PLA-1827, Forwards Response to Generic Ltr 83-28, Required Actions Based on Generic Implications of Salem ATWS Events. Rept Will Be Revised to Reflect BWR Owners Group or INPO-NUTAC Responses
| ML18040B032 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 11/04/1983 |
| From: | Curtis N PENNSYLVANIA POWER & LIGHT CO. |
| To: | Eisenhut D Office of Nuclear Reactor Regulation |
| References | |
| REF-GTECI-A-09, REF-GTECI-SY, TASK-A-09, TASK-A-9, TASK-OR GL-83-28, PLA-1827, NUDOCS 8311070331 | |
| Download: ML18040B032 (66) | |
Text
e P'EGULATO 1tlRMATION DISTRISUTIONS1 (RIDS)
ACCESSION NBR: 8311070331 DOC G DATE e 83/11/04 NOTARIZED! YES DOCKET FRAIL:Stl 387 Susquehanna Steam Electric Stations Uni,'t 1< Pennsylva 05000387 50 388 Susquehanna Steam Electric Stations Unit 2s Pennsylva 05000388 AUTH BYNAME AUTHOR AFFILIATION CURTISrN ~
WE Pennsy 1 venia Power II Light Co.
REC'IP NAME RECIPIE.NT AFFILIATION EISENHUTpDGGG Division of Licensing
SUBJECT:
Forwards response to Generic Ltr 83 28i,"Required Actions'ased on Genetic Implications of Salem ATWS Events>"
Rept will be revised to reflect BWR Owners Group or INPO NUTAC eespenses, ~
DISTRIBUTION 'CODE:
B003S COP IES RECE VEDi LTR JENCL L.
SIZE:
TrTLf: I.icensing Submittal: Anticipated Transients Without
'Scram'(ATNS)
NOTES: 1cy NMSS/FCAF/PM, 1cy NMSS/FCAF/PM, LPDR 2cys
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LPDR 2cys
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05000387 05000388 RECIPIENT IO CODE/NAME NRA LB2 BC 05 INTERNAI.: ELD/HDSO NRR/DHFS DEPY08 NRA/DSI/ADCPS06 NRA/DSI/CPB 07 NRA/DST/GIB 09 COPIES L,TTR ENCL 1
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1 1
1 1
1 RECIPIENT ID CODE/NAME PERCHgR
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01 NRR THADANIg A13 NRR/DL DIR NRR/DS I/AEB NRR/DSI/ICSB 10 G
F IL 00 COPIES LTTR ENCL 1
1 1
1 i
1 EXTERNAL: ACRS NRC PDR NTIS NOTES'2 5
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LPDR NSIC 03 06 2
2 1
1 TOTAL'UMBER OF COPIES REUUIREDe LTTR 25 ENCL 24
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eO Pennsylvania Power 8 Light Company Two North Ninth Street
~ Allentown, PA 18101
~ 215 i 770.5151 Norman W. Curtis Vice President-Engineering 6 Construction-Nuclear 21 5/770-7501 NOH 04 883 Mr. Darrell G. Eisenhut, Director Division of Licensing U.S. Nuclear Regulatory ~ssion Washington, D.C.
20555 SUSQUEHANNA STEM EIZCTRIC STATION
RESPONSE
TO GiZKRIC L1'TTER 83-28 ER 100450/100508 FILE 841-2 PLA-1827 Docket Nos.
50-387 50-388
Dear Mr. Eisenhut:
The attached report is in response to Generic Letter 83-28, ~red Actions Based on Generic Implications of Salem ATNS Events.
Our previous letter, PLA-1815 (September 6, 1983), identified several areas that willbe addressed generically by either the BWR Owners Group or the INPO-NUTAC.
The attached report will be revised to reference those efforts when they are complete.
We are available to meet with you and your staff to discuss our response and our planned changes.
Very truly ours, N. W. Curtis Vice President-Engineering t'onstruction-Nuclear I
Attachment cc: R. L. Perch USNRC G. Rhoads USNRC
~y (I)
@sf f070331 83' 04 PDR ADDCK 05000387
~ ~
~ ~
AFFIDAVIT CCNNMWEALTH OF PENNSYLVANIA)
SS COUNTY OF LEHIGH
)
I, NORMAN W. CURTIS, being duly sworn according to law, state that I am Vice President, Engineering 6 Construction-Nuclear of Pennsylvania Power 6 Light Company and that the facts set forth on the attached response by Applicants to Generic Letter 83-28 are true and correct to the best of my knowledge, information and belief.
N W. Curtis Vice President, Engineering 6 Construction-Nuclear Sworn to and s+yribed before me this 9 day of November, 1983
-", '"Notary Public MARTHA..C.;BA'RTONotaryPublic Allentown, Lehigh County, Pa.
Nygommlsslon Expires Jan. 13, 1986
Pennsylvania Power 6 Light Ccmpany Susquehanna Steam Electric Station Units 1 6
2 I
Response to Generic Letter 83-28 November 6, 1984 Docket~~o
~
Cnntrol4'~~I~~+~~
Dw o+8j Qio~of Document:
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Section 1:
Post-Tri Review 1.1 Pr am Descri tion and Procedure PAL has developed a systematic program for ensuring that unscheduled reactor scrams are thoroughly evaluated.
This program is implemented by plant procedures.
PPGL's Nuclear Safety Assessment Group (NSAG) recently completed an independent review of the scrams to date at SSES.
NSAG attempted to determine if each trip was analyzed in sufficient detail to justify the conclusion that the reactor could be safely restarted.
NSAG concluded that in each case the decision to restart the reactor was justified That is, sufficient facts were in hand and sufficient analysis had been done to ensure safety.
The following respond to the respective requirements of Generic Letter 83-28 section 1..
Susquehanna SES is in full compliance with the NRC position and no changes are planned.
The general criterion for determining the acceptability for restart is:
The plant shall be able to be restarted and operated safely.
To implement this general criterion, we impose the following specific criteria:
a)
The trip parameter that scr~ the reactor shall be determined.
b)
The system or component that is the source of the abnormality causing the trip shall be determined and corrective actions accomplished, as required:
1)
Zf the suspect system or canponent is safety-related, the cause of the malfunction shall be identified and corrective actions completed prior to restart.
2) lf the suspect system or component is not safety-related, the plant may be restarted ifit is confirmed that the plant is within technical specifications..
c)
. The correct functioning of safety-related systems and ccmponents during the event shall be verified by:
1)
Verifying the proper sequence of safety-related system/component actuation frcm detection/initiation to shutdown.
Section 1:
Post-Tri Review 2)
Camparing the event information with known or expected behavior.
The responsibilities and authorities of personnel who will perform the review and analysis of these events are as follaws:
a)
The Superintendent of Plant or Duty Y~ger is responsible for authorizing the Shift Supervisor to re-start the plant follawing a scram.
b)
Section Heads are responsible for resolving items identified in the post scram critiques.
They shall ensure resolution of items which are identified as "required for unit startup" are dispositioned prior to Duty Manager authorization of a plant startup.
c)
The Shift Technical Advisor is responsible to provide post scram data and preliminary evaluations of plant performance including whether equipment functioned as designed.
d)
The Supervisor, Scram Investigations is responsible for supervising and coordinating the investigation and inmediate corrective actions following reactor scrams.
e)
The Senior Ccmpliance Engineer is responsible to verify appropriate notifications were made within the designated times based on appropriate incident classification by on shift personnel.
The necessary qualifications and training for the responsible personnel are:
a)
Superintendent of Plant required to meet the criteria of section 4 of ANSI/ANS-3.1-1978 for Plant Manager.
b)
Duty Manager required to meet the criteria of section 4.2 of ANSI/ANS-3.1-1978 for managers.
c)
Section Heads-Senior Plant Staff personnel who as a miniiwm, meet the requirenents of ANS-3.1-1978 for their respective positions.
d)
STA - Minimum qualifications:
B.S. or equivalent in Engineering or Related Sciences.
Four years pawer plant experience of which three or a+re are nuclear pawer plant experience.
Section 1:
Post-Tri Review Meet. requirements for personnel whose duties require the use of self-contained respirators Abilityto meet requirements for access to vital areas of Susquehanna SES.
Abilityto successfully complete the required training and certification which includes:
STA Systems Training -'onsisting of lectures, quizzes and examinations covering system design
- bases, flow paths, camponents, and instrumentation and controls.
STA Science consisting of reactor theory, thermodynamics g heat transfers fluid flowg BWR chemistry, reactor plant materials, radiation protection, electrical science, and instrumentation.
STA Simulator consisting of technical specifications, operating procedures, emergency procedures, and transient analysis.
Mitigating core damage training.
Bnergency response team training Management Training consisting of problem
- analysis, decision naking, and camnunication.
e)
Supervisor, Scram Investigations These are senior plant staff personnel who as a minimum, meet the reouirements of ANS 3.1-1978 for their respective positions and is appointed by the plant superintendent.
f)
Senior Compliance Engineer Required to meet the criteria similar to Section 4.4 of ANS-3.1-1978.
The normal sources of plant information necessary to conduct the review and analysis are the General Electric Transient Analysis and Recording System (GENES), the plant process
- computer, and chart recorders located in the control roam and throughout the plant.
Section 1.2 "PostWrip Review-Data and Information Capability" contains the detailed capabilities of these sources.
When applicable, the plant performance is evaluated against the transients in the FSAR.
In addition, the Post Transient/Reactor Scram Evaluation procedure has a safety systems actuations and
Section 1:
Post-Tri Review performance table which is used to compare actual to expected plant behavior.
Plant Operating Review Conmittee meetings are conducted to review all scrams fram critical and recmmend startup to the Plant Superintendent.
A technical Section instruction, SCRM Report Data Collection, provides guidelines for the proper retrieval of data for analysis and evaluation.
This data is stored in the Don@rent Control Center as part of the permanent plant records.
Section 1:
Post-Tri Review 1.2 Data and Information Ca ilit Safety-related systems provide actions necessary to assure safe shutdown, to protect the integrity of radioactive material barriers, and/or to prevent the release of radioactive material in excess. of allowable dose limits (SSES FSAR 7.l.la.3.1).
Recording, recalling and displaying data
'to permit diagnosing the causes of reactor shutdowns and for ascertaining the functioning of safety-related equipment are clearly not safety-related functions.
Acquiring and storing this data is necessary for the proper and efficient operation of a nuclear purer plant. As such, SSES is equipped with systems which allow for the acquisition, processing, recording and display of data for timely diagnosis and co~ion of conditions rendering the plant inoperable.
These systems are more than adequate to perform the functions for which they were intended.
The following respond to the respective requirements of generic letter 83-28 section 1.2.
Ca ilit for assessin s
ence of events The primary mechanism for assessing the sequence of events prior to and during an unscheduled reactor shutdown is the plant computer.
An overview of the SSES plant ccmputer is in FSAR
'ection 7.7.1.7.
The computer is equipped to produce two post-trip review logs and a sequence of events log (SOE).
Taken
- together, these logs allow for the detailed assessment of the sequence of events just prior to and during an unplanned reactor shutdown.
The plant cczputer is progranaed to produce both a nuclear steam supply system (NSSS) post-trip review log and a balance of plant (BOP) post-trip review log.
The NSSS post-trip review log is automatically activated five minutes after a scram. It provides a log of 24 NSSS data point values (listed in Attachment 1) collected at 5 second intervals for the 5 minute period before and after a scram event.
The log is printed on the logging line printer using the format shown in Attachment 2.
The BOP post-trip review log is similar to that for the NSSS except that 48 BOP data points are monitored and logged (see ).
The log provides the data collected at 15-second intervals for the 30 minutes prior to and follmring a trip.
The log. is either automatically initiated 30 minutes after the trip or can be demanded by the operator and is~ on the logging line printer using the same format as in Attachm nt 2.
The SOE log provides a chronological listing of major events occurring during plant operations.
Vpon detection of a status change of any of the preselected sequential events contacts, the
Section '1:
Post-Tri Rev'ew SOE log initiates.
As soon as 64 contact changes have been sensed or 30 seconds have elapsed since the first, detected
- change, the log begins printing on the log line printer in the Control Rocm.
The log continues until all contact changes have been printed including any changes sensed subsequent to log initiation.
The system has the capability of establ'shing an order of occurrence between state changes with a resolution of 4 milliseconds.
Attachment 4 is the list of parameters which are inputs to the SOE log and Attachment 5 is an example of the output format.
