not full open Ch-d 1188 A10-C4 1C71 603 MSIV Not Full Open Trip -

.C1513 D511 Contet Hi Press Ch-A 1205 A9-B1 1C78 603 Pri Cntat Hi Press Trip C1514 'D512-Contet Hi Press Ch-B 1205 A9-81 IC71 603 Pri Cntet Hi Press Trip C1515 D513 ContmL Hi Press Ch-C 1189 A10-81 1C71 603 Pri Cntet Hi Press Trip C1516 D514 Contet Hi Press Ch-D 1189 A10-Bl IC71 603 Pri Cotat Hi Press Trip

.C1517 D515 Reactor Hi Press Ch-A 1206 A9-82 1C71 603 Rx Vessel Hi Press Trip p C1518 D516 Reactor Hi Press Ch-B 1206 A9-B2 1C71 603 Rx Vessel Hi Press Trip C1519 D517 Reactor Hi Press Ch-C 1190 A10-82 1C71 603 Rx Vessel Hi Press Trip C1520 D518 Reactor Hi Press Ch-D 1990 A10-82 1C71 603 Rx Vessel Hi Press Trip ,

C1521 D519 Reactor Lo Wtr Lyl Ch-A 1207 A9-83 1C71 603 Rx Vessel to Lvl Trip

.C1522 D520 Reactor Lo WLr Lvl Ch-B 1207 A9-83 1C71 603 Rx Vessel Lo Lvl Trip C1523 D521 Reactor to Wtr Lyl Ch-C 1191 A10-83 1C71 603 Rx Vessel to Lvl Trip i

-C1524 D522 Reac to r lo Wt r ' Lyl ' Ch-D 1191 A10-83 1C71 603 Rx Vessel Lo Lvl Trip l 603 Mn Ste Line HI Rad T rip

• C1525 D523 Main Steam Line Hi Rad A 1208 A9-84 1C71 C1526 D524 Main Steam Line Hi Rad B 1208 A9-84 1C71 603 Mn Stm Line Hi Rad Trip i C1527 0525 Main Steam Line ill Rad C 1192 A10-B4 1C71 603 Hn Stm Line Hi Rad trip 3 )'

C1528 D526 Main Steam Line HI Rad D 1892 A10-84 1C71 603 Mn Stm Line Hi Had Trip .

C1529 D527 Neutron Monitoring Ch-Al 1212 A9-C3 1C71 603 Neutron Monitoring Sys Trip C1530 0528 Neutron Monitoring Ch-B1 1212 A9-C3 1C71 603 Neutron Monitoring Sys Trip _,

C1531 0529 Neutron Monitoring Ch-A2 1196 A10-C3 1C71 603 Neutron Monitoring Sys Trip ,

C1532 D530 Neutron Monitoring Ch-82 1196 A10-C3 IC71 603 Neutron Monitoring Sys Trip C1535 D566 RX Man Scram Ch-C 1216 A10-A3 1C71 603 Rx Man Scram Sys B C1536 D567 - RX Man Scram Ch-D 1216 .A10-A3 IC71 603 Rx Man Scram Sys B C1537 D533 RX Man Scram Ch-A 1215 A9-A3 1C71 603 Rx Man Scram Sys A C1538 D534 RX P4an Scram Ch-B 1215 A9-A3 1C71 603 Rx Man Scram Sys A i C1539 D535 RX Auto scram Ch-A 1213 A9-A2 1C71 603 Rx Auto Trip Ch A2 l C158 0 D536 RX Auto Scram Ch-B 1214 A10-A2 IC71 603 Rx Auto Trip Ch B2 C1552 D537 NSS Spare Sequence Annun IC71 C1553 D538 .TSV Closure Ch-A 1211 A9-DI 1C71 603 Turb Stop Vv Closure Trip C1554 0539 TSV Closure Ch-B 1211 A9-D1 1C71 603 Turb Stop Yv C19sure irrp

.C1555 D540 TSV Closure Ch-c 1195 A10-D1 IC71 603 Turb Stop Vv Closure Trip C1556" D541 - TSV Closure Ch-D 1195 A10-D1 1C71 603 Turb Stop Vv Closure Trip C1557 D542 TCV fast Close Ch-A 1209 A9-D2 IC71 603 Turb Control Vv fast close Trip C1558- D543 TCV Fast Close Ch-B 1209 A9-D2 1C71 603 Turb Control Vv fast close Trip C1559 D544 TCV Fast Close Ch-C 1193 A10-D2 1C71 603 Turb Control VV Fast close Trip C1569 D545 TCV Fast Close Ch-D 1193 A10-D2 IC71 603 Turb control Vv rast close Trip C1561 -D546 APRM Ch-A Upscale Lvl 1228 A10-A9 1C51 603 APRM Bus A Upsc8 Trip or inop C1562 ' : D547 - APRM Ch-B Upscale Lvl .1229 A10-A10 1C51 603 APRM Bus B Upsci Trip or inop u --

7 - _

3

- . ** -. = _ _ . . _ - _ _ _ - - _ . _______ -

6

y .

n e

).

SEQUENCE OF EVEOTS COMPUTER PolNTS VS. ASSOCI ATED ALARM PolCTS (NSSS/ BOP) Pags 2 cf' 4 July II, 1933 *w SHOREHAM NUCLEAR POWER STATION - UNIT 1 Rov. 2/JkS -

Comp. ' Comp. Assoc. Annun Annun .

Ciid Pt. Alarm Window Sys. Pnl .

