ML20080J029
| ML20080J029 | |
| Person / Time | |
|---|---|
| Site: | Zimmer |
| Issue date: | 09/16/1983 |
| From: | Williams J CINCINNATI GAS & ELECTRIC CO., COLUMBUS & SOUTHERN OHIO ELECTRIC CO., DAYTON POWER & LIGHT CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, RTR-REGGD-01.097, RTR-REGGD-1.097 GL-83-33, LOZ-83-0125, LOZ-83-125, NUDOCS 8309260108 | |
| Download: ML20080J029 (75) | |
Text
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Pi g THE CINCINNATI GAS.& ELECTRIC COMPANY
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CINCINN ATI, OHIO 45201 September 16, 1983 LOZ-83-0125 J. WILLIAMS, JR SENIOR WCC PRESIDENT NUCLEAM OPERATIONS Docket No. 50-358 Mr. Harold Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C.
20555
Dear Mr. Denton:
r-RE:
MM. H. ZIMMER NUCLEAR POWER STATION - UNIT 1 NRC GENERIC LETTER 82-33 (DECEMBER 17, 1982),
REQUIREMENTS FOR EMERGENCY RESPONSE CAPABILITY, SUPPLEMENT 1 TO NUREG-0737 This letter is in response to NRC letter dated May 18, 1983 from Mr. B. J. Youngblood, which enclosed a schedule for implementing Supplement No. 1 to NUREG-0737.
Enclosed are forty-three (43) copies of the following two i
reports:
1.
Report for Wm. H. Zimmer Nuclear Power l
Station - Regulatory Guide 1.97, Revision 2 and NUREG-0737, Supplement No.1.
2.
Safety Parameter Display System Imple-mentation Plan (Attachment 1 to this report contains information proprietary to S. Levy, Inc.
Three copies of this attachment are being provided under sepa-rate cover to Mr. L. Kintner, Licensing
~
Project Manager, for use by the NRC Staff).
i Appendix L of the FSAR, " Requirements Resulting from TMI-2 Accident," will be changed to reflect this information in a forthcoming FSAR revision.
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8309260108 830916 PDR ADOCK 05000358 F
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Mr. Harold Denton Director Septembe-16, 1983 LOZ-83-01.5 Page 2 CG&E has no request for exemptions from the requirements of the Emergency Response Facilities described in Supplement No.1, NUREG-0737.
Very truly yours, W
THE CINCINNATI GAS & ELECTRIC COMPANY By 2
_au
. WILLIAMS, JR.
SENIOR VICE PRES DENT SJT/sfr State of Ohio
)
Ss County of Clermont)
Sworn to and subscribed before me this
/k day of September, 1983.
lILEth W Qf M
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!!otary Public PATRICIA LOUISE EROWN l
Notary Public, Stata of Ohio My Comminion D;;tes May 1,198S l
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Mr. Harold Denton Director September 16, 1983 LOZ-83-0125 Page 3 cc:
John H. Frye III M. Stanley Livingston Frank F. Hooper Troy B. Conner, Jr.
John E. Dolan James P. Fenstermaker SA Steven G. Smith William J. Moran Stephen F. Koziar, Jr.
Samuel H. Porter Gregory C. Ficke T.P. Gwynn Lynne Bernabei, Esq.
John D. Woliver Deborah F. Webb David K. Martin George E. Pattison Andrew B. Dennison L. Kintner (3 copies of Attachment 1 to the SPDS Implementation Plan)
s WM. H. ZIMMER NUCLEAR POWER STATION - UNIT 1 DOCKET No. 50-358 f
REGULATORY GUIDE 1.97.
REVISION 2 NUREG-0737, SUPPLEMENT 1 REPORT FOR ZIMMER NUCLEAR POWER STATION September, 1983 The Cincinnati Cas & Electric Company i
Columbus & Southern Ohio Electric Company The Dayton Power and Light Company
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.s TABLE OF CONTENTS SECTION TITLE PAGE
1.0 INTRODUCTION
1-0 2.0 CG&E POSITION ON REGULATORY GUIDE 1.97 (REY. 2) FOR WM.