All three logs produce hard copy records which are retained in the SSES Records Management System.
The plant computer is pcwered from an uninterruptable non-1E power supply backed up by an engineered safeguard supply (FSAR 8.3.1.8).
An alternative mechanism for assessing the sequence of events in case of a loss of the computer is to take data from the strip chart recorders in the control room'nd in the plant (see next section) and supplenent this data with operation observations.
Ca ilit for assessin the time histo of anal variables needed to determine the cause of unscheduled reactor shutdowns, and the functionin of safet -related The primary mechanisms for assessing the time history of analog variables at SSES are the historical recording capability of the plant con@uter system, control room strip chart recorders and the General Electric Transient Analysis and Recording System (GETARS).
The plant computer has a historical recording function which records certain preselected changes in input signals fram both NSSS and BOP systems.
Changes to the software which controls what data is placed in historical records is under plant administrative control.
Data is temporarily stored on a drum and transferred to magnetic tape for pernenent retention each time the assigned drum area is filled.
Data may later be retrieved from the magnetic tape and analyzed.
Attachment 6 is an exanple output.
Strip chart recorders in the control roam can be used to assess the functioning of safety-related equipment.
With few exceptions, these recorders run at 1 inch per hour.
Strip charts provide a continuous indication of analog param ters.
Section 1:
Post"Tri Review GEARS is composed of a mini-computer, magnetic tape unit, magnetic disk drive unit and a printing/plotting station that are integrated by software.
The system is capable of handling inputs from more than five hundred channels.
These channels may then be monitored and/or recorded..
The channels that are selected for recording are placed in what is called. a "work file" created by utilizing system software.
The number of points in this file and the frequency of scanning is a balance between the scanning rate of the equipment, the storage capability of the equipment, the desired discrimination time between scans and the desired time following an event that is desired.
At the present time we scan approximately 300 channels once every 6.67 milliseconds (see Attachment 8 for a listing of these points).
The recording is triggered by selected points exceeding a
predetermined set point.
The triggers presently in use are shown in Attachment 9.
On triggering, the items selected to be in the work file are recorded for a predetermined tim prior to the trigger until the recording capability is filled.
This recording tim period is dependent on the number of channels in the work file and the selected scan rate.
The information that is stored is in both digital and analog form depending on the specific parameter.
These may be plotted at a later time to aid in accident/transient analysis.
GETARS is powered from an uninterruptable non-1E power source tbwt is capable of being powered fram the plant diesel generators.
This is considered a highly reliable source of pM~r o Attachment 7 is a list of param ters developed for this response which are required to determine if safety-related systems functioned properly to meet the criteria of 1.1.1 above.
Many more parameters, identified in the attachments, are monitored and recorded at SSES than are required to determine the acceptability for restart.
However, during development of Attachment 7 we identified approximately 40 parameters, not currently known to be available in the necessary form, which would be useful for post-trip review.
Ne are continuing to evaluate these parameters and will include the results of that evaluation when this report is revised to include the results of the BWR Owners Group program (see section 2.1).
One additional point should be made with regard to ascertaining the proper functioning of safety-related equipment by post-trip reviews.
- SSES, as do all licensed reactors subject to technical
Section 1:
Post-Tri Review specifications, has a comprehensive surve'llance and in-service inspection program.
The purpose of this program is to assure the continued high reliability and operability of safety-related equipment.
Ascertaining the proper functioning of safety-related equipnent during post-trip reviews is an informative supplenent to the current program.
- However, PP&L continues to rely on the existing surveillance and in-service inspection program to assure the proper functioning of safety-related equipment.
Other data and information rovided to assess the cause of unscheduled reactor shutdowns Other data and information available to assess the cause of unscheduled reactor shutdowns include:
plant instnunentation associated with nonsafety-related systems; data obtained from analysis of samples; data from further engineering analysis; and, in extreme cases, information fram disassembly and examination of equipment.
Together, all of the above information allows us to determine the ultimate cause for rendering the plant inoperable.
Schedule for an lanned chan es to existin data and information ca ilit The results of the evaluation of Attachment 7 parameters referred to in 1.2.2 above willbe included in a revision to this report which will be submitted by February 29; 1984.
Section 2 nt Classification and Vendor Interface 2.1 Reactor Tri S stem nents t Classification All reactor trip system ccxqmnents are identified as safety-related on docm~ts, procedures and information handling systems used in the plant to control safety-related activities, including maintenance,.work
- orders, and parts replacement.
No changes are planned.
Vendor Interface The Susquehanna SES reactor protection (trip) system was supplied by the General Electric Ccmpany.
General Electric is still under contract to PAL for the completion of Unit 2.
This contract provides for current vendor information through contract termination.
As identified in our letter PLA-1815, the BWR Owners'roup is undertaking an effort to address continuing maintenance of vendor information generically.
Additional information will be submitted when the BNRCG program is available.
Also, PAL has an existing Industry Events Review Program (IERP) to review relevant industry information and experience for impact on SSES.
This program is described in section 2.2.2.
2.2 Pr ams for All Safet, -Related nents 2.2.1.
nt Classification PP&L has a comprehensive program for ensuring that safety-related
- systems, components and structures are classified and maintained as such.
The PPGL program is implemented through a thorough set of procedures containing the necessary ingredients to assure that safety-related quality is maintained throughout the life of SSES.
The procedures define the aethod of detexmu~g the classification of equipment, and ~rements for design, aedification, procurement, maintenance,
- handling, storing, and testing safety-related equipment.
SAFETY RELATED is a generic term applied to:
1.
Those systems, structures, and components that meet one or narc of the following requirements:
(a)
Maintain the integrity of the Reactor Coolant System pressure boundary.
(b)
Assure the capability to prevent or mitigate the consequences of accidents that could cause the release of radioactivity in excess of 10CFR100 limits.
Section 2 t Classification and Vendor Interface (c)
Preclude failures which could cause or increase the severity of postulated accidents or could cause undue risk to the health and safety of the public due to the release of radioactive material.
(d)
Provide for safe reactor shutdown and iaxnediate or long term post accident contxol.
2.
Those activities that affect the systems, structures and components discussed in. Item 1 above such as their design, procurement, construction, operation, 'efueling, maintenance, modification and testing.
PPSL maintains a controlled "Q-list" identifying systems, structures and components which have been desi'gnated as safety-related.
The content of this list is based on Table 3.2.2 of the FSAR.'he SSES Q-list does not include every safety-related ccxqmnent of evexy safety-related system.
However, as discussed below, the PAL p~ures for determining unlisted ~nent and part quality are very conservative.
As such, this system ccxnplies fullywith the intent of NRC's position.
The Q-list is one of PP&L's Quality Consideration Lists (QCL).
Quality systems, components and structures are defined as those which appear on any of the QCL.
The QCL are controlled docunents which take precedence over other lists or documents as the source of quality classification.
Development and updating of the QCL are the responsibility of the Nuclear Plant Engineering group.
The initial QCL and requests for changes thereto are required to be reviewed by quality assurance.
PAL procedures for modifications, engineering, maintenance, procurement and material control require a determination if the activity affects a quality system, camponent or structure.
For procurement of spare parts, parts used in safety-related (Q-listed) assemblies or systems are classified as either Ql, Q2, Q1E, Q2E or N as defined below:
"Ql" Items A sub-classification of "Q" - These items are considered essential to the safe operation 'of the plant and are designed and fabricated in accordance with a QA program approved by PAL that is consistent with the pertinent provisions of 10CFR50, Appendix B, National Standards or Codes, as appropriate, and other requirerrents set forth in the purchase order.
The supplier shall provide, as a min+num, doazrentation for all specified requirements and a certificate of compliance certifying that all Ql items comply with the purchase order requirements in order for the p~s to be considered acceptable.
10
Section 2 -
i t Classification and Vendor Interface "Q2" Items A sub-classification of "Q" These items are considered essential to the safe operation of the plant and may be obtained ccxrnercially as catalog/off-the-shelf items (i.e,
they do not recpuxe unique or special engineering specifications; or they are manufactured to national standards or by processes generally automated or highly repetitive; or there is little chance for error during manfacture to affect their safety-related characteristics; or they may be obtained ccxmnercially in accordance with Article IRK-7000 of the ASME Code).
The manufacturer of Q2 items need not have a QA program meeting the requirements of 10CFR50, Appendix B, but should have the usual processes or quality controls normally associated with the production of like items.
"Q1E" A sub-classification of "Q" These Class 1E items are classified as Ql items and are essential to ~ency reactor
- shutdown, containment. isolation, reactor core cooling, and containment and reactor heat removal, or otherwise essential in preventing significant release of radioactive material to the enviroment and are environmentally qualified in accordance with the requirements of NUREG 0588, Category I, and IEEE-323-74 as identified by SSES "Environmental Qualification Report for Class 1E ~ipnent."
The supplier shall provide, as a miniunm, the docum ntation and certificate of campliance consistent with all Ql items.
In addition, for items classified as 1E, the supplier will be mcyired to submit:
l.
An Environmental Qualification Test Plan including test set-up, test procedures, and acceptance criteria for at least one of each type of part.
2.
An Environmental Test Report, with data, tlmt demonstrates parts qualification in accordance with the Environmental Qualification Test Plan.
"Q2E" Items A sub-classification of "Q" These Class 1E items are classified Q2 items and are essential to emergency reactor
- shutdown, containment isolation, reactor core cooling, a'nd containment and reactor heat rental or otherwise essential in preventing significant release of radioactive material to the environs and are environmentally qualified in accordance with the requirements of NUREG 0588, Category I, and IEEE-323-74 as identified by SSES "Environnental Qualification Report for Class lE Equipment."
As a minUaum, the documentation required of the supplier will consist of a certificate of campliance and:
1.
An Environmental Qualification Test Plan, including test set-up, test, procedures and acceptance criteria for at least one of each type of part.
11
Section 2 nt Classification and Vendor Interface 2.
An Environmental Test Report, with data, that demonstrates parts qualification in accordance with the Environmental Qualification Test Plan.
"N" Items Those items not within the scope of the PP&L QA programs Attachment 10 shows the logic followed to classify these parts.
Upon receipt of quality parts, PPGL material control proCedures require that acceptance be documented and that the part be clearly identified by means of a color coded tag.
The tag remains with the part through handling, storage and installation.
For other than procurement of spare parts, parts or ccmponents of safety-related systems are all considered safety-related.
All corrective maintenance and implementation of modifications on all plant structures, systems and components is controlled by a work authorization system.
Before physical work is initiated, the work authorization system requires that the work group foreman classify the work authorization as quality or not and ASME code related or not.
The work activity classification is conservatively based on the system, structure and major equipment level, rather than on the classification of individual canponents.
This is an important and deliberate feature of the system for classifying work activities intended to assWe that work activities which may affect safety-related
- systems, structures, or equipment are appropriately conducted and controlled even though the objective of the work activity may be maintenance, repair, replacement, etc. of a non-safety ccnponent.
The PPSL operational Quality Assurance Program (FSAR 17.2) applies to all safety-related SSES structures,
- systems, canponents and activites.
This program provides the necessary managenent controls, including periodic reviews and audits, to verify that covered procedures are followed.
The procedures for preparation, validation and routine utilization of the information handling systems described, including the QCL, are subject to the controls of the PPsZ Operational Quality Assurance Program.
Deaenstration that appropriate design verification and qualification testing is specified for procureaent of safety-related ~nents requires consideration of the status of the component being procured.
For procurement of the types of safety-related components acquired during construction, 12
Section 2 't Classification and Vendor Interface demonstration is provided by the Equipment Qualification Document Files which have been audited by the NRC and found acceptable (NUREG-0776, Supplement 3, p3-7).
Design veri ication and qualification testing is not required for procurenent of replaceaezt parts and components identical to or equivalent to items already qualified.
For procurement of replacement parts/components for which an alternative to an identical or equivalent is selected, appropriate design verification and qualification testing is required.
Procurement specifications for such equ'pment include design control requirements which comply with ANSI N.45.2.11-1974, including generation and approval of inputs and verification by independent review.
Attachment 11 is taken fram a specification for new containment structure electrical cable penetrations prepared by our architect engineer.
This specification includes requirements for qualification to harsh environmental conditions and appropriate qualification testing.
To date, PP&L has not had occasion to directly procure equipment ~ring environmental qualification for a harsh environment.