Numbtf Id. f,nAggier Poini Descriotion NumbgI Location Hem & AIa re Do3qdat ion C1563 APRM Ch-C Upscale Lvl 1228 A10-A9 1C51 603 APRM Bus A Upsci Trip or Inop D548 - 1229 A10-A10 603 APRM Bus 8 Upscl Trip or inop C1564 - D549 APRM Ch-D Upscale Lvl 1C51 C1565 0550 APRM ch-E upscale Lvl 1228 AIC-A9 1C51 603 APRM Bus A Upsci Trip or inop -

C1566 0551 APRM Ch-F Upscale Lvl 1229 A10-A10 1C51 603 APRM Bus 8 Upsci Trip or inop C1567 D552 IISS Spaee Sequence Annun

'C1568 D55 3 .- MSS Spare Sequence Annun C1569 D554 IRM Ch-A Upscale Lvl 1239 A10-A7 1C51 603 IRM Trip Sys A Upscl/Inop C1570 ... D555 .IRM Ch-B Upscale Lvl 1241 A10-A8 1C51 603 IRM Trip Sys 8 Upscl/inop -

C1571 -- 0556- IRM Ch-C Upscale Lvl 1239 A10-A7 1C51 603 1RM Trip Sys A Upscl/Inop C1572 D557 BRM Ch-D Upscale Lyl 1241 A10-A8 1C51 603 IRM Trip Sys 8 upscl/Inup 603 C1573l D558 'IRM Ch-E Upscale Lvl 1239 A10-A7 1C51 IRM Trip Sys A Upscl/Inop C1574 D559 IRM Ch-F Upscale Lvl 1241 A10-A8 1C51 (03 IRM 1 rip Sys 8 Upscl/Inop C1575' D560

• IRM Ch-C upsca le Lvl 1239 A10-AT 1C51 603 IRM Trip Sys A Upscl/Inop C1576 0561 IRM Ch-H Upscale Lvl 1241 A10-A8 1C51 603 IRM Trip Sys 8 Upscl/inop C1579 D562 IISS Spare Sequence Annun C1584. D563 NSS Spare Sequence Annun C1591' D564 ' NSS Spare Sequence Annun C1592 D565 .NSS Spare Sequence Annun C1756 -D626 APRM Thennel Lvl Trip A 1228 A10-A9' ICSI 603 APRM Bus A UpscI Trip or inop C1757 D627 APRM Thermal Lvl Trip 8 1229 A10-A10 1C51 603 APRM Bus 8 Upsci Trip or inop C1758 D628 APRM Thermal l_vi Trip C 1228 A10-A9 1C51 603 APRM Bus A Upsci Trip or inop .

C1759 'D629 APRM The rma l Lvl Trip D 1229 A10-A10 1C51 603 APRM Bus 8 Upsci Trip or inop C1760 D630. APRM The rma l . Lvl Trip E 1228 A10-A9 1C51 603 APRM Bus A Upsci Trip or Inop -

C1761' 'D631 APRM Thermal Lyl Trip F 1229 A10-A10 1C51- 603 APRM Bus 8 Upsci Trip or inop C3801 D140 Mn Turb Exh Hood Hi Temp 0142 209F-E8 1N32 8 Main Turb Exh Hood Hi Temp 1 rip C3802 D141' Mn Turb Stator Cig Loss 0132 209F-C10 IN32 8 Win Turb Stator Cig Loss Trip C3803 D142 Turb Shaf t to PP Lo P 0129 209F-E7 1N32 8 Main Turb Shaft Pump Lo Press Trip C3804 D143 Mn Turb ihrust Srg Wear 0134 209F-C8 .1N32 8 Main lurb Thrust Brg Wear Trip =

C3805~ D144 Mn Turb 125V DC 1 rip 0144 209F-D9 1N32 8 Main Turb EHC 125V DC Trip Loss C3806 D145 [HC Loss or Fluid 0141 209F-D8 IN32 8 EHC Loss or Fluid Trip

.C3807 D146 MN lurb Ilo Speed Signal 0138 209F-810 IN32 8 Main Turb No Speed Signal Trip C3808' D147 Mn Turb Vib Hi 0143 209F-D10 IN32 8 Nin Turb vib Hi Trip C3809 Dl48 ~Mn lurb Lo Cond Vac 0130 209F-E9 1N32 8 Main lurb Cond to vac Trip

.C3810 D149 toss of 24V DC Pwr 0140 209F-A9 IN32 8 Main Turb Tripped C3811 .D150 Mn Turb Man Trip 0139 209F-88 1N32 8 Main Turb Manual Trip C3812 D151 Mn Turb Tripped 0140' 209F-A9 IN32 8 Main Turb Tripped C3813 Dl52 Mn Turb Backup overspeed 0137 209F-89 IN32 8 Main Turb Backup overspeed Trip C3814 D153 Mn Turb Msr Drn ik Hi 0135 209F-C9 IN32 8 Main Turb Msr Drn ik Lovel Hi Trip C3816 ' D155 : Loss or Emerg Trip Fluid 0140 209F-A9 IN32 8 Main Turb Tripped C1532 D159 80P Spara Sequence Annun C3833 D160 80P Spa re - Sequence Annun C3834 D161 80P Spare Sequence Annun C5417 D156 Loss or Circ Wtr 0140 209F-A9 1 Nil 8 Main Turb Tripped C5433 D154 Loss or Circ Wtr 0140 209F-A9 1N71 8 Main Turb Tripped fC7720 D101 Volts /Hz CT 110 P/C Prot 0236 209C-C5 1R61 8 Main Con Volts per Hertz HI

b

• ~

f s. _ .

. f p- -

SEQUENCE CF EVE".iS COMPUTER PolNIS VS. ASSOCI ATED ALARM Polc15 (CSSS/80P) PeCo 3 af 4 '

July 11 1983 .

SHOREHAM NUCLEAR POWER STAil0N - UNIT 1 Rev. 2/JKS -

Ccap. Comp. Assoc. Annun Annun =.

Nid . Pt. Alarm Window Sys. Pol .

Ilumber I d . . .. Gggstutor Point De sc ript ion Number Location L Ioc. Alare Description C7721 0102 -Volts /Hz GT 118 P/C Prot 0236 209C-C5 1R61 8 Main Gen Volts per Hertz Hi .