2-0 l
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ZIMMER NUCLEAR POWER STATION 3.0
SUMMARY
OF INFORMATION FOR CCMPLIANCE WITH REGULATORY GUIDE 3-0 1.97 REV. 2 4.0 JUSTIFICATIONS 4-0
5.0 REFERENCES
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1.0 INTRODUCTION
This report presents the Cincinnati Gas & Electric Company (CG&E) position on NRC Regulatory Guide 1.97, Revision 2, " Instrumentation for Light-Water Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident" for the Wm. H. Zimmer Nuclear Power Station.
It also describes how Zimmer meets the re-quirements contained in both Reg. Guide 1.97 and Supplement 1 of NUREG 0737, Section 6.0, "RG 1.97 Application to Emergency Response Facilities".
Zimmer instrumentation that will be used to monitor iype A, B, C, D and E variables listed on Table 1 of Regulatory Guide 1.97, Rev. 2, have been reviewed for compliance with Regulatory Guide requirements.
The results of the review are summarized in Section 3.0, Tables 3-1 through 3-5, of this report. Deviations from the guidance in Regulatory Guide 1.97, Rev.
2, are explicitly shown and supporting justification or alternatives are provided in Section 4.0.
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4 2.0 CG&E POSITION ON REGULATORY GUIDE 1.97 (P.EV. 2) FOR WM. H. ZIMMER NUCLEAR POWER STATION The intent of Regulatory Guide 1.97 Rev. 2 is to ensure that necessary and sufficient instrumentation exists at each nuclear power station for assessing plant and environmental conditions during and following an accident, as required by 10CFR Part 50, Appendix A and General Design Criteria 13,19 and 64. CG&E concurs with the intent of Reg.
Guide 1.97 Rev. 2.
Regulatory Guide 1.97 requirements will be implemented except in these instances in which deviations from the letter of the guide are justified technically.
In assessing RG 1.97, CG&E has drawn upon information contained in several applicable documents, such as BWR Owners Group Position on NRC RG 1.97 (Rev. 2), ANS 4.5, NUREG/CR-2100, NUREG 0737 Supplement 1, and BWR Owners Group Emergency Procedures Guidelines. CG&E does not believe that literal compliance with the provisions of the guide, because of their specific nature, is appropriate.
Some RG 1.97 requirements call.for excessive ranges or inappropriate categories.
Other requirements could adversely affect operator judgement under certain conditions.
CG&E has followed the criteria used by the NRC for establishing Category 1, 2 and 3 instruments.
The following compliance statement is applicable to the regulatory position defined in RG 1.97 (Rev. 2):
1.
Existing accident-monitoring instruments required to be environ-l mentally qualified will be qualified in accordance with the ac-ceptance criteria specified in Category II of NUREG-0588; " Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment" and the requirements contained in the NRC's Safety Evaluation Report of the Zimmer environmental qualification program (NRC letter dated July 8, 1983, Reference Docket No.
50-358). Replacement for existing unqualified equipment and new equipment added to comply with NUREG-0737 (THI Action Plan) will
'be qualified to NUREG-0588 Category I criteria. This will ensure that accident-monitoring instruments comply with 10CFR50.49 requirements 2.
Accident-monitoring instruments required to be seismically qualified will be qualified to the NRC Seismic Qualification Review Team (3QRT) criter,ia contained in the " Safety Evaluation Report Related to the Operation of Wm. H. Zimmer Nuclear Power Station" (NUREG-0528, Supplement No. 3, Section 3.10).
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To assist the control room operator, identification of instruments designated as Categories 1 and 2 for variable types A, B, and C '
will be made with due consideration of human factors engineering.