As an example of a direct PPSL procureaent, attachment 12 is taken fram a procurement specification for an additional diesel generator.
This specification includes designation of mild environment conditions appropriate for this component and requires a certification of compliance.
Also, dynamic load qualification requirements are included in this example.
With respect to the equipment classification program in use at SSES for structures, systems and components important to safety, we are participating in the UtilitySafety 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 components important to safety constitute a broader class than the safety related set.
We nevertheless believe that non safety related plant structures, systems and ccxqmnents have been designed, and are maintained, in a manner cmxnensurate with their importance to the safety and operation of the plant.
The procedures and specifications described above form a system which controls activities affecting safety-related ccxqmnents and which camplies with NRC's position.
No changes are planned in this area.
13
I ra
Section 2 nt Classification and Vendor Interface 2.2.2.
Vendor Interface PP6L has an existing Industry Events Review Program to review relevant industry information and experience for impact on SSES.
Th's program provides for the periodic review of:
INPO Significant Operating Experience Reports (SOER).
General Electric Service Information Letters (SIL) and Technical Information Letters (TIL).
INPO Significant Event Reports (SER)
SSES Incident and Event Reports Other industry information such as Atomic Energy Clearing
The implementation of the action ite~ identified by this program incorporates the lessons learned frcm industry experience in PP&L practices.
The original set of SSES Installation and ~rating Manuals (ICM's) were obtained and reviewed in accordance with administrative and engineering procedures implementing the A/E's Quality Assurance Program which was reviewed and approved by PPGL and subject to audits and management reviews.
Revisions to the ICN's were controlled by these sane procedures up to the time of turnover of control to PPGL.
New ICY's, or revisions to ICN's are directed to the Docum nt Contxol Center (DCC) at SSES.
DCC does a preliminary administrative review.
A technical review is also performed by individuals which neet the following requirements:
a.
Have a minimum of five (5) years experience in the appropriate discipline, e.g., Instomentation and Controls, Chemistry and Radiochemistzy, Radiation Protection, Electrical Ecpu.pment, and Mechanical Eauipment.
b.
One of the. five years shall be nuclear power plant experience in the appropriate discipline.
c.
Two of the five years should be related technical training.
d.
A ma~ of four of the five years may be fulfilledby related technical or academic training.
14
Section 2 t Classification and Vendor Interface After a review is conducted to determine that the ICN is appropriate for the applicable equipment and approved by the technical reviewer's supervision, the ICN or ICN revision is
- issued, as follows:
DCC logs review as complete and files original review sheet,.
Appropriate number of copies of the ICN Review Form are made e
A copy of the review form shall be inserted in front of the applicable ICN(s).
This certifies the revision, manual or correction as approved.
DCC makes the appropriate controlled distribution.
DCC updates the Index of Approved ICN's by entering the appropriate manual identif'cation and date of approval into the index.
ICN's are controlled documents and may be changed by issuing a properly reviewed and approved Drawing Change Notice.
As identified in our letter PLA-1815, a Nuclear UtilityTask Action ~ttee (NUTAC) has been formed under the auspices of ZTPO to address this issue generically.
Additional information will be submitted when the NUTAC program is available.
15
Section 3 Post-Maintenance Te tin 3.1 Reactor Tri S stem nents A review was made of test and maintenance procedures and Technical Specifications to assure that post-maintenance operability testing of safety related ccmponents in the reactor trip system is performed as required to denunstrate the equipment is capable of performing its safety function.
No problems were identified.
A Maintenance Planners Guide was developed to provide'larification for the planners in assuring that all post-reintenance testing required by Technical Specifications is identified.
In addition to the testing required by Tech Specs, the maintenance work groups are responsible for reccmending operational testing on the Equipment Release Form (ERF) for a particular job.
In many cases this testing is fulfilledby performing the applicable surveillance test for a system or ccaqmnent.
Shift supervision makes the final determination on the operational testing required to restore the'ystem or component to operable status.
We conclude that this system ccmplies with NRC's position and plan no changes in this area.
Appropriate vendor and engineering recaanendations have been incorporated in SSES test and maintenance procedures and Tech Specs for the reactor trip system.
The process for ensuring that reconxnendations are incorporated is the same as for all other safety-related components described in section 3.2.2.
To date, we have identified no instances where post-maintenance testing required by Tech Specs degrades rather than enhances safety.
- However, we intend to continue to pursue this subject as addressed under item 3.2.3.
3.2 All Other Safe Related A review of the SSES test and maintenance procedures and Technical Specifications has been completed.
SSES procedures require post-maintenance functional testing and a review for and performance of appropriate operability testing for safety-related caqmnents.
The required testing demonstrates that the equiprent is capable of performing its safety functions before being returned to. service.
Based on our review, we are in full compliance with the NRC position.
No changes are. planned.
3.2.2 Appropriate vendor and engineering reccxaaendations have been reviewed for incorporation into SSES test and maintenance 16
Section 3 Post-Maintenance Testin procedures and Tech Specs for all safety-related equipment.
Incorporation is accxzrplished by appropriate handling of vendor supplied Installation and 9gerating Manual (ICH) information and vendor bulletins such as described in section 2.2 above.
The plant work authorization system includes identification of required instructions and procedures in planning of work
~ activities.
Each work authorization includes instructions written specifically for it.
Appropriate portions of ICN's are incorporated verbatim, paraphrased, or by reference as appropriate.
The work authorization is subject to review and approval before use.
To date, we have identified no instances where post-maintenance testing required by Technical Specifications degrades rather than enhances safety.
However, we share a BWR industry wide concern that the current frequency of surveillance tests required by sane technical specifications may degrade rather than enhance safety.
We have joined with other industry members in the formation of a Technical Specification Review Camnittee as part of the BWR Owners Group.
We intend to pursue the revision of tech spec requirements which may be shown to degrade safety through this axmittee.
One example of a tech spec requirement that has the potential to degrade rather than enhance safety is the ICO Action Statements for tech spec 3.8.1q AC Power Systems.
The Action Statements require the ope'rable diesels to be started within 2-4 hours of entering the ICO's and once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter.
The continual starting and shutting down of the diesels has a potential to reduce their reliability by continually subjecting them to transient conditions.
We offer this only as an example.
We will pursue justified revisions to tech spec requirements through the BWR Owners Group.
17
Section 4 Reactor Tri System Reliabilit rovements 4.1 Vendor Related Modification Not applicable to SSES 4.2 Preventive Maintenance and Surveillance Pr am for Reactor Tri Breakers Not applicable to SSES 4.3 Automatic Actuation of Shunt Tri Attachment for Westin house and BGW P1Bll'tS Not applicable to SSES 4.4.
rovements in Maintenance and Test Procedures for B&W Plants Not applicable to SSES 4.5 S stem Functional Testin The SSES reactor trip (reactor protection) 'system includes the motor~enerator power supplies,
- sensors, relays, bypass circuitry, and switches that cause rapid insertion of control rods (scram) to shutdown the reactor (FSAR 7.2.1.1.1).
SSES Tech Specs (NUREG 0931) 4.3.1.1, 4.3.1.2 and 4.3.1.3 require periodic functional tests of each instrumentation channel, functional tests of the logic system and demonstration of the system response time respectively.
Testing of the system response tea includes de~ergization of the scram pilot valves.
Taken together, these tech specs require, and SSES performs, on-line functional testing of the reactor trip system, including independent testing of the diverse trip features.
,On-line functional testing of the backup scram valves is currently not performed at SSES.
As identified in our letter, PLA-1815, the BWR Owners Group will address this issue and the review of tech spec intervals generically.
Additional information will be suhnitted when the Owners Group program is available.
18
Attachment 1
NSSS Post-trip Review Log Data Points Data Point Descri tion MAP03 NFL01 NFL04 NAR01 NAR05 NFP02 NM501 NM503 NM505 NN101 NN103 NN109 NN111 NN113 NN115 NFFOS NFF05 NFF06 NFF07 NFF01 NFF02 NFF03 NFF04 NLT01 Contnm. Press.
A LOCA Range Rx. Narrow Range Lvl. A Rx. Upset Range Lvl.
Mn. Steam Line Rad.
A Offgas Pretreat Rad.
'"A Rx. Press.
Wide Range APRM Flux Pwr. Lvl. A Chan.
APRM Flux Pwr. Lvl. C Chan.
APRM Flux Pwr. Lvl. E Chan.
SRM A Log Count Rate SRM C Log Count Rate IRM A Flux, Reading IRM C Flux Reading IRM E Flux Reading IRM G Flux Reading Mn. Turb. Steam Flow FW Flow A FW Flow B FW Flow C Rx. Steam Flow A Rx. Steam Flow B Rx. Steam Flow C
Rx.
Steam Flow D Rx. Bottom Head Drn. Temp.
UNIT X, PA6K XX OF XX DATE NO/DX/TR TINE 10:30t 10 NSS POST TRIP REVIEW lOG PRE-TR)P DATA PPPPPPPP PPPPPPPP'PPPPPPP PPPPPPPP PPPPPPPP PPPPPPPP PPPPPPPP PPPPPP$ 'P PPPPPPI'P PPPI'PPPP PPPPPPPP PPPPPPPP Ill!
10:20!10 SXXXX.XXASXXXX.XXASXXXX.XXA)XXXX.XXASXXXX.XXA SXXXX.XXA SXXXX.XXA SXXXX.XXA SXXXX.XXA SXXXX.XXA SXXXX.XXA SXXXX.XXA 10:20:15 SXXXX.XXASXXXX.XXASXXXX.XXASXXXX.XXA SXXXX.XXASXXXX.XXA SXXXX.XXA SXXXX.XXA SXXXX.XXA.SXXXX.XXASXXXX.XXA SXXXX.XXA
.10:20:20 IO:20:25 10:20:30 POST-TRII'ATA PPPPPPPP PPPPPPPP PPPPPPPP PPPPPPPP PPPPPPPI'PPPPPPP.
PPPI'PI'PP PPPPPPPP PPPPPPPP I'PPPPPPP PPPPPPPP PPPPPPPP 10;25:10 SXXXX.XXASXXXX.XXASXXXX.XXASXXXX.XXA SXXXX.XXA SXXXX.XXA SXXXX.XXA SXXXX.XXASXXXX.XXASXXXX.XXA SXXXX.XXA SXXXX.XXA IO:25ll5 SXXXX.XXASXXXX.XXA SXXXX.XXASXXXX.XXA SXXXX.XXA SXXXX.XXA SXXXX.XXA SXXXX.XXA SXXXX.XXA SXXXX.XXA SXXXX.XXA SXXXX.XXA IO:25:20 10:25:25 10:25:30 DATA RECORDED fOR 2i 4'OINTS AT 5-SECOND INTERVALS fOR 5 HINUTES NDTF DOP Post Trip Revlnt log ls sane as NSS I.og except, 80P log data Is recorded for 4$ points at 15 second Intervals for 30 alnutes..before and 30 alnutes after trip.
Attachment 3
BOP Post-trip Review Log Data Points Data Point GNE02 GNE03 GNE04 GNJ01 GNU02 GNU03 GST02 TAP03 TCY01 TCY02 TEL01 TEL02 Description Gen.
Phase A-B Voltage Gen.
Phase B-C Voltage Gen.
Phase C-A Voltage Generator Power Gen. Reactive Power Gen. Frequency Gen. Stator Clg. Wtr. Out Turb. Seal Stm. Press.
Turb.
CV Position Turb.
BYP. Vlv. Position MSEP A Drain Tank Level MSEP B Drain Tank Level THV09 THV10 THV11 THV12 TLP01 TLP02 TLP03 TLT15 TLT16 TNP01 TNP02 TNP03 Gen.
BRG.
//9 Vibration Gen.
BRG.
I!10 Vibration Alt. BRG.
811 Vibration Alt. BRG.
//12 Vibration Turb. L-0 Brg.
HDR Press.
Turb. L-0 OPER Oil Press.
Turb.
HYD Fluid Press.
Turb L-0 CLR. in Temp.
Turb L-0 CLR. DSCH. Temp.
Condenser A Press.
Condenser B Press.
Condenser C Press.
TET08 TET09 THS01 THV01 THV02 THV03 THV04 THV05 THV06 THV07 THV08 LPT B Exhaust Hood Temp.
LPT C Exhaust Hood Temp.
Turbine Speed Turb.
BRG.
81 Vibration Turb.
BRG.
I/2 Vibration Turb.
BRG.