C7722 .D117 Unit Dirr PRI Prot 0212 209C-A5 1R61 8 Unit Pri Prot 1 rip C7723 0118' . Con Ilout Crd Pri Prot 0212 209C-A5 1R61 8 Unit Pri Prot T ri p -

C77'4 D119 Max Exc Lim Unit Pri Prot 0212 209C-A5 1R61 8 Unit Pri Prot Trip

-C7725 D120 MI6 XfMR 1A Sudden P 0212 209G-A5 1R61 8 Unit ?ri Prot Trip C7726 D121 MN XFMR-18 Sudden P 0212 209G-A5 1R61 8 unit Pri Prot Trip C7727 D122 Mel Gen Line Dirr Prot 0147 209C-04 1R61 8 Main Gen Prot Lcts C7728 .D123 Unit SKR 1350/1360 Fall 0212 209G-A5 1R61 8 Unit Pri Prot Trip -  ;

-C7729 . DIP 4 Unit Antimotoring Prot 0212 209C-A5 1R61 8 Unit Pri Prot Trip i C7730 D125 .951 XFMR 1A Dirr S/U Prot- 0213 209G-A6 1R61 8 Unit Backup Prot Trip C7731' D126- DWI XIMR 18 Dirr 8/U Prot' 0213 209C-A6 1R61 8 Unit Backup Prot Trip C7732 D127 ' , MII Con Dirr 8/U Prot 0213 209C-A6 1R61, 8 Unit Backup Prot Trip C7733. D128 Exc-Alt Dirr 8/U Prot 0213 209C-A6 1R61 8 Unit Backup Prot Trip C7734: D129 Con Neut Grd 8/U Prot 0213 209G-A6 1R61 8 Unit Backup Prot Trip C7735 .0130- Gen Grd Sus S/U Prot 0213 209C-A6 1R61 8 Unit Backup Prot Trip 4 C7736 'D131 Loss or Exc 8/U Prot 0213 209G-A6 1R61 ft Unit Backup Prot frip -

l C7737 D132- Con Rev Pwr B/W Prot 0213 209C-A6 1R61 8 Unit Backup Prot Trip C7738 D135 set XFMR 1A Crd 8/U Prot 0214 209C-86 1R61 8 . Sys Backup Prot Trip C7739 D136 sul XIM'1 18 Grd 8/U Prot 0214 2090-86 1R61 8 Sys Backup Prot Trip C7740 D137 Nog Phase Seq 8/W Prot 0214 209C-86 1R61 8 Sys Backup Prot Trip j

.C77%I D138 Ortset MMO Aly 8/U Prot 0214 209C-86 1R61 8 Sys Backup Prot Trip .

Unit Pri Prat Trip I

-C7742 D139 - Con Fid Grd PRI Prot 0212 209C-A5 1R61 8 C7743 .D157 Unit Gen Leads OC Prot ' 0212 209C-A5 1R61 8 Unit Pri Prot Trip

• l

-C7760 D162~ . Sep Spa re Sequence Annun r C7761- ~D163 Sop Spare Sequence Annun C7762 ~0164 Sop Spare Sequence Annun C7763 D165 - Sop Spare sequence Annun =

}

C7821 D166 Sep Spare Sequence Annun

.C7822 0167 - Sop Spare Sequence Annun ll C7823' D103 MSS 8/U OC Prot. 0219 209H-A2 IR62 8 NSS Xfer Backup Prot Trip' l

l C7824 D104 NSS 8/W CR0 OC x WDC 0219 209H-A2 1R62 8 NSS Xrer Backup Prot Trip C7825 DIOS .NSS 8/U CAD OC Y WDG . .0219 209H-A2 1R62 8 NSS Mrar Backup Prot Trip C7826 D106 NSS Xrer Sudden Press .0219 .209H-A2 1R62 8- MSS Xrar Backup Prot Trip

<C7827 D107' NSS Xfor Dirr Pri Prot 0218 209H-A1 1R62- 8 NSS Xrar Pri Prot Trip C7828 DIOS - NSS Skr 1350/1360 fall 0219 209H-A2- 1R62 8 NSS Xror Backup Prot Trip C7829 D109' RSS Xrar Dirr Pri Prot 0220 209H-A6 .1R62 8 RSS Xfer Pri Prot Trip l 8 RSS XFor Backup Prot Trip C7830- D110 Ras 8/U OC Prot .

0221 209H-A5 1R62 C7831 D111 .RSS 8/U Grd OC x WDC 0221 209H-A5 1R62 8 RSS Xrar Backup Prat Trip

( C7832 Dil2 . RSS 8/U Grd OC Y WDC 0221 209H-A5 1R62 8 RSS Xrar Backup Prot Trip C7833 D113 RSS Xrar Sudden Press 0221 209H-A5 1R62 8 RSS Xror Backup Prot Trip C7834- D114 RSS Xrer 69KV Skr 8/U . 0221 209H-A5 1R62 8 RSS Xrer Backup Prot Trip C7835 D115~ RSS Xrar 69KV Bus Diff 0220 209H-A6 1R62 8 RSS Xrar Pri Prot Trip t' -CT836 ~D116- 13kV Cas Turb Skr faiI 0221- 209H-A5 1R62 8 RSS Xter Backup Prut Trip .

~C7837 D133 Gen Skr 1310 8/U Prot 0219 209H-A2 1R62- 8 MSS Xrar Backup Prot Trip

[: 'C7838' D134 Con Skr 1330 8/U Prot 0219 209H-A2 1R62 8 NSS Xrer Backup Prot Trip F

I L

u I

l r.

L L

. - - . . . . . . .. . - = -. ---

g. .;

SEQUENCE OF EVECTS COMPUTER pol *GIS VS. ASSOCIATED ALARM Pott!i3 (CSSS/ BOP) Psgo 4 sf 4 '

July 11, 1983 *

. SHORDIAM NUCLEAR POWER STAT 104 - UNIT 1 Rev. 2/Jks

• Comp. Comp. Assoc. Annun Annun - .

. Cid Pt.- . . Alarm Window Sys. Pnt .