This position is taken to clarify the intent of RG 1.97, which' specified that these instruments be easily discerned for use during accident conditions. CG&E will comply with requirements contained in NRC's Control Room Design Review / Audit report forwarded to CG&E April 1,1981, and the CG&E response to the subject report ( Attachment L 1 of FSAR Appendix L). Implementation of Detailed Control Room Design Review is on hold pending completion of outstanding control room construction. The Detailed Control Room Design Review will incorporate methods and recommendations for identification of RG 1.97 types A, 8 and C instruments.
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SUMMARY
INFORMATION FOR COMPLIANCE WITH REG. GUIDE 1.97 REY. 2 3.0 CG&E p6aitions on the implementation of the variables listed in Table 1 of RG 1.97 and on the assignment and compliance with design and qualification criteria for the instrumentation proposed for their measurement is summarized in Tables 3-1, 3-2, 3-3, 3-4 and 3-5.
The variables are listed in Tables 3-1 through 3-5 in the same sequence used in Table 1 of RG 1.97; however, for convenience in cross-referencing entriee and supporting data, the variables are designated by the letter and number.
3.1 Zimmer Type A Variables Regulatory Guide 1.97, Revision 2, designates all type A variables as plant-specific, thereby defining none in particular. Type A variables are defined in RG 1.97 as "those variables to be monitored that provide primary information required to permit the control room operator to take specific manually controlled actions for which no automatic control is provided and that are required for safety systems tc accomplish their safety functions for design basis accident events".
Regulatory Guide 1.97 defines primary information as "information that is essential for the direct accomplishment of the specified safety functions". Variables associated with contingency actions that may be identified in written procedures are excluded from this definition of primary information.
The Zimmer type A variables listed below and shown on table 3-1 are based on the BWR Owners Group Emergency Pro-cedures Guidelines.
Variable A1 - RPV Pressure Operator Action: (1)
Depressurize RPV and maintain safe cooldown rate by any of several systems, such as main turbine by-pass valves, RCIC, and RWCU.
(2) Manually open one SRV to reduce pressure below SRV setpoint if any SRV is cycling.
Safety Function: (1) Core Cooling (2) Maintain reactor coolant system integrity.
Variable A2 - RPV Water Level Operator Action: Restore and maintain RPV water level.
Safety Function: Core Cooling.
3-0
Variable A3 - Suppression Pool Water Temperature Operator Action: (1) Operate available suppression pool cooling system when pool temperature exceeds normal operating limits.
(2) Scram reactor if temperature reaches limit for scram.
(3)
If suppression pool temperature cannot be maintained below the heat capacity temperature limit, maintain RPV pressure below the cor-responding limit.
(4) Attempt to close any stuck-open relief valve.
Safety function: (1) Maintain containment integrity.
(2) Maintain reactor coolant system integrity.
Variable A4 - Suppression Pool Water Level Operator Action: (1) Maintain suppression pool water level within normal operating limits.
(2)
Transfer RCIC suction from the condensate storage tank (CST) to the suppression pool in the event of high suppression pool level.
(3)
If suppression pool water level cannot be maintained below the suppression pool load
- limit, maintain RPV pressure below corresponding limit.
Safety Function: (1) Maintain containment integrity.
Variable AS: Drywell Pressure Operator Action:
Control primary containment pressuch by any of several systems, such as containment pressure control systems, suppression pool sprays, drywell sprays.
Safety Function: (1) Maintain containment integrity.
(2) Maintain reactor coolant system integrity.