I!3 Vibration Turb.
BRG.
/l4 Vibration Turb.
BRG.
85 Vibration Turb.
BRG.
P6 Vibration Turb.
BRG.
87 Vibration Turb.
BRG.
II8 Vibration TET07 LPT A Exhaust Hood Temp.
YBB01 THX02 THX03 THX04 TNL01 FPT01 FPT02 FPT03 FTSOl FTS02 FTS03 GSC02 Sync.
BKR. 230 KV Grid Turb. Diff. Expansion Turb. Shell Expansion Turb. Rotor Expansion Condenser Hotwell Level RFP A Discharge Temp.
RFP B Discharge Temp.
RFP C Discharge Temp.
RFPT A Speed RFPT B Speed RFPT C Speed Gen. Stator CLG.
CNVTY.
Attachvent 4
SEQUEZKE OF EVENTS INPUTS POINT IDENT
'PZ03 CPZ04 CPZ05 CPZ06 EBZ01 EBZ02 EBZ10 EBZll EBZ20 EBZ26 EBZ51 EBZ52 EBZ76 EKY02 EKY07 EKY08 ETZ01 ETZ02 EZZ03 EZZ04 ENGLISH DESCRIPTION COND PP A ABNORKK TRIP COND PP B ABNORhSI TRIP COND PP C ABNORMAL TRIP COND PP D ABNORMAL TRIP T BUS OA106-BUS 11A BKR T BUS OA106-BUS 11B BKR BUS 10-T BUS OA106 BKR T BUS OA106-OA107 BKR BUS 20-T BUS OA107 BKR FDR TO SU BUS 10 UNDERVOLT T BUS OA 107-BUS 2A BKR TIE BUS OA107-BUS 2B BKR FDR TO BUS 20 UNDERVOLT SU XFl4IR 10-BUS 10 BKR SU XFMR 20-BUS 20 BKR AUX XH4R ll-BUS 11A BKR AUX XV'1-BUS llB BKR UNIT AUX XFlR DIFF UNIT AUX XFMR GRD DIFF UNIT AUX XFMR PP2BE OC UNIT AUX~ GRD OVOLT
Attachment 4
SEQUENCE OF EVENTS INPUTS FZI01 FZ103 FZ104 FT105 FZ201 FZ203 FT204 FT205 FZ301 FT303 FZ304 FT305 GEZ02 GNZ01 GNZ02 GNZ03 GNZ04 GNZOS GNZ07 GNZ08 GNZ09 ENGLISH DESCRIPTICN RFPT A MASTER TRIP RFPT A VACUUM TRIP RFPT A ACT THR BRG WEAR RFPT A IN'HR BRG RFPT B MASTER TRIP RFPT B VACUUM TRIP RFPT B ACT THR BRG WEAR RFPT B INARCH THR BRG RFPT C MASTl& TRIP RFPT C VACUUM TRIP RFPT C ACT THR BRG NEAR RFPT C INACT THR BRG EXCITER DIF1. ER1WZIAL MAIN GEN DI1UERENTIAL MAIN GEN NEUTRAL OVOLTS MAIN GEN LOSS FIELD A KQN GEN UNDER FREQUENCY UNIT PRI LOCKOUT A/C GEN OUT OF STEP MAIN GEN GRD OVERVOLTAGE MAIN GEÃ NEUT OVOLT STRT
Attachment 4
S@gENCE OF EVENTS INPUTS POINT IDENT GNZ10 GNZll GNZ12 GNZ15 GNZ18 GNZ20 GNZ22 GNZ23 GNZ24 GNZ38 GNZ44 GNZ45 GNZ46 GNZ47 GOZOl GOZ03 NMQ51 NMQ52 NMQ54 NMQ55 ENGLISH DESCRIPTION UNIT DIFFERENZIAL MAIN GEN LOSS FIEU3 B UNIT BKUP LKOUT B-D MAIN GEN NEG SEQEZKZ MAIN GEN VOLT/AERFZ MN GEN BKUP VOLT/HZ BKUP ANTI-MZORING RELAY GEN SPAN PRCT1KTION A GEN SPAN PROHKTION B GEN LOAD UNBALANCE UNIT.1 SYNC BKR LO AIR PRS 230 KV GEN BKR FAILUFZ OVERSPEED PROTEC TRIPPED PRIMARY ANTI-MOTOR RELAY GEN MN SEAL OIL PUMP GEN EMERG SEAL OIL PP UPSC NEUT TRIP APRM A UPSC NEUT TRIP APRM B UPSC NEUT TRIP APRM C UPSC NEUT TRIP APED D UPSC NEUT TRIP APRM E
Attachment 4
SEQUENCE OF EVENTS INPUTS ENGLISH DESCRIPTION NMQ56 NMQ58 A+59 NMQ60 NMQ61 5%@63 NNQ51 NNQ52 NNQ53 NNQ54 NNQ55 NNO56 NNQ57 NNQ58 NPQ51 NPQ52 NPQ53 NPQ54 NPQ55 NPQ56 UPSC NEUT TRIP 'APRM F UPSC THEfM TRIP APRH A UPSC'fKRM TRIP APRM B UPSC THERM TRIP APRM C UPSC TOM TRIP APRM D UPSC THERM TRIP APRM E UPSC THER% TRIP APED F IRN UPSCL TRIP CHAN A IRM UPSCL TRIP ~1 E IRM UPSCL TRIP CHAN C IRM UPSC IBM UPSC TRIP CHAN G TRIP CHPÃ B IRM UPSC TRIP CHM F IRM UPSC TRIP CHAN D IRM UPSC TRIP CHAN H DSCH VOL HI LVL TRIP A DSCH VOL HI LVL TRIP B DSCH VOL HI LVL TRIP A DSCH VOL HI LVL TRIP B MS1V NOZ FL OPEN TRIP Al MSIV NVZ FL OPEN TRIP Bl
Attachment 4
SEQUENCE OF EVENTS INPUTS NPQ57 NPQ58 NPQ59 NPQ60 NPQ61 NPQ62 NPQ63 NPQ64
'NQP65 NPQ66 NPQ67 NPQ68 NPQ69 NPQ70 NPQ71 NPQ72 NPQ73 NPQ74 NPQ75 NPQ76 NPQ77 ENGLISH DESCRIPTION MSIV NOZ FL OPEN TRIP A2 MSIV NOZ FL OP%Ã TRIP B2 PRI CONTN TRIP A PRI CCNTN TRIP B PRI CCNTN TRIP A PRI CCÃZN TRIP B RPV HP TRIP A RPV HP TRIP B RPV HP TRIP A RPV HP TRIP' RPV LOW WZR LVL TRIP A RPV LOW NIR LVL TRIP B RPV LCM NZR LVL TRIP A RPV LCN NZR LVL TRIP B MS LINE HI RAD TRIP A MS LINE HI RAD TRIP B MS LINE HI RAD TRIP A MS LINE HI MD TRIP B NM SYS TRIP A NM SYS TRIP B NM SYS TRIP A
Attachment 4
SEQKRKE OF EVENTS INPUTS NPQ78 NPQ79 NPQ80 NPQ81 NPQ82 NPQ83 hVQ84 NPQ85 NPQ86 NPQ88 NPQ89 NPQ90 NPQ91 NPQ92 TAZ94 TBZ02 TBZ04 TBZ06 TBZ08 TBZ10 ENGLISH DESCRIPTION NM SYS TRIP B MAN SCRAM TRIP A OR C MAN SCRAM TRIP B OR D AVIQ SCRAM TRIP A.OR C AUTO SCRAM TRIP B OR D TURB STOP VLV CLS TRIP A TURB STOP VLV CLS TRIP B TURB STOP VLV CLS TRIP A TURB STOP VLV CLS TRIP B TURB CV FAST CLS TRIP A TURB CV FAST CLS TRIP B TURB CV FAST CLS TRIP A TURB CV FAST CLS TRIP B RECIRC PUM TRIP SYS A TRIP 13 RECIRC PUM TRIP SYS A TRIP 13 VACUUM PUMP TURB BYPASS VLV gl TURB BYPASS VLV 52 TURB BYPASS VLV 53 TURB BYPASS VLV 54
'lURB BYPASS VLV 55
Attachment 4
SEQUENCE OF EVENTS INPUTS
'ICZ06 TCZ08 TCZ10
- TCZ12, TDZ01 TDZ07 TDZ15 TDZ16 TDZ17 TDZ18 TDZ19 TDZ21 mZ22 ENGLISH DESCRIPTION TURB CONTROL VALVE 1 TURB COBOL VALVE 2
'URB CONTROL VALVE 3 TURB CONTROL VALVE 4 TURB KM.'ER TRIP TURB OVERSPEED TRIP TURB BACKUP OSPD TRIP TURB EXH HOOD TEMP TRIP LOSS OF STATOR CLG TRIP TURB SHAFT PP DSCH PRESS TURB THR WEAR OR BRG OIL TURB EHC 125DC PCNER TURB FED PRESS LCM TPZP TDZ23 TDZ24 TDZ35 TDZ36 TDZ40 TDZ51 TDZ52 TDZ55 TURB VACUUM TRIP TURB MANUAL TRIP TURB VIBRATIONTRIP TURB EHC SPD SIGNAL LOST TURB MSEP HIGH LVL TRIP TURB EHC POS VOLTS LOST TURB EHC NEG VOLTS LOST TURB QUILL SHAFT
Attachment 4
SEQUEL OF EVENTS INPUTS WEZ29 avz30 K:Z31 WCZ32 YTZ01 YTZ02 YTZ03 YTZ04 YTZ05 YTZ07 YTZ08 YTZ30 YTZ31 YTZ57 YTZ58 YTZ80 YTZ81 ENGLISH DESCRIPTION CIRC WZR PUMP A CIRC WZR PUMP B CIRC VZR PUMP C CIRC NZR PUMP D MN XFMR LEAD DIFF MN ERR A DIFF MN cZKR A SUDDEN PRESS MN XFMR B DIFF MN XFMR B SUDDEN PRESS SU 328R 10 PRI LKOUT RLY SU XFMR 10 PRI BKUP RLY TRF 10 NZR OPER AIR BRKE SU XFTIR 10 HSGS 1R106 SU ERR 20 PRI LKOUT RLY SU XEHR 20 PRI BKUP RLY TRF 20 MZR OPER AIR BHKE SU XFMR 20 HSGS 2R106
Attachmen 5
UN IT I
PAGE I
'ECUERCE F
EVENTS LDG 96-24-S3 Il-.lb TINE ll-lo'31 584 11:Ie:31 esa 11 -1b 695 11-1c -31 775 EI-I& 31 902 11-1 &-35 -696 II 43-908 II-16-43 975 11 -1h 47 II-I6-50 257 11: Ie:50.258 11-16 -50. 259 I I-Le-50
- 262, 11-.16-.50.262 1 I: L & '. 5'0 26 2 11 16 50 262 ll-lo:50 263 11-Eb-50 263'I 16 50 263 11-16-50 263 PT ID YTZ07 EKYOI Y'TZD8 YTZ31 YTZ,31 YTZ30
~ ESZ26 EBZII E-BZ2 6 NNR55 NH051 NMQ53
."LPQ53 NPQ79 hPQ7~
NP ~259
') P95 1 hPQ& l NPQ61 NPu55 NA:4E SU XFHR IQ PRE LKOUT RLY SU XAlR 10-BUS IO BKR SQ XR4R 10 PRI BKUP RI.Y SU XFHR 10 HSGS IRIDAL SU XFHR 10 RSGS I R106 TRF 10 HTR CPER AIR BRKF FOR TO SU BUS 10 UNDERVO T
BUS QA106-OA107 BKR FBR T13 Slj BUS 10 UNDER VD UP SC TRIP APNEA CHAN E
UP SC TRI?