Hugber Id. - Computer Point Descript ion Numbe r tocation Ngm toc. Alare Description C7841 0158~ NSS Gen tsads OC Prot'. 0219 209H-A2 1R62 8 NSS Xter Backup Prot Trip C7842 D168 Bop Spa re Sequence Annun -

' C7843 D169 .Dop Spare Sequence Annun .

)

O O

N I

e e

k 4

.e b

'e

y-APPENDIX C 12d35-63' I33558 ALM JA543 Rod OUT BLOCK 04 135143 h0RM A523 APH:4 UPaCL NOR:4 13'143 2 NORas A343 400 OUf eLV;( OFP c

il3 Dol 5 ALM A32d APWM UPdCL ALA -

, '135dl5 ALM AS43 OtU OUT BLOCK ON 140347 44 SEu D547 APdM CHNL M UPSCALE LVL f.4I P. g I40347 44 St!J ud28 NcUTRoi M041Toa1:G Cil 9i T4iP 140347 44 *SEJ Ib29 hEd!' HON mk4110glNLi Cd 42 THIP

.140347 44 SEO 0536 HX AUfo SCHAM CHtIL 3 T41P $qq b k d .$ QQ g l40332 35 SEJ Dd26 NEdIRON MONIl~ORING Cd HI HSEf 140552 1 35

• SEQ 0529 NtiUTRON *>NITORitAi CH A2 RSdT l 140332 Jo *SEJ Dd47 APRM CdhL 3 UMCALE LVL HSEf d t 140551 ALM A537 APM tsf PASSED CH E . UN I<0553 NORM ~A528 APR4 UPSCL NOR4' l40553 NORM AS43 ROD OUT BLOCK - UFF I40D55 57 SEJ Dd36 RX AUTO SCdAM CdhL R RSET OFF r 11C0d36 NORM A531 APRM BfPASSED CH B . } --

IC0923 ALM -El27 EMER Bus 101 DG PRI PROF l'0?8 HIP

.1(0927 ALM ., A537. APRM Bf PASSED CH B IC0936.NONM E127 EMER BJd 101 D3 PRI P40f NOHM 140939 NORM A531 APdd HfPASSED CH B. OFF gg ggg .,

11<0751 NORM E237 DG 103 LUBE O!L HfR NORM .,

-0.4 STATUS

)i ,

101002 NORM.B001 APRM B 4 PodcR -

i, l' Tsun cyctr PT.tn. 13 A*q

" a - 141203 ALM- E127 EMER Hus 3 01 DG PWI PRuf ON IRIP '

=m 141204 ALM A537 APR4 Bf PASSED CH D ====== n .m u charaners b l("l5.140RM hl21 EMER RUS 101 15 PRI PWOT r40R4 ICl2*5 ALM--El21 EMER BJS 101 DG PHI PWof 4 HIP -

' l41250 NORM cl21- LWR BJS 301 in PHI PROT N044 F ICl3l0 ALM - cl21 EMcR BUS 101 D3 PRI PROT TRIP _

848321 NORM E121 EMER BUS '801 DG PRI PROT NORM

,141630 ALM E236 DG 102 LUBE UIL dTR LCTL d

101631 ALM . hl21 EMER BJS 101 - D ' PRI PRuf fRIP / \

141703 LRL 8003: APRM D % PoncR

'l41111 NORM A537'APR4 RfPASStu Cd D OFF Icid24 ALM A523 APdM UPSCL ALd ON  %

I41824 ALM A543 Hud OUT BLOCK. -- -

142342 32 St0 Do*9 APR4 CHNL D UPSCALE LVL TRIP 1 142242 33 bEQ D530 hEuldON MON!f0HING CH 42 TRIP 142542 33 SEO 0536 RX AUTO SCRAM CHNL R T419

-142624 49' SEJ D530 hEdfRu1 Mutil10 RING CH H2 WSET

%%NCQ Ok (\)(C MQ

', 142624 47 *Sc0 Dd49 APR4 CddL 0 UPSCALL LVL RScT 142023 ALM- Aa3d APR4 BYPAdSED CH D ON RSEf 842623 ~5 5 SEJ Da36 RX Auf0 dCdAM CHNL 8 142625 NORM A52d APR:4 UPSCL NORM l<2925 H0HN A543 R(U 00f BLOCK OFF

. IG629 ALM ' E233 DG 103 AlH CPRN 3C . LCTL 14L629 ALM ~c234 UG 103 AlH cpu 5R.4C- LCTL I4/J02 NORM B003 APda D 4 P(mER -0.6 i 42703 140RM : A33J APRM Hf9ASSED Cd D. - OW E '

142950 ALM - Abel APRM Bf PASSED CH F ON

> 143303 LdL 6005. APH 4 t- A 904ER 3 43105 NOR:4 A541 APRM Rf PAdSED CH r 2 OW t 14330V NORM E127 EMER HJs not DG PRI PkOT NO42 103406 ALM L A323 APR 4 UPSCL AL4

-1C3436 4LM A343 won udf HLi>;( .UN r.0$t

• I 43432 NOR4 A323 APR 4 UPSCL 143442 NOHN A343 HUD '00f RLLA;A OH

'1432J0 \LM Aa2J APR4 dP WL ALA

APPENDIE D Process Computer Post Trip Log Analog Parameters A. NSSS Log .