3-1
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fp WM. H. ZDelER NUCLEAR POWER STATION St99tARY INFORMATION FOR COMPLIANCE WITH R.C.1.97 REV. 2 Table 3.1 TYPE A VARIABLES Page 1 of 1 ENVIRONMENTAL SEISMIC QUALITY POWER CR TSC BOF VARIABLE QUALIFICATION QUALIFICATION ASSURANCE REDUNDANCY RANCE SUPPLY DISPLAY IDCATION LOCATION SCHEDULE C009 TENTS ctor Will comply Will comply Will comply Yes 0-1500 peig IE Recorded Yes Yes Installed See llote l'
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pressure with with with and on CRT j
(Cat. 1) 10CFR50.49 SQRT Criteria 10CFR50. App.B f
i A-2 Reactor Will comply Will comply Will comply Yes
+60" to -150" IE 3 Channels Yes Yes Inst'alled See Notes water level with with with and recorded
-1, 2, and
-110" to -310" 1 Fuel zone FSAR Sect.
j (Cat. 1) 10CFR50.49 SQRT Criteria 10CFR50, App.B channel L.39 indicated All on CRT 4
A-3 Suppression Will comply Will comply Will comply Yes 30-230*F IE One channel Yes Yes Prior to See Note 1 pool bulk with with with recorded Fuel Load Both channels water temp-10CFR50.49 SQRT Criteria 10CFR50. App.B on CRT erature (Cat. 1)
A-4 88* NOL' I Suppression Will es y Will comply Will comply Yes t of ECCS IE One channel Yes Yes Prior to
.j pool water wit' with with suc* ion recorded Fuel Imad and FSAR lina to 5ft.
Both channels Sect. L.30.5 level 10CFF.
.9 SQRT Criteria 10CFR50. App.B above low on CRT l
(Cat. 1) 8 water level a
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Drywell W
comply Will comply Will comply Yes
-10 to 4140 IE One Channel Yes Yes Prior to pressure alth with with psig recorded and Fuel Load and FSAR other channel seeg, L,30,4
.FR50.49 SQRT Criteria 10CFR50. App.B j
(Cat. 1)
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WH. H. ZIMMER NUCLEAR POWER STATION S199tARY INFORMATION FOR COMPLIANCE WITH R.C.1.97 REV. 2 TYPE B VARIABLES Table 3.2 ENVIRONMENTAL SEISMIC QUALITY Ptaa 1 of2 POWER CR TSC EOF VARIABLE QUALIFICATION QUALIFICATION ASSURANCE REDUNDANCY RANCE SUPPLY DISPLAY lhCATION LOCATION SCNEDULE COMtENTS bl Neutron flua See Comment '
See Comment Will comply Yes SRM i
(Cat. 1)
Recorders.
Yes Yes See See Sect. 4.1 with W I'o 10 CPS IE Indicators.
Comment t
6 10CFR50, App.E APRM and CRT 0-125% PWR b2
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Control rod N/A N/A N/A N/A Full in h11able Indicating No No Installed l - FSAR q
position Full out Non-IE lights
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(Cat. 3) t
.nd Positfon 221.370 B-3 RCS soluble N/A N/A N/A N/A 100-1000 boron Station No No No Installed concen t ra tior ppm po,,,
(In chen lab)
See Note 12 (Cat. 3) l B-4 Coolant level in reactor Addressed in (Cat. 1)
Item A-2 B-5 i
BWR core No No No No No No No No No Do Not See Sect. 4.2 I
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(Cat. 1)
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WH. H. ZlHMER NUCLEAR POWER STATION EUMMARY INEVRMATION FOR COMPLIANCE WITH R.C. 1.97 REV. 2 TYPE B VARIABLES Table 3.2 Page 2 of 2 ENVIRONMENTEL SEISMIC QUALITY POWER CR TSC EOF VARIABLE QUALIFICATION QUALIFICATION ASSURANCE REDUNDANCY RANCE SUPPLY DISPLAY IDCATION LOCATION SCHEDULE C009HDrTS B-7 Drywell pressure Addressed in (Cat. 1)
Item A-5 u
B-8 Drywell sump level Addressed in (Cat. 1)
Item C-6 B-9 Primary containment Addressed in
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pressure Item A-5 (Cat. 1) i B-10 Primary Will comply Will comply Will comply See Closed IE Open - closed Yes.
Yes Prior To See Notes i
containment with