APISH CHA h C
DSCH VOL HI LVL TRlP A
RAil SCRAA TR I P A
QR C
HS LINE HI RAD TRIP A
PRI CQNTN TRIP A
DSCH VOL HI ~VL TRIP A
PR I CQMTN TR I'P A
HSIV NOT FL OPEN TRIP AI S7 ATUS STARTED TRI PP ED S:IARIEG CLGSEC OPEN YES Zl OSF.:3 YES YES YE" YES YES YES YES YES YES NOT OPEN
Attachment 6
UNIT I y PAGE 1
. 6/14/83 13 39-12 HISTORICAL DATA RETRIEVAL ANO REVIEM SERVICES(MAG TAPE BETMEEN 9 45 QN 6/14/83 ANO 10 0 'QN 6/14/83 ALARM ACTIVITY t CONTACT)
T ICE PQI.NT ID POINT DESCRIPTION STATUS
+6 L 'PQ05 AUTO SCRAM TRIP 4
OR C
OK 9 +b I ".i NPQ06 AUTO SCRAM TRI.P 8
QR D
QK
'ALUE UNETS LRT EXC 9-46 12 NPQ58 l
9 =46 12 NP Q56 9=46-14 NND IO
+6 1<<NN011 46-I 6 TCZ12 9-46 Ib TCZLO 9 46-lb TC108 9-46-'I 6 TC2.'06 9= <<6-16 TDZ 11 9-+6 1S MV?33 9 46 I & TC001 9=<<b 1& TC002.
9 <<6'S TCD03 9=46-18 TCD04 9 =46'& T0001 9
<<6'- 18 T4003 9
46 I 8 T4005 MSIV NOT FL QP'EN TRIP B2*OK NSIV NOT FL OPED TRIP Bl QK PERIOD 8
NEGATIVE PERIOD C
NEGATIVE TURB CONTROL VALVE 4 QK TURB CONTRQL VALVE 3 'K TURB CONTROL VALVE 2 OK TURB CONTROL VALVE 1 OK TURB CV COMMAND CLOSED OK HVAC 0 IV 2 CQMM3N SYSTEM QK TURB CV 51 INTER POSN ALM TURB CV 42 INTER PQSN At M TURB CV N3 INTER PQSN ALA TURB CV N4 INTER POSN ALM TURB TRIP /MODE 5 H RUN/SU ALM SPE 8
STM EXH VLV/RUN/SU ALM NOT OPEN NOT QP N
CLOSED CL3SED
~3'S E D LJ QSED YES TR9LE NQ
<<6 2
FV 74 FMHTR IA SU VENT VLV ALM CL3SED 9=<<6 2')
COMMON SYSTEM QK 9=<<6 22 T EZ06 TURB STM LEAD 3 DRN VLV ALM
<<6-2~
CR001 CQNO RE JECT CV 4
INTER ALA TRBLE QT CL NO
SUSQUEHANNA STEAM ELECTRIC STATION SALEM ATWS EVENT NRC GL 83-28
RESPONSE
ATTACHMENT 7 1.2: Post-Tri Review Variable List Explanation of column entries are as follows:
NAME:
Description of variable or event Egg:
~Priced ie reason for recording.
IE means to determined initiating event SSP means to analyze safety-related system performance.
SOURCES: Devices by which data is recorded, save and/or displayed.
G is GETARS R is Recorder (Strip chart-part of plant instrumentation).
f!R indicates more than one (theI/) such recorders exist.
S is Sequence of Events Log LOG is the operations (manual) log.
Post Trip Log points are operator defined and therefore not listed here.
The Post Trip Log allows for 24 NSSS and 48 BOP real or pseudo analog points.
Additional data is also retrievable from the historical records processing function of the Plant Computer System.
ATTACHMENT 7 PAGE 1
NAME USE SOURCES REACTOR Scram Discharge Vol. Level Total Scram Channel A Scram Channel B Scram Manual Scram Total Isolation Channel A Iaolation Channel B Isolation SRM Signals (4)
IRM Signals (8)
APRM A APRM B APRM C APRM D APRM E APRM F LPRM A LPRM B LPRM C LPRM D LPRM E LPRM F LPRM Group A
LPRM Group B
NM Sys Trip A NM Sys Trip B NM Sys Trip A NM Sys Trip B
Flow Biased Rod Block Setpoint Flow Biased Scram Setpoint Rx Vessel Shut Down Level Rx Vessel Upset Level Alternate Reactor Water Level Setpoint Narrow Range Level (0 60)
Upset (Wide) Range Level (0 180)
Low Range Level (-150 to 60)
Bottom Drain Temperature Narrow Range Pressure (850 - 1050)
Wide Range'ressure (0-1200)
Total Core Flow A Recirculation Loop Flow B Recirculation Loop Flow A Recirculation Loop Temperature B Recirculation Loop Temperature A Recirculation Pump Drive Flow B Recirculation Pump Drive Flow A Recirculation Motor Generator Set Speed B Recirculation Motor Generator Set Speed A Recirculation (RPT) Breaker IE IE IE IE IE IE IE IE IE IE IE IE IE IE IE IE IE IE IE IE IE IE IE IE IE IE IE IE IE IE IE IE IE IE IE IE IE IE IE IE IE IE IE IE IE IE IE IE IE S
G G~S G,S G,S G
G G
R RIS G,R,S G,R,S G,R,S G,R,S G
G G
G G
G G
G S
S S
S G
G R
R G
G,R G,R,S G,R G,R G,R G,R,S G
G G
G G
G G
G G
G ATTACHMENT 7 PAGE 2
NAME USE SOURCES A Recirculation Trip 813 B Recirculation (RPT) Breaker B Recirculation Pump Trip 813 RCIC IE IE IE Reactor Core Isolation Cooling Initiation Reactor Core Isolation Cooling Pump Flow.
RCIC Turbine Speed RCIC Pump Suction Pressure RCIC Pump Discharge Pressure RCIC Steam Supply Pressure RCIC Turbine Exhaust Pressure RCIC Steam Flow dp-A RCIC Steam Flow dp-B RCIC Trip Signal RCIC Control Valve Position Reactor Core Isolation Cooling Controller Output HPCI SSP SSP SSP SSP SSP SSP SSP SSP SSP SSP SSP SSP High Pressure Coolant Injection Initiation High Pressure Coolant Injection Pump Flow High Pressure Coolant Injection Turbine Speed HPCI Pump Discharge Pressure HPCI Steam Supply Pressure HPCI Turbine Exhaust Pressure HPCI Steam Flow dp-A HPCI Steam Flow dp-B HPCI Control Valve Position HPCI Vessel Injection Valve Position High Pressure Coolant Injection Contr. Output RHR/LPCI SSP SSP SSP SSP SSP SSP SSP SSP SSP
.SSP SSP A. Residual Heat Removal System Flow B Residual Heat Removal System Flow SSP SSP G,R G,R AUTO DEPRESSURIZATION/SAFETY RELIEF VALVES ADS Initiation Relief Valve ABC Initiate Relief Valve GJKL Initiate Relief Valve DEHFP Initiate Relief Valve RSMN Initiate Relief Valve B Position Relief Valve D Position Relief Valve F Position Relief Valve H Position Relief Valve K Position SSP SSP SSP SSP SSP SSP SSP SSP SSP SSP ATTACHMENT 7 PAGE 3
NAME USE SOURCES Relief Valve L Position Relief Valve N Position Relief Valve R Position Relief 'Valve A Position Relief Valve C Position Relief Valve E Position Relief Valve G Position Relief Valve J Position Relief Valve M Position Relief Valve P Position Relief Valve S Position SSP SSP SSP SSP SSP SSP SSP SSP SSP SSP SSP MAIN STM LINES MSIV Initiation Signal A Inboard Main Steam Isolation Vlv Position B Inboard Main Steam Isolation Vlv Position C Inboard Main Steam Isolation Vlv Position D Inboard Main Steam Isolation Vlv Position A Outboard Main Steam Isolation Vlv Position B Outboard Main Steam Isolation Vlv Position C Outboard Main Steam Isolation Vlv Position D Outboard Main Steam Isolation Vlv Position A Steam Line Flow B Steam Line Flow C Steam Line Flow D Steam Line Flow A Main Steam Line Pressure Near Relief Valves B Main Steam Line Pressure Near Relief Valves C Main Steam Line Pressure Near Relief Valves D Main Steam Line Pressure Near Relief Valves Main Stm. Line Tunnel Temp Main Stm. Line Tunnel Delta Temp Main Stm. Line Radiation Main Stm. Line A HI Rad.
Main Stm. Line B HI Rad.
Main Stm. Line C HI Rad.
Main Stm. Line D HI Rad.
SSP SSP SSP SSP SSP SSP SSP SSP SSP SSP SSP SSP SSP SSP SSP SSP SSP IE IE IE IE IE IE IE TURBINE/GENERATOR First Stage Shell Press Main Stop Valve A Trip Main Stop Valve B Trip Main Stop Valve C Trip Main Stop Valve D Trip Main Control Valve A Trip Main Control Valve B Trip Main Control Valve C Trip Main Control Valve D Trip Generator Gross Megawatts Output IE IE IE IE IE IE IE IE IE IE R
S S
S S
S S
S S
G ATTACHMENT 7 PAGE 4
NAME USE
- SOURCES, Grid Frequency Grid Voltage Main Turbine Speed Main Turbine Trip (Note 5)
Generator Breaker Open (Note 3)
Control Valve
/11 Position Control Valve g2 Position Control Valve i/3 Position Control Valve 84 Position Total Bypass Valve Position FEEDWATER IE IE IE IE IE IE IE IE
'E IE G,R G,R G
G,S G
G,S GPS GJS G,S G
A Loop Feedwater Flow B Loop Feedwater Flow C Loop Feedwater Flow A Feedwater Line Temperature B Feedwater Line Temperature C Feedwater Line Temperature A Condenser Vacuum B Condenser Vacuum C Condenser Vacuum Feedwater Pump Suction Header Pressure A Feedwater Pump Turbine Trip (Note 2)
B Feedwater Pump Turbine Trip (Note 2)
C Feedwater Pump Turbine Trip (Note 2)
Startup Level Controller Output Feedwater Turbidity IE IE IE IE IE IE IE IE IE IE IE IE IE IE IE G,R G,R G,R G,R G,R G,R G
G G
G,R GPS Gis G,S G
R CONDENSATE CST Level
LOG CONTAINMENT Sup.
Pool Temp Sup.
Pool Level Drywell Temp Drywell Pressure Containment Area Rad. Monitor Suppression Pool Activity Primary Containment Trip (4 Signals)
STANDBY GAS TREATMENT SSP SSP SSP SSP SSP SSP IE R
R R
G,2R R
Note 1
S SGTS Stack Monitor Flow SGTS Air Flow Area Radiation Including Refueling Floor and Railroad Access Area SSP SSP SSP R
R R
ATTACHMENT 7 PAGE 5
NAME USE SOURCES Vent Stack Radiation including SGTS STANDBY LI UID CONTROL SSP SBLC Flow (Tank Level
& Time)
SBLC Initiation SSP SSP LOG LOG ATTACHMENT 7 PAGE 6
NOTES 1.
Not required as function of time.
Lab. Analysis of sample.
2.
Seq. of Events has four separate trip initiation inputs per turbine.
3.
See Seq.of Even'ts Inputs list for twenty-five Generator/Electrical
~ System trip initiation inputs.
4.
All five bypass valves are input to seq.
of Events log.
5.
See Seq. of Events log for twenty-seven individual trip initiation event inputs.
6.