1. APRM A

2. APRM B

3. Reactor Pressure

4. Core Dif ferential Pressure

5. Jet Pump Total Flow

6. Feedwater Flow A

7. Feedwater Flow B

8. Reactor W ter Level

9. Total Steam Flow

10. Feedwater Temperature A B. BOP Log

1. Generator Current Phase A

2. Generator Voltage

3. Condenser Pressure A

4. Condenser Pressure B ~

5. Condensate Booster Pump Discharge Pressure

6. Condensate Pump Discharge Header Pressure

7. Condensate Booster Pump A Flow

8. ' Condensate Booster Pump B Flow

9. Feedwater Pump A Suction Flow

10. Feedwater Pump B Suction Flow

11. Feedwater Pump A Discharge Pressure 12 Feedwater Pump B Discha_rge Pressure

13. Generator Stator Cooling Water Temperature

14. Generator Field Current

15. Generator Field Voltage

16. Drywell Pressure

17. Low Pressure Turbine A Exhaust Hood Temperature

18. Low Pressure Turbine B Exhaust Hood Temperature -

19. Turbine Shaft Driven Oil Pump Discharge Pressure 20 Turbine Bearing oil' Supply Pressure

21. MSR Drain Tank A Level

22. MSR Drain Tank B Level

23. Turbine Thrust. Bearing Wear

24. thru 31. Main Turbine Bearing Vibration (8 Points)

32. and 33. Exciter Bearing Vibration (2 Points)

34. Turbine Speed

35. Spare

36. Condensate Pump A Flow (Calculated Valve)

37. Condensate Pump B Flow (Calculated Valve);

3 8. - Unit Gross Power (Calculated Valve)

39. Normal Station Service Transformer Power (Calculated Valve)

40. . Reserve Station Service Transformer Power (Calculated Valve)

/

• APPENDIX E ERP PHASE I DATA SET AVAILABL FROM PROCESS COMPUTER PARAMETER INSTRL"fENT

1. Reactor Pressure 1B21*PT004A

2. Reactor Pressure 1B21*PT00LB

3. Reactor Utr Lev (WR) 1B21* LIT 004A

4. Reactor Wtr Lev (WP.) 1B21* LIT 0043

5. Reactor Wct Lev (FZ) IB21* LIT 007A

6. Reactor Wtr Lev (FZ) 1E21* LIT 0073

7. ADS /SRV Tailpipe Press IB21*PT153A

8. AD3/SRV Tailpipe Press IB21*PT153B

9. ADS /SRV Tailpipe Press IB21*PT153C

10. ADS /SRV Tailpipe Press 1B21*PT153D .

11. ADS /SRV Tailpipe Press 1B21*PT153E

12. ADS /SRV Tailpipe Press 1B21*PT153F

13. ADS /SRV Tailpipe Press 1B21*PT353G

14. ADS /SRV Tailpipe Press 1B21*PT153H

15. ADS /SRV Tailpipo Press IB21*PT153J

16. ADS /SRV Tailpipe Press 1B21*PT153K

17. ADS /SRV Tailpipe Press 1E21*PT153L

18. RCIC Pump Disch Flow 1E51*FT003

19. RHR Sys A Flow -

1 Ell *FT001A

20. RHR Sys B Flow 1E11*Fr0015

21. RHR EX A Outlet Temp 1E11*TE012A

22. RHR HX B Outlet Temp 1E11 ATE 012B

23. RHR HX A Inlet Temp 1 Ell *IE011A

24. 'RHR RX B Inlet Temp 1E11*1E0llB

25. HPCI Pump Disch Flow 1E41*FT003

26. Core Spray Sys A Flow '

1E21*IT002A

27. Core Sptay Sys B Flow 1E21*FT002B

28. RHR HX A - SW Outlet Temp lEll*:E013A

29. RER HX B - SW Outlet Temp IE11*!E013B

30. RHR Svce Wtr A Flow 1 Ell *FT006A 5 1 Ell *FT0063

31. RHR Svce Wtr B Flow -

32. Reactor Bldg. Flood Level 1G11*LTS645A

33. Reactor Bldg. Flood Level 1G11*LTS645h

34. Drywell Pressure 1293*PI003A

35. Drywell Pressure 1Z93*PT003B

36. Suppression Chamber Press 1Z93*Pr004A

37. Suppression Chamber Press 1Z93*PI004B

38. Suppression Pool Wer Temp (1 ft) 1293*!E110Z

39. Suppression Pool Wtr Temp (1 ft) 1Z93*TElllW

40. Suppression Pool Utr Temp (1 ft) 1Z93*TEll2Y Suppression Pool Wtr Temp (1 ft)

~

41. 1Z93*7 Ell 3X t

• a x

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

e%

• PARAMETER 1NSTRUMENT

42. Suppression Pool Wtr Temp (2 ft) IZ93*TE132A

43. Suppression Pool Wtr Temp (2 ft) 1Z93*TE133B

44. Suppression Pool Wtr Temp (2 ft) IZ93*TE134A

45. Suporession Pool Wtr Temo (2 ft) lI93*TE135B

. 46. Suporession Pool Utr Level 1293*LT001A

47. Suporession Pool Wtr Level 1293*LT001B 4 s' . Drywell Hydrozen Conc. 1748*I2Z115A

49. Drywell Hydrogen Conc. 1T48*I2Z115L

50. Suporession Chamber H2 Conc 1743 *I2Z116A

51. Suporession Chamber H,'

Conc IT4J*E2Zll6B

52. Drvwell Oxycen Conc 1748*G2Z123A

53. Drvwell Oxveen Cone IT48*C2Z123B

54. Suooression Chamber 02 Conc IT48*02Z124A

55. Suporession Chamber 02 Conc IT48*C2Z12/ 3

56. Reactor Bldg. Press. 1T41-FDT0ll

57. ADS /SRV Air Hdr. A Press IP50*Fr116A

58. ADS /SRV Air Hdr. B Press IP50*Fr116B 39.. Drywell Temoerature 1T47TE02,7.A 60.

61. j( jg 62.

63.

64.

65.

66. *

67. Y )

68.