Also see Seq. of Events log for various electrical system trip initiation event inputs.
ad/me] 057i: del ATTACHMENT 7 PAGE 7
GETARS SIGNAL LIST (TMS) ~
CN.e NAME UNITS DESCRIPTION 1
2 3
4 5
6 7
8 9
10 11 12 13 14 15 16 17 18 19 28 21 22 23 24 25 26 27 20 29 38 31 32 33 34 35 36 37 38 39 48 41 42 43 44 45 46 SCRAM SCRi%
SCRMB MSCRM ISOLT ISOLA ISOLB APP.MA APRMB APRMC APP.MD APRME APRMF LPRMA LPRMB LPRMC LPRMD LPRME LPRMF LPRM1 LPRM2 HFLX1 HFLX2 FBRBS FBSCM TCFLO CPDP RSLFA RSLFB RSLTA RSLTB RSDFA RSDFB RSDPA RSDPB MGSDA MGSDB ALF RSMCO MGSCOA MGSCOB MGMAA MGMAB MGSTA MGSTB RPT"A DIG DIG DIG DIG DIG DIG DIG 4(
0/
4/
Pl
/0 op 4/
Pt po
/4
à 4)
MLBH PSID MLBH MLBH DEGF DEGF KGPM KGPM PSID PSID
%SPD xSPD
/0
à A
Pt DIG TOTAL SCRAM CHANNEL 'A'CRAM CHANNEL 'B'CRAM MANUAL SCRAM TOTAL ISOLATION CHANNEL 'A ISOLATIOH CHAHNEL 'B'SOLATION APRM
'A'PRM
'D'PRM E
'B'PRM
'C'PRM
'D'PRM
'E'PRl'1
'F'PRM GROUP
'A'PRM GROUP
'B'EAT FLUX el HEAT FLUX 4'2 FLOW BIASED ROD BLOCK SETPOIHT FLOW BIASED SCRAM SETPOINT TOTAL CORE FLOW CORE PLATE DIFFERENTIAL PRESSURE
'A'ECIRCULATION LOOP FLOW
'B'ECIRCULATION LOOP FLOW
'A'ECIRCULATION LOOP TEMPERATURE
'B'ECIRCULATION LOOP TEMPERATURE
'A RECIRCULATION PUMP DP, IVE FLOIJ
'B'ECIRCULATION PUt& DRIVE FLOIJ
'A RECIRCULATION PUMP D IFFEPENTIAL PRESSURE
'B'ECIRCULATIOH PUMP DIFFERENTIAL PRESSURE
'A RECIRCULATION MOTOR GENERATOR SET SPEED
'B'ECIRCULATION MOTOR GEHERATOR SET SPEED AUTOMATIC LOAD FOLLOIJIHG MASTER RECIRCULATION COHTROLLER OUTPUT
'A' SET COtlTROLLER OUTPUT
'B MW SET COtkTROLLER OUTPUT
'A' SET WA STATIOH OUTPUT
'B' SET WA STATION OUTPUT
'B'W SET SCOOP TUBE POSITION
'A'ECIRCULATION (RPT)
BREAKER
49 SB 51
~ 52 53 54 55 56 57 58 59 68 61 62 63 64 65 66 67 68 69 78 71 72 73 75 76 77 78 79 88 81 82 84 85 86 87 88 89 98 91 92 93 94 95 96 97 98 99 MGFBB MGY-A MGV-8 RCICI RCFLO RCTSP RCPSP RCPDP RCTIP RCTEP RCETA RCETB RCVSA RCVTT RCVCO RCVYI RCFCO RCPSC RCEGM RCRGS HPC I I HPFLO HPSPD HPPDP HPTIP HPTEP HPETA HPETB HPVSA HPVST HPYCO HPVVI HPCO HPEGN HPRGS RHF-A RHF-B SMLFA SMLFB RHP-A RHP-B RHPCA RHPCB RHL"A RHL-B RHLCA RHLCB MISMA MSFLO FIJFLO FMLFA FIJLF8 4
PT-B FBA D IB DIG DIG VOLT YOLT DIG GPM RPM PSIG PSIG PS IG PSIG INCH INCH co 4/
4/
0/
DIG GPM RPt1 PSIG PSIG PSIG INCH IthCH 4/
0/
4/
KGPM KGPN KGPt"l KGPt1 PSIG PSIG 4p 4/
A
/4 MLBH MLBH MLBH NLBH
'B'ECIRCULATION (RPT)
BREAI'ER
'A'ECIRC ~
MOTOR-GENERATOR SET DRIVE l FIELD BKR.
'B'ECIRC.
MOTOR-GEHERATOR SET DRIYE i FIELD BKR.
'O'EC IRCULATIOH MOTOR-GENERATOR SET VOLTAGE
'B'ECIRCULATION MOTOR-GEHERATOR SET VOLTAGE REACTOR CORE ISOLATIOH COOL IHG INITIATION REACTOR CORE ISOLATIOH COOLING PUMP FLOW RCIC TURBINE SPEED REACTOR CORE ISOLATIOH COOLIHG SUCTION PRESSURE RCIC PUMP DISCHARGE PRESSURE RCIC STEAM SUPPLY PRESSURE RCIC TURBINE EXHAUST PRESSURE RCIC STEAM FLOW DELTA " P 'A" RCIC STEAM FLOW DELTA -
P '8"
~ RCIC STEAM ADt1ISSIOH VALVE POSITION RCIC TRIP 1 THROTTLE VALVE POSITION RCIC CONTROL YALYE POSITION RCIC VESSEL IHJECTION VALVE POSITION REACTOR CORE ISOLATIOH COOLIHG CONTROLLER OUTPUT RCIC SUCTIOH PRESSURE COHTROLLER OUTPUT RCIC TURBINE EGM OUTPUT RCIC RAMP GEHERATOR SIGHAL / CONVERTER OUTPUT HIGH PRESSURE COOLAHT INJECTION INITIATION HIGH PRESSURE COOLAHT INJECTION PUMP FLOW HIGH PRESSURE COOLANT INJECTION TURBINE SPEED HPCI PUMP DISCHARGE PRESSURE HPCI STEAN SUPPLY PRESSURE HPCI TURBIHE EXHAUST PRESSURE HPCI STEAM FLOW DELTA - P "A" HPCI STEAM FLOW DELTA " P 'B" HPCI STEAN ADMISSIOH YALVE POSITION HPCI STOP VALVE POSITIOH HPCI COHTROL YALYE POSITION HPCI VESSEL INJECTIOH YALVE POSITIOH HIGH PRESSURE COOLAHT INJECTIOH CONTROLLER OUTPUT HPCI TURBINE EGM OUTPUT HPCI RAMP GENERATOR i SIGNAL CONYERTOR OUTPUT
'O'ESIDUAL HEAT REMOVAL SYSTEM FLOW
'B'ESIDUAL HEAT REMOVAL SYSTEM FLOW
'A RHR SYSTEM SERVICE IJATER FLOIJ
'8'HR SYSTEM SERVICE IJATER FLOW
'A'ESIDUAL HEAT RENOVAL Hx PRESSURE
'B'ESIDUAL HEAT REMOVAL Hx PRESSURE
'O'HR Hx PRESSURE CONTROLLER OUTPUT
'B'HR Hx PRESSURE CONTROLLER OUTPUT
'O'ESIDUAL HEAT REMOYAL Hx LEVEL B'ESIDUAL HEAT REMOVAL Hx LEVEL
'O'HR Hx LEYEL COHTROLLER OUTPUT
'B'HR Hx LEYEL CONTROLLER OUTPUT STEAN FLOIJ / FEED FLOIJ MISMATCH TOTAL STEAN FLOIJ TOTAL FEEDMATER FLOW
'O'OOP FEEDIJATER FLOW
'B'OOP FEEDIJATER FLOW
183 184 185 186 187 188 189 118 111 112 113 114 115 116 117 118 119 128 121 122 123 124 125 126 127 128 129 138 131 132 134 134 135 136 137 138 139 140 141 142 143
, 144 145 146 147 148 149 158 151 152 153 154 FIJLTB FWLTC FWPFA FIJPFB FIJPFC F WPDA F WPDB FIJPDC VAC-A VAC"8 VAC-C CPDHP FIJPSA FIJVSU FWYCA FIJVCB FIJVCC FWTSA FWTSB FWTSC FWTRP FWTRA FWTRB FWTRC FWCVA FIJCYB FWCVC FWSCO FLNCO FWMRA FWMAB FWMAC FIJFGA FWFGB FWFGC FLEA FWMFB FWMFC IJLSET NRIJL WRWL.
LRWL MS IA MS IB MS IC MS ID MSOO MSOB MSOC MSOD MSLFA MSLFB 1
IJLFC 1
FWLTA MLBH DEGF DEGF DEGF GPt1 GPM GPM PSIG PSIG PSIG IHCH INCH INCH PSIG PSIG N
Po Pt RPM RPM RPM DIG DIG DIG DIG 4p 4/
0/
4/
4/
Pt INCH IHCH INCH INCH DIG DIG DIG DIG DIG DIG DIG DIG MLBH MLBH
'C'OOP FEEDWATER FLOW
'O'EEDIJATER LIHE TEMPERATURE
'B'EEDWATER LINE TEMPERATURE
'C'EEDWATER LINE TEMPERATURE
'O'EEDWATER PUMP FLOW
'B'EEDWATER PUMP FLOW
'C'EEDIJATER PUMP FLOIJ
'A'EEDIJATER PUt1P DISCHARGE PRESSURE B'EEDIJATER PUMP DISCHARGE PRESSURE
'O'EEDIJATER PUMP DISCHARGE PRESSURE
'O'ONDEHSER VRCUUM
'8'ONDEHSER VACUUM
'C'ONDENSER VACUUI1 COHDEHSATE PUMP DISCHARGE HEADER PRESSURE FEEDIJATER PUMP SUCTION HEADER PRESSURE FEEDIJATER STRRTUP VRLVE POSITION FEEDIJATER FLOW CONTROL VALVE 'O'OSITIOH FEEDWATER FLOIJ CONTROL VALVE 'O'OSITION FEEDIJATER FLOW COHTROL VALVE 'C'OSITIOH
'O'EEDIJATER PUMP TURBIHE SPEED
'8'EEDWATER PUMP TURBIHE SPEED
'C'EEDIJATER PUMP TURBIHE SPEED ALL FERDWATER PUMP TURBINE TRIP
'A'EEDWATER PUMP TURBIHE TRIP
'B'EEDWATER PUMP TURBIHF TRIP
'C'EEDIJATER PUMP TURBIHE TRIP
'A'EEDWATER PUMP TURBINE CONTROL VALVE POSITION
'8'EEDIJATER PUMP TURBINE CONTROL VALVE POSITION
'C'EEDWATER PUMP TURBItlE CONTROL VALYE POSITIOH STARTUP LEYEL CONTROLLER OUTPUT FEEDWATER MASTER CONTROLLER OUTPUT
'O'EEDIJATER TURBINE BIAS M / R STATIOH OUTPUT
'B'EEDIJATER TURBIHE BIAS M / A STATIOH OUTPUT
'C'EEDIJATER TURBIHE BIAS M i A STATION OUTPUT
'O'EEDIJATER TURBINE FUNCTION GEHERATOR OUTPUT
'8'EEDIJATER TURBIHE FUNCTION GENERATOR OUTPUT
'C'EEDWATER TURBINE FUNCTIOH GENERATOR OUTPUT
'O'EEDWATER PUMP MINIMUM FLOW CONTROLLER OUTPUT
'8'EEDIJATER PUMP MINIMUMFLOIJ CONTROLLER OUTPUT
'C FEEDWATER PUMP MIHIMUMFLOW COHTROLLER OUTPUT REACTOR IJATER LEVEL SETPOIHT NARROW RAHGE LEVEL (8 " 68)
WIDE LEVEL (8 " 188)
LOIJ RANGE LEYEL (-158 TO 68)
'O'HBOARD MAIH STEAM ISOLATIOH VALVE POSITION
'B IHBOARD MAIH STEAM ISOLATIOH YALVE POSITIOH
'C'NBOARD MAIH STEAM ISOLATION VALVE POSITION
'D INBOARD MAIN STEAM ISOLATIOH VALVE POSITION
'A OUTBOARD MAIH STEAM ISOLATION VRLVE POSITIOH
'B'UTBOARD MAIN STEAM ISOLATION VALVE POSITION
'C'UTBOARD MAIH STEAM ISOLRTION VRLVE POSITION
'D'UTBORRD MAIN STEAM ISOLATION VALVE POSITIOH
'A STEAM LINE FLOIJ
'8'TEAM LINE FLOIJ
'0%
p R 6 wO GD X 0 ITI 00
158 159 160 161 162 163 164 165 166 167 168 169 170 171 172 173 174 175 176 177 178 179 180 181 182 183 184 185 186 187 188 189 190 191 192 193 194 195 196 197 198 199 200 201 202 203 204 205 206 207 208 SRVII SRV 12 SRVI3 SRVI4 SRVPB SRVPD SRYPF SRYPH SRVPK SRVPL SRVPN SRVPR SRVPA SRVPC SRVPE SRVPG SRVPJ SRVPM SRVPP SRVPS MWE GFREQ GVOLT MTSPD TBSF MTT MGB PLUNB LDSET TPSET PRO CVAO PRSPA PRSPB TIPA TIPB CVPT CYP 1 CVP2 CYP3 CVP4 SVP 1 SYP2 SVP3 SVP4 BPVT
'PVP 1
BPVP2 BHDT DWP HRDP 1'FC 1
%LFD 15?
ADS I MLBH MLBH DIG DIG DIG DIG DIG 5VDC 5VDC 5VDC 5VDC
'5YDC 5VDC 5VDC 5VDC 5VDC 5VDC SVDC 5VDC 5VDC 5VDC 5VDC 5VDC MWE HZ KV RPM MLBH DIG DIG DIG Po op 4/
/4 Ye A
DEGF PSIA PSIG
'C'TEAM LIHE FLOW D'TEAM LINE FLOW ADS IHITIATIOH RELIEF VALYE ABC INITIATE RELIEF YALYE GJKL IHITIATE RELIEF VALVE DEHFP IHITIATE RELIEF YALVE RSMN INITIATE RELIEF VALVE '8'OS ITIOH RELIEF VALVE 'D'OSITION RELIEF VALVE 'F'OSITION RELIEF VALVE OS IT ION RELIEF VALVE 'K'OSITIOH RELIEF VALYE 'L'OSITION RELIEF VALVE OS ITION RELIEF VALVE 'R'OSITIOH RELIEF VALVE 'A'OSITIOH RELIEF VALVE 'O'OSITIOH RELIEF VALVE 'E'OSITIOH RELIEF VALVE 'G'OSITIOH RELIEF VALVE 'J'OSITION RELIEF VALVE 'M'OSITIOH RELIEF YALVE 'P'OSITIOH RELIEF VALVE OS ITIOH GENERATOR GROSS MEGAWATTS OUTPUT GRID FREQUENCY GRID VOLTAGE MAIH TURBINE SPEED TURBINE STEAM FLOW MAIN TURBINE TRIP GEHERATOR BREAKER OPEN POWER l LOAD UHBALANCE LOAD SET TRAHS IEHT PRESSURE SETPO INT PRESSURE REGULATOR OUTPUT CONTROL VALVE AMPLIFIER OUTPUT
'A'RESSURE REGULATOR SETPOINT
'8'RESSURE REGULATOR SETPOINT
'.A'RESSURE REGULATOR SENSED PRESSURE
'B'RESSURE REGULATOR SEHSED PRESSURE TOTAL CONTROL VALVE POSITIOH CONTROL VALVE <<1 POSITION COHTROL VALYE <<2 POSITION COHTROL VALVE <<3 POSITIOH COHTROL VALVE <<4 POSITION STOP VALVE <<1 POSITIOH STOP VALVE <<2 POSITIOtl STOP VALVE <<3 POSITIOH STOP VALVE <<4 POSITION TOTAL BYPASS YALVE POSITIOH
<<1 BYPASS VALVE POSITIOtl
<<2 BYPASS VALVE POSITIOt.l BOTTOM DRAIN TEMPERATURE DRYIJELL PRESSURE NARROW RANGE PRESSURE (850 1050)
DP 2
ICBUS 211 1DBUS 212 1EBUS 213 1FBUS 214 EVEHT 215 MSLPA 216 MSLP8 217 MSLPC 218 MSLPD PSIG DIG DIG DIG DIG DIG PSIG PSIG PS IG PSIG IJIDE RANGE PRESSURE (8 - 1288) eIC BUS POWER
<<ID BUS POIJER 41E BUS POlJER
%IF BUS POWER EVEHT MARKER
'A'AIN STEAM LINE PRESSURE HEAR RELIEF VALVES
'B'AIN STEAM LINE PRESSURE HEAR RELIEF VALVES
'C'AIN STEAM LINE PRESSURE HEAR RELIEF VALVES
'D'AIN STEAM LINE PRESSURE HEAR RELIEF VALVES 21 1
12
'1 13 1
14 1
15 1
1 1
193 1
194 1
195 1
196 1
4 6
6 6
7 7
4 4
4 4
38 28 31 32 1
18 5
6 7
8
l 4
SCRAH TII1ING SIGNALS.
CH.e NAME UN ITS SUB-CHAllNELS LINK SB LNK PORT 403 484 485 486 487 488 489 498 491 492 493 494 495 496 497 498 499 588 581 582 S83 584 585 586 587 SGB 589 518 41 Spare CRD Channel CRD82 DIG4 CRD S
34-83 Spare CRD Channels 38-83.
42-83.
14-87.
CRD86 D IG4 CRD87 DIG4 CRD'S 26"11. 38"11. 34-11. 38-11.
CRD'S 42-11 46-11 58-11, 6-15.
Spare CRD Channels CRD11 DIG4 CRD12 DIG4 CRD13 DIG4 CRD S 2"19 6-19.
18-19.
14-19.
CRD S
18-19 22"19. 26-19.
38-19.
CRD'S 34-19 38-19.
42-19.
46-19.
Spare CRD Channels CRD28 DIG4 CRD21 DIG4 Spare CRD Channels CRD26 DIG4 CRD27 DIG4 CRD20 DIG4 CRD S
2-35.
6-35 18-35.
14-35.
CRD S
18-35. 22-35. 26-35. 38-35.
CRD'S
.34"35. 38-35.
42"35. 46"35.
CRD'S 26-27. 38-27.
34"27. 38-27.
CRD'S 42-27. 46-27.
58-27.
54-27.
13 1
13 1
13 1
13 1
13 1
13 1
13 1
13 1
14 1
14 1
14 1
2 11 12 13 28 21 1
2 3
I 0
'I
8539 5
8548 5
8541 51 8542 514
.8543 515 8544 516 CRD34 8545 517 CRD35 8546 518 CRD36 8547 519 8548 528 8549 521 8558 522 8551 523 CRD41 8552 524 8553 525 8554 526 8555 527 8556 528 8557 529 8558 530 8559 531 Spare CRD Channels DIG4 CRD'S 18"43.
14-43.
18-43.
22-43.
DIG4 CRD'S 26'-43. 38-43.
34-43. 38-43.
DIG4 CRD'S 42-43. 46-43. 58-43. 54-43.
Spare CRD Channels DIG4 CRD'S 18-51, 22-51, 26-51, 38-51.
Spare CRD Channels RD29 DIG4 CRD'S 58-35. 54-35.
58-35.
2-39.
14 1
14 1
14 1
14 1
14 1
I 9
~
18 11 16
$D S WO 0 2 H) R 00
Attachment 9 SSES Unit, 1 GETARS Points Presently Used to Trigger Recording Channel Descri tion Manual Scram 46 A Recirc RPT Breaker 47 B Recirc RPT Breaker 183 Safety Relief Valve "S" Position 214 Narrow Range Reactor Vessel Pressure
ITEM CLASSIFICATION LOGIC FOR PARTS OF ASSEMBLIES ATTACHMENT 10 tl
~
NO Safety Related Assembly' ASME Sec. III Class 1,
2 or 3
Assembly YES Safety Related Assembly?
YES YES ls the part in a Class IE Circuit?
NO Active Assembly?
YES NO YES Is the part a
Subcomponent?
NO t
YES YES NO I
Must the part perform its function during or after a Design Basis Event?
Pressure Retaining Part?
NO YES I
0 G1 Pressure Retaining Part?
YES I
Qi N
Commercial Grade Part?
Critical Non-Press Retaining Part?
NO YES Ql Q2 Commercial Grade Part (with the excep-tion that qualification to NUREG 0588, CAT. I and IEE 323-74 is, required)?
NO I
QN ES Commercial Grade Part?
NOTES:
YES Q2E I
NO I
QQ1E NO YES I
Ql Q2 1.
Commercial Grade Item An item that:
a.
is not subject to design or specification requirements unique to facilities/
activities licensed by the NRC.
b.
is used in applications in addition to facilities licensed by the NRC.
c.
is listed in a manufacturer's or distributor's published product description, e.g.: catalog.
2.
Critical Non-pressure Boundary parts are non-pressure retaining parts or appurtenances whose failure would compromise the proper operation of an active engineered safety component.
Specif ication 8856-E-402B Revision 4
ATTACHMENT ll
.Page l of 7
. 0 CONDITIONS OF SERVICE 6.1 Elevation
- The Cable Penetrations will be installed indoors between elevations of approximately 707 to 712 feet above sea level.
6.2 Service Each electrical penetration assembly shall be suitable for installation into a contairment structure penetration nozzle which is part of the contaiment structure.
6.3 Design Life The assemblies shall have a minimum design life of 40 years.
At the end of 40 year life, they shall be.capable of meting the requirements for normal and emergency cperation in accordance With the follovirg sect lonse 6.4 Normal Envirorxnent - (For Penetrations'No.
2H104E and 2W100H)
The outboard end of Cable Penetrations vill be subject to the folloving corditions during normal, contiruous cperation.
a.
Pressure b.
Ambient temperature
-0.375" vg 115'F (maximum contimous) 60'F (minizwm) c.
Relative Humidity 10% to 90%
d.
Radiation dose (Gaum@)
0.100 Rads per knur (ave-rage) 3.5 x 104 rads over 40 years
'The inboard end of the Cable Penetrations vill be subject to the fol-lmirg corditions durirg
- normal, contirxvs ceration.
(The values inside bracket are for wetwell penetrations.
'The other values are for
~rvell penetrations.)
a.
Pressure b.
Ambient Temperature c.
Relative Humidity d.
Radiation dose (Gamma) 0.1 to + 1.5 psig (0.1 to + 1.5 psig) 150'F Max. Cont.
20% to 905 (10(S contimous) 4 rads per hour (0.1 rads per hour)
Aver~ 1.1 x
10 rads (3.5xl04 rads) over 40 years P126/1-3.2
Specif ication 8856-E-402B Revision 4
ATTACHMENT 11 Page 2 of 7 The. above data is without margins.
Margins shall be added for 'qualifica-
~
~
~
~
~
~
~
~
tions as required by IEEE Std. 323-1974 per NUREG-0588 Category I.
- 6. 5 Khergency Environment -
(For Penetrations No.
2H104E and 2N100H)
The outboard end of the cable penetrations will be subject to the fol-lowing conditions during abnormal condition:
a o Pressure b.
Temperature 2o 2 pslg 300'F 0-60 seconds 130'F 6~econds-100 days c.
Relative Humidity d.
Radiation dose (Gamma) e.
Radiation dose (Beta) 100't, 1-12 Hrs.
90%
12 Hrs - 100 days 8.3 x 105 rads per hour (averagej 1.7 x 10~ rads (TID) 5.0 x 103 rads per hour (average
)
1.1 x 106 rads (TID).
TID shown in (d) and (e) includes normal 40 years dose plus accident dose s All Cable Penetrations shall maintain their pressure seal and all elec-trical circuits shall remain operable when the interior ends of the penetrations are subject to the following emergency conditions at 100%
relative humidity:
(Ihe values inside brackets are for wetwell pene-trations.
'Ihe other values are for drywell penetrations.
)
a.
First 45 seconds Pressure Thngerature 44 psig (29 psig) 40oF (130 DF)
The tea of rise frcm normal to the above temperature and pressure will be within 10 seconds.
b.
Next 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> Pressure Ibmperature 35 psig (30 psig) 340 F (200 F) c.
Next 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> Pressure
.Taqoerature 35 psig (30 psig) 3200F (210'F)
F126/1-13
K Specification 8856-E-402B Revision ATTACHMENT 11 Page 3 of 7 d.
Next 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> Pressure Terrgerature 20 psig (15 psig) 250 F (210 F) e.
Next 99 days Pressure Tenqerature 10 psig (10 psig) 200'F (200 F-next 120 hrs.t 140'F-120 hrs.-99 days) f.
Radiation Dose (Gamna) 1.6 x 10 Rads/hr. (1.9x10 Rads/hr) (average) 7.1 x 107 Rads(2.7xl07 Rads) (TID) g.
Radiation Bose (Beta) 2.3 x 10 Rads/hr. (2. 3x188 Rads/hr) (average) 1.9 x 109 Rads (1.9x109 Rads)
(TID)
All above data is without margins. ~ margins shall be added for qua-lification as required by IEEE Std. 323-1974 per
[NUREG-0588, Category I.
TID shcwn in (f) and (g) includes 40 year dose plus accident dcse.
6.6 Eesign Ehta
'Ihe cable penetrations shall be suited for the following conditions:
a)
Radiation Streaming along axis of cable penetration 15-100 mr/hr.
b)
External differential pressure (drywell to interior of the terminal boxes) 5 psig Q)
CoHfo itiwwC I e45t +4( de
~
n Pr~S~<C r 'F 5S pZ l q 5'alfer s tt ~<+ ca(fa ~~+~~* 42 lsd iH e(ado;W M4~4C
~a
~ cr 45H sf i fabri ~~4'gag.