69. Drywell Temperature 1T47TE027L

70. RBCLCW HX A outlet Temp 1P42-TE001A

71. RBCLCW EX B Outlet Temp IP42-12001B

72. Circ Wtr Pmp A Disch Press 1N71-Ir083A

73. Cire Wtr Pop B Disch Press IN71-FIO83B

74. Circ Wtr Pmp C Disch Press IN71-Fr083C

75. Cire Wtr Pmp D Disch Press. IN71-Fr083D

76. Main Condenser Pressure IN21-PI005A

77. Main Condenser Pressure 1W21-F:005B

78. Condensate Storage Ik. Level IPil-IT002

, 79. Feedwater Temperature 1B21-Tr001A

80. Feedwater Temperature IB21-TIO01B

81. Feedwater Temperature 1B21-T!001C

82. Feedwater Temperature IB21-TIO01D

83. Feedwater Flow 1C32-F?001A

84. Feedwater Flow . IC32-F:001B

85. Neutron Flux Level . APKM A

86. Neutron Flux Level APRM B

87. Neutron Flux Level APKM C

88. Neutron Flux Level AFTG( D

89. Neutron Flux Level APRM F.

90. Neutron Flux Level APRM F

91. Neutron Flux Level TIP A

92. Neutron Flux Level TIP B

93. Neutron Flux Level TIP C

94. Neutron Flux Level TIP D

95. Control Rod Position (Core Mdp Graphic Display) .

L

' ' APPENDIX F

• Radioloc.ical and Meteorological Data Base .

y-1is00.02-7t>8c 12/13/s2 C33 *

~ f.IS Page i of 5 0.Is *

• DATA 8.89 CilAfedEL IR DESCRIPTION M WANGE ALAmel LIMITS TEGif SPEC LImIM t.20 f.23 i

t. Amot Rest Pumps wsf (e) . s- sooosmEas/HR 9 stem /im s.2s -

2. Assoa es>C iumeINE f(G) .f-LOO 0am Etea m 1.2T SamEN/Ist.

3. Amo3 mm PutePs Easi tie) . s- sooosmEas/HR Sa:REas/em t.re ,'

4. Asso4 mm suAT Exc usf.  :(e) . -tooosentm/ tat SaeREm/im s.as

s. Asso5 anst suAT ExC EAsi 9(G) . 8-lOOOesREse/Im SemEst/lut 5.30

= s. Assos ' PaV AREA Wsf 1(1) 8- SO'esREse/im I.31

7. AetoF PSV AREA EAsi f(e) 1- tO'aIREes/HR I . 3;r *

s. AasPs PERSONNEL st4TCH 1(e) .1- 90000eREal/tti SeeREN/Het t.33

e. Assoe tousP HATCil f(G) . 1- 80000eREst/im SesREel/im t.34 '

to. Amto FUEL Pool CLEdaf UP PURIPs t(e) . I- 8000eRE st/lgt SeeREst/im I.35 e it. Aas e t Rm Ct.EAN UP PUsePs I(G) . 9-looseREst/let t.3s

12. Amt2 FUEL POOL 64 EAT ExC f(e) . 1-tOOOsWEas/ gut 3.37 ,

$3. Am13 CcNTAal EQUIPesENT f(G) . t- to000eREm/im BaeREte/ ten 1.3e .

e4. Ass 4 rutt root tie) . t-toooseREse/im SeeREse/ sui 3.3e ts. Asels MEA lEAT INsul sie t(G) . t- 90000eREm/ist lesREst/lut 1.40 so. Ase se OECow AREA mEst 1(a) . t- sooosattet/tm SesREse/ena s.45

37. Aas t7 CaseENsATE PUsers tie) . t-loo 008 test /im SeeREN/Im . l.42 so. Aase oEcost AREA EAsi 1(G) . 1- SnnremEN/im 9.43 as. Asese HI press TUltetNE l(G) . 1-lOOoseREft/im t.44

20. Ase20 Hogsi AREA 1(G) 1-innnmetas/em

. t.45

23. Amat 'oscose AREA tie) . t- toooseREm/est e.4s .

22. Am22 aum LAS (e) . o t- tooseREm/est sesREes/ sat s.4r

23. Amas taATER eAis f(e) . s- soooseREm/tm SaeREse/ sat s.4e

24. Ase24 oECoN AMEA esDATH f(G) 3-lOOcteREse/tm SIGREN/Sgt

. 1.49 as. Asers masTE ETAP t(e) . t-loooseREm/em SaeREu/ set t.50

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e t .

Y-t1 02-754c I2/tr/02 033

• e*,

4 8 Page 3 of 5 DATA fila 48dE L. ID DESCRIPTION LIg[3 TEQt SPEC E 808I15 RANG $ ALAR 00 LIMI Q .

St. SeET4 Wits DIRECTION 33 t(tdo) 0-540 DEGREES 2.IG .

52. MEIS DELTA TEIIP 33 TO ISO f(T) - tO* F -20* F

• 2.87 53 MEis IEMPERAtuRE AT 33 f(T) -29* F- 126* F I 2.la s4. PMi1A seAIN STEAM LINE A l(G) 1-IO'leREN/HR

• 2.89 ss. PMI es seAIN STEAM LINE S f(G) l-lO'feREN/Ist 2.20

54. Pelt IC MAIN STEAM LINE C f(G) 8- lO'IEREN/let 2.21

57. PMi sO 10AIN STEAM LINE D l(G) 1-lO*80RE91/IIII 2.22 ,

se. PMlas HIGH RANGE Sic.WNI EMH f(A) 1- SO* TIC /CC 2.23

59. P98 827 - PAS West se0NITOR PART tiG) tO-'*-tO-*teC/CC 2.24 y sO. PMl2R PAS \NT IeONITOR CAS 1(G) 2x tO- * -2M SO* 8eCC/CC 2.25

08. PMI2A STEA80 M T AIR EM A f(G) t- lO*flREN/llt 2.26

62. PM128 STEAM JET AIR EJE B f(G) 1- tO*IIRE M/lR 2.2T C3. PMt3 LIG IIADWASTE DI5cilARGE f(G) IX IO- * - t X lO- '88CC/CC 2.2R C4. PMl3 LIG RADWASIE DISCHARGE 4(F) 0-20CGPM 2.29 cs. PM134 Hint RANGE Re5vs f(G) l- tO'asC/CC 2.30 ss. Put4 STEAae JET AIR EJE C f(G) 1- tO'teREN/flR 2.38 C7. PMl74 WEFUEL SEVEL VENT A 8(G) .Ol- SOceeREN/881 2.32 se. PMl7e IIEFUEL LEVEL VENT 5 f(G) .Ol-800MREN/ilR 2.33 ss. PMIS CONT DRV FILT VRAIN EX l(G) IO- * - lO- 8 80C/CC 2.34