- 7. 0 SEISNIC A%)
ROD IC 7.1 We Cable Penetrations shall meet the requirements of Specif ication 8856~22, "General Project Requirerrents for Design Assessment and Qualification of Seismic Category I Equiprent and Equiprent Supports for Seismic and Hydrodynamic loads".
In addition, the cable penetra-tions shall be in cargliance with IEEE Standard No.
- 344, "Recarrrended practice for Seismic Qualification of Class I Electric Equipnent for Nuclear ~r Generating Stations."
7.2 During and after all KBE's penetrations must maintain pressure seal integrity.
In addition, the penetraticns indicated in Table I, Note 1.2 (see Material Requisition) mast remain functional.
P126/1-14
Specification 8856-E-4028 Revxszon 4
ATTACHMENT ll
.Page 4 of 7
.0 FZNISH
- 10. 1.Surface preparation, prix'oat application, shop priming system, mterial, dry film thickness, inspection and tests shall greet require-ments of Speci fication 885~
"General Project Requirenents for Shcp Priming of Mechanical and Electrical Equipment".
10.2 All machined surfaces of carbon steel which vill not be welded upon shall be temporarily protected against oxidation during shipment ard storage by liberal application of a
grease coating conforming to MilitarySpecification MIL&-16173, Grade I: "Corrosion Preventive Ccmpouna, Solvent Cutback, Cold-Application".
10.3 All machined surfaces of carbon steel to be welded, and all carbon steel surf aces within 2 inches of field welds shall be protected, against corrcsion during shipnant and storage by a
coat of EEWX Aluminate or Buyer approved equal, which can readily be repaved prior to field welding.
10.4 Seller shall not finish paint the equipnant.
10.5 'eleted.
10.6
'Zhe cable penetration design shall permit field cleaning, priming ard painting 'in those areas that vill require it after field welding.
ll.0 DEFECTIVE MA1'EQUAL Warranty and limitation of liability shall be as stated in the Purchase Order.
- 12. 0 QJALITY ASSURANCE 12.1 The Cable Penetrations listed in the Material Requisition shall pex-1 ~f
'*1 p
p will require a high level of quality control ard eke~ ntation of design and nanufacture.
For this equipnant the Seller shall maintain a Quality Assurance Program which ccmplies with Specification 8856-G-9, "General Project Requirem nts for Quality Assurance on Purchase Orders for 'O'esignated
- items, ard the requirements af the appli-cable specif ications.
The Seller shall furnish the documntation as specified below and on foans G-321& and 8856-QQ, in accordance with the applicable require-ments of Specification 855~9.
P126/2-11
Specif ication 8856-E-4028 Revision ATTACHMENT ll Page 5 of 7 12.2 Each Cable Penetration shall be qualified by analysis and successful
, use under similar conditions. or by actual test to demonstrate its ability to perform its function under normal and emergency conditions of service as defined in Section 6.0 of this Specification.
Compliance
-.with the following requirements shall be donmented section by section.
A general statement of conformance will not be sufficient.
'Ihe analysis and testing program shall meet or exceed the requirements contained in:
12.2.1 IEEE No.
323-1974, "Qualifying Class I Electric Equipment for Nuclear Power Generating Stations," per NUREG-0588 Category I.
l2.2.2 IEEE Standard 317, "Electrical Penetration Assemblies in Con-tainment Structures for Nuclear ~r Generating Stations,"
and as required in specification paragraph 13.1.
12.2. 3 (1KLETED) 12.3 Intentionally left blank.
12.4 Intentionally left blank.
12.5 Intentionally left blank.
- 12. 6 written docanentation and/or certified test results shall be sub-mitted verifying that the Cable Penetrations meet the following:
12.6.1 Factory Production Tests as required in. specification paragraph
- 13. 2.
12.6. 2 Materials Manufacturer' or Materials Supplier' Certified Material Test Reports and/or Certificates of Gcapliance and ASME Design Reports as required in specification ra raph 13.4.
g 12.6. 3 NorMestructive Test Wsults as required in Specif ication paragraph 13.5.2
- 12. 7 Evidence of ccxrpliance to seismic requirements test procedures and test reports shall be furnished in accordance with speci fication paragraph 7.0 Seismic Requirements.
12.8 Certificate of ccepliance and test
- reports, indicating that the cable penetration has been qualified to the Section 12.2 requirements, shall be suhnitted.
12.9 Before the penetration is placed in servicei copies of the appropriate Certificate Holder's Data Report shall be filled in by the owner with the Cmmanwealth of Pennsylvania which is the enforcenent authority.
P126/2-12
~ 1 I
I
~
~
3,3.0
'V&IS AND TEST REPORTS 13.1 Desi Qualification Speci fication 8856-E-402B Revision ATTACHMENT 11 Page 6 of 7 The prototype testirg shall be rret by one of the Wo follcrwing methods or a canbination thereof:
Method I:
Prior testirg of pzototypes of like d sign which in the judgement of the Purchaser are adequate to neet the requirements state herein.
The buyer will use, as one basis for judgem nt, the IEEE Std No.
- 323, the guidance in 14BKG&588, Cat. I and IEEE Std Qo.
- 317, IEEE Std.
No. 344 and Specification 885~22.
Method II: Prototype testing perfozmed after award of a
purchase order on individual cable penetrations of the
- types, sizes and conductor grcupirgs as specified herein.
If the Bidder proposes to qualify the cable penetration by Method I, he shall subnit tests he has perfozmed and adequately docum nted and which specif ic types of assemblies he believes are qualified under Method I abcve.
Docunentation shall include ccmplete description of the prototype cable penetrations, date and place of tests, test procedures, re-sults and ccpies of certified. test reports.
Method I will be ac-ceptable only if the prototype assemblies are clearly and urquestion-ably representative of the actual product supplied to the Buyer.
In the event the Buyer judges that Method I. above is inadequate to prcperly qualify irdividual asseablies or all assemblies, it will be necessazy that prototype testing under Methcd II be per-formed.
13.1.1 Prototype Tests Prototyp. tests, if required, shall be performed on a prototype asserrbly of each type as defined in section 8.1 and each rating as def ined in section 8.2.
These tests shall be perfozmed in accordance with section
- 6. 0, "Conditions of Service",
ard section 8.0 "Type and Ratirg" of this Specification.
13.1.2
~
~
13.1.3 Existing qualification test data may be used to broaden the cover-age of a particular test when it can be shown that the existirg test data are valid for the penetration assembly being supplied.
The Seller shall suhnit docunantation showirg that each of the materials used in the fabricated Cable Penetration, includirg bush-irgs, pigtails ard/or connectors, is capable of withstandirg the radiation enviroment described in Section 6.0<
"Conditions of Ser-vice" without ccmpranisirg its ability to remain cperational.
It shall be demonstrated by test (using an equivalent
- sard, gravel or concrete mix) that the nozzlemcncrete interface temperature will not exceed 150'F at any. point under the rrcst adverse conditions of rxnaal cperation stated in section 9.1.5 Wile all corductors carry rated current.
This deaanstration raed only be applied to the lm voltage or medium voltage assemblies with the largest calculated
4i 4
0 ki 13.1.4 Specification 8856-E-402B Revision ATTACHMENT ll Page 7 of 7 After the tests it shall also be demonstrated that these circuits identified in Table I as required to operate in an emergency will remain functional and the remainder maintain leaktight -integrity.
13.1 ~ 5 No Cable Penetration that has been subjected to the above prototype tests will be accepted for installation at the jobsite. ~ver, the header plate may be reused if it meets the quality assurance requirements.
13.1.6 At the Buyer's option all of the prototype tests may be witnessed.
The Buyer shall be notified 30 days in advance of the testing.
0 rtified test reports are required for each cable penetration tested.
13.2 Facto Production Tests (Non-Destructive j Production tests shall be performed on each penetration assembly h'
~nths Before tests are to be performed, the Seller shall submit a written description of the tests and of the test procedures.
In addition to meeting the requirements of IEEE
- 317, the fol-lowing supplementary requirements shall also apply:
- 62. pig(pl<<c the ~rgi~x pac lEEE-Zi7) 13.2.1 Each assembly shall be proM. tested across pressure barrier seals 'ith Heliun to verify structural integrity, for field test. P~e~~ti~ +<<< ~ Po<<
l-2 ap IDGE-3tg s4 (I ~e f-Wr~i'rs ~i +%SF Sect. w6'5Zo 13.2.2 We leakage rate of each penetrat'i'on assembly shall be determined and shall not exceed 1
x 10 6
std cc of Helium per second.
- 13. 2. 3 After the routine leak rate tests each cable or conductor and associated connectors, except coaxial and triaxial conductors and connectors, in all cable penetrations shall be tested per and shall meet the requirements of N~ICEA Standards of Voltage Tests after Installation except that the cables need not be bmersed in water if the type of construction shall be tested against every other conductor and also tested to the metallic end plates.
I C
I C'.
13.2. 4 Cbaxial and triaxial cables and connectors shall receive a dieletric withstand voltage test. ~ dielectric with-stand test shall be at the manufacturer's maximxn rated voltage between conductors and between conductors and inner, or first, shield for a period of cne minute.
P126/3-1
ATZACHhKNT 12
.1-1082 Revision 1
Largest motor E'o be,sta ted 2000 K'1640 amps at 0.18 pf during star"ing)
Elevation at site 676 ft 5.2.2 Design Environmental. Conditions a)
Temperature - Engine off (Min/Max).
b)
Temperature - Engine Running:
72'/104'F Max. At Generator At Engine 120 F 160'r c)
Outside Air Temperature Dry-bulb air temperature,
.minimum/maximum
-19 F/105 F Wet-bulb air temperature, uMdznsm/maximum
-13'F/82'F d)
Pressure Atmosphe"e e)
Radiation - integrated Dose 1.8 x 10 Rads 2
f)
Relative Humidity, Max/YM 90'/5Z g)
Vendor shall supply a certificate of compliance for the listed environment conditions.
5.2.3 Cooling Mater Available a)
Cooling water for the heat exchangers will be raw water from the spray pond receiving approximately 100 gpm of warm circulating ~ater for deicing purposes.
Supply pressure Max. Temperature Min. Temperature The cooling water will have 130 psig Max.
95'F 33'F the following analysis:
Calcium as Ca Magnesium as Mg Sodium as Na Average 121 ppm 35.5 ppm 9 '
ppm Max.
240 ppm 81 ppm 34 ppm
w ATZACHhKNT 12 Page 2 of 2
<-1082 Revision 5,2.4 Diesel Generatoz Building The diesel generator unit will be housed in stzucture.
The unit is completely enclosed protected cell.
The unit is also protected such as tornadoes, floods, lightning, rain, a reinforced concrete, Class I in its own concrete and missile against other natuzal hazards ice or snow.
5.2.5 Seismic Requirements:
a)
The diesel generatoz system is a Seismic Category E and safety related system.
The diesel generator with aux~liary systems and supports shall be de'signed to function through and after five (5) Operational Basis Earthquake followed by-a Safe Shutdown Earthquake (SSE) and shall meet the requirements oz the General Specification foz Design Assessment and Qualification of Seismic Categozy I Equipment and Equipment Supports for Seismic h Hydrodynamic Loads, G-1024.
Use of the Seller's standard more stringent seismic qualification procedure for control devices may be substituted if the procedure is reviewed and approved by Buyer prior to testing.
b)
The Seller shall design and submit seismic calculations for each diesel generator, control panels, control devices and all other accessories as required either in accordance with the attacned General Specification G-1024, or to the pertinent manufacturer's procedures providing such procedures have prior approval by the Buyer.
c)
The, Seller shall also state the max~ allowable loads for the Seller's equipment nozzle interfacing with Buyer' supplied components or piping.
5.3 PERPOR BNCE RZQUXR ANTS:
5.3.1 The diesel generator will be installed as a backup to the four existing diesel generatozs, which are used in the event of loss oz all normal sources of power.
Upon loss of powez the diesel generators shall stazt automatically and attain full ooerating voltage and rated zzequency within ten seconds.
Each diesel generator feeds an independent 4.16 kV bus arrangement foz each.
zeactor unit (2 units).
Diesel-generator voltage setpoint control will be set by the Buye to produce a full operating voltage within the range of 4160 volts plus or minus 5X as permitted by NE>.LG1-22.47 to suit plant operating conditions.
5.3.2 The diesel engine shall be capable of.starting from cold.
Reliability of automatic start and operation is of utmost importance and shall be of prime consideration in the design of the equipment.