70. PN28 REAC10ft SLDG $8 VENT I f(S) 90- * - lO'ISC/CC 2.35

78. PN29 ptACTOR ELDG 58 VENT 1 4(F) 0-4CF06 2.36

72. l'M22 REACT 08t SLDG 55 VENT 2 f(G) IO- * - lO'80C/CC 2.3T

73. PN22 REACTOR BLDG 58 VENT 2 4(F) 0-4 CFM 2.38

74. PN23A Blut IIEAT EXC OUILET A f(G) 90 lO- 880C/CC 2.39 7s. PN23e Star stEAT EXC OUTLET S f(G) 10-*-10-' 00C/CC 2.40

w-

,. y-Il600.02-75ac 82/83/ 2 033 -

e s -

Peee 4 or s DATA coiAque t in IpefX twscairilgeg Retsit At Aast t tilli5 Y$CH SPEC LINIIS

73. em24 aEA SLOG CLO LOOP Waita 8(G) 10-*-SO-'teC/CC 2.48 .

77. PN25A CONT aOOs0 VENT l l(at) 10 SO- 88ec/CC 2.42

73. Pmass CANT soost VENT 2 lle) 90-*-lO**MC/CC 2.43

=

73. Pm264 CONI moest VENT 3 1(G) 10- e . gg- e/aIC/CC 2.44 eO. rmasa, CONY a00se VENT 4 1(G) SO-*-SO-*mc/CC 2.45 st. PN21- CONT a00ss Aile05 PAar 1(G) 10-**-lO-easc/CC 2.46 c2, emas C&eef moons ATseoS eA5 1(a) 80-*-lO-amC/CC 2.47 ,

c3. PNas aEA stoo VENT EX GAS 1(G) 10 SO- 'sec/CC 2.48 "

c4. nea0 . MEA stoG VENT EX PAa! 1(G) lO- * *- lO- *80C/CC 2.48 7 e5. Puit gTATIONVENTEXPAsi 1(G) 80-'*-lO-*8ec/CC 2.50 ,

sG. PM48 ,

STAllON VENT EX PARI 4(F) 0-4CFt8 2.58 C7. 1"M42 STATI0ef VENT EX GAS 8(e) 10-*-80-80eC/CC 2.52 as. PM42 STATION VENT EX GA5 4(F) 0-4Cf88 2.53 am. ew43 SiAllON VENT EX 100 1(N) SO-**-SO-*tec/CC 2.54 sO. PM43 STATION VENT EX 100 4(F) 0-4 CFM

• 2.55

09. Pwa5 mADWASTE STEAN GEN MIG 1(G) 80-*-IO-88ec/CC  !

2.56 ca. em46 WASTE EvaroaAion lie) SO* - SO' Nata /im 2.57 C3. PM47 aEGENEaATE EVAPonATOR f(G) SO* ,10'leatst/les 2.58 c4. PN4a SPENT aESIN taANS T A8et 1(G) SO'- lO'Itaitt/lut 2.52 e5. em4u EVAP acTioses TAeoc lle) SO*-le'anaEN/sen 2.60  ;

~ es . east Ann aEM Pusee DI504ARGE 8(G) 'lO * *- SO- 'acC/CC 3.1 .l C7. ens 2 teof wATEa leEAT SV5 & EXC 8(e) 80-8-lO-*mC/CC

• 3.2 I

sa. essa STEAas SEAL EVAP0a4TOs 1(G) 80-*-10-8 teC/CC 3.3  !

se. Poes4 Tusi stne CLo Loop watra lle) 80- a- SO- a sec/CC 3.4 800. eess5 aAo stoa VENT EX GA5 1(G) 80-*-80-* sec/CC 3.s 1

e

. y-3If,00.C2-158c 92/13/02 033 je s l*

w. ,

Pogs 5 et E , .

- I DATA CluMNEL ID DE SCRIfilpet llegl M ALAIMI LIMI15 TECH SPEC LIMIIS -

80 8. PN66 RAD SLOG VENT EX PART 1(G) IO.se.80-800C/CC 3.6 102. Pn67 fue SLOG WENT Ex gas 1(G) 10 8-lO-'mC/CC 3.7 [

~

103. f(G)

PeeSS TUR BLDG WENS EK PART 10-**-lO-'teC/CC 3.s fi.

IS4. PeeG I CONI DRVWELL PA8f 8(G) IO* * *-lO * 'acC/CC 3.9 SOS, PNG2 CONT DRVWELt. GAS f(G) IO- * -lO- e nC/r4 3.30 i

806. PesssA otFGAs sVs Disal A 8(G) .Ol-loosmEN/sm 3.88 107. Pae6be GFFG45 SVS DISCH S 1(G) .Ol=100tIREN/NR 3.12 s los. Puss REACTOR stDG PAes 1(G) 10 90' sec/CC 3.83 ,

toe. Pa6s REACioR StDG PAm 4(F) (L18) 3.14 Y Iso. Poss? TuRSINE BLOG PAN 8(G) 10-8-lO' esc /CC 3.95 ,

'iei. Poes? TuRSINE SLDG PAtt= 4(F) (LIR) 3.96 ll2. PeeGs RADWASTE SiDG PAN l(G) 10-8-lO'MC/CC 3.87 il3. eseca RADWASTE SLOG Pm 4(r) (LTR) 3.la 184. PeeGS ' STATION VENT PAN f(G) 10- s - lO'8eC/CC 3.le  !

115. PeeG9 siATicN VENT PAM 4(F) (LTR) 3.20 186. PN73 WASTE DEWATERIWS TA8e( f(G) . 8-lO*80RE88/im 3.29 887. PN74A' SOLID WASIE CASM A 1(G) IO- 00'80REld/im 3.22 '

i

-lit. PM745 SOLID WASTE CASE ( 3 f(G) lO-lO'leREte/iB 3.23 ,

lle. PN74C SOLIO WASTE CASK C f(G) 10-lO'DeREte/4R 3.24

)

120. PM74D SOLID WASIE CASK D f(G) lO-lO'8 mete /les 3.26 ii 128. PM77 RADWASTE Tate ( WENT $(G) 80-8-lO-'sec/CC 3.2G 822. Puis sal.1WAIER DRAIN TANK f(G) 80- * - lO- 8 00C/CC 3.27 e

_ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ --w,- w - _ ___ . _ _ _ _ . _ _ _ _ _ _ _

6' .

k

. .:t/47L ,

APPENDIX G

[f '

g 00/20 S

-t ERF. Phase ~I Analog Display Formats *'

5.P.D.S. MAIN FCl4 - C' tart:!C !*J' 10 ' -TIP.E-y? ' . -

x:c::3: 4.

t: .

l -

(

I REACTIVITY CONTROL 6' .

.l 'O NO. 11 - ATPtt LEVELS ,

U

- NO. It - C0ffrROL ROD FOSITICNS A C

~~

REACTOR CORE CCOLA:;T / HEAT REtb, OVAL

• g 0 NO.13"- RX OCrt! PRESS. W/FZ LEVELS. REF. LEG DM TEMP.

, O NO.14 - CCRE GFRAY FLO!!, EU:'? FCOL TEMPE:fATU"E E' -

S

. REACTOR C00LAtfr SY3 TEM INTEMITY - 0 0 NO. 15 - RX CC't: PRESS. C*A PRESS. SUPP POOL ftE55 h k- ~O i 3.16 - CM AVG TE!!P, ItlP? FOGL MATE"i LEVELS (~

O 'NO. 17 - ADS /IRV PC31 TIC 13. C /TZ LEVELS fI > $

C0itTAItttENT INTEC'<ITY / HEAT ret *0 VAL g O NO.18 - RX BLCG FLCCO LEVEL. PX BL3G CELTA T2E33.

.. - ' DM PRESS, SUSP l 00L' FRESS, E'/ AVG TErp.

I 0 NO.19 - RX DC':2 rRESS, S'JP? 000L LEVEL. Ot.PP PCOL TEMP.

O NO. 20 - DM / SUPP FCOL H2 CC :CE ITRATIC'8 g

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.A B- A D A D A D L940 L941 L943 L944 ESO3 E501 L993 L994

• XXXX XXXX XXXX XXXX XXXX XXXX XXXX XXXX ,

O i

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L943 REACTOR LEVEL A WR INCH L944 REACTOR LEVEL B WR It:CH L994 DW T EL 145 AZ 233 CEGT 3 6 4

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I . .w. . . fS U%b AW* .WMVM G Y. . . .'.'. .. . . .. . . . . . . . . . . . . . . ... . . . . . . . . . . . . . . . . . . . . .. *

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• INCH INCH INCH INCH C407 TIuE

• LO41 L944 F303 E50I 8000 0001 B002 R001 8004 8005 w.

4

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• PRE 9S PRE 9S A PRF99 PRC99

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• PSIG P91G PSIG P9IG PSIG ESIS E520 ES22~-f.947 L949 F527 E529 G703 f.9 74 L9 77

,. r T i wP --* - L940 - L941 - E507- E5 il - E514-m 7~ -25.2 -25.6 -25.7 -25.4 -30.I 1,1 M 29 -29.9 -27.9 89.9

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• COOLANT FLOWS

• Cn0LANT TEMPER 4TUR

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+

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• DISCH ' DISCH ~ FLOW- FLOW A TEMP TEMP , 1

• ~

~

• GPW GPW GPM GPW GPW GPM DEG.F DEG.F DEG.F 350.F GPM GPM DEG.F DEG.F G704 G713 G709 G709 G714- G718 G701 G705 G702 G706 G719 F652 G7to G7tl TIME ^

• 991.9 938.3 24.4 70.1 71.6 74.9 9906.4 3495.1 61.1 58.1 114429 44.0 -660.0-1015.5 -

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ELFV ELEV ELrV ELEV FLEV ELEV ELEV ELFV FLEV ELEV

• WIDE WIDE ELEV -

• HANGE RANGE 68 FT 68 FT 90 FT 91 FT 83 FT 83 FT 102 FT 102 FT 13? FT 11) FT 162 FT

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• PSIG PS17 DEG.F DFG.F OFG.F OEG.F DEG.F DFG.F DEG.F DEG.F OEG.F DFq.F DFG.F F602 f.825 f.918 I956 f.957 f.096 L997 1.093 LO94 G562 TIME

• F655 F656 f.777

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• 1 ******************************************************************************************.**.****.*******

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• R ANGE . R ANGE LEVFL LEVEL I FOOT I FOOT I Ft0T I fry)T 2 FmT 2 Ftv)T 2 cent 2 F'Y)T i ~~'* A R A 9 DEPTH ' DEPTH DEPTH DEPTH OFPTH DFPTH 9CPTH OFPTH

g. *

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• H2 H2 02 02 H- H2 n2 02 FLrYt) FLn00 PRE 9s g

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• CONC C097 CONC CONC Cf. ..C CONC CO'lO CONC 1EVEL t.EVEL

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• * * -t T T T t t INCH INCH IN H2O V TIC

• L929 ~~ = L938 ~ 'E503 ~ E504 " L939 ~'E507' E505' LO73 ~~~ F651 FM54 G715' t

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• 7 **+******************************************************************************e************************

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• 0

• '~~**********************************************************************************************************

%

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• DEG.F DEG.F FEET ULB/HR UL9/HR DEG.F DEG.F DEG.F DRG.F VAC VaC P91G P914 PSIG P9IG

." TIME- * .L851 L852- F591' B022 8023 8050 8051 R052 005.1 F507 F508 LRI1 L9td LRIS L816 G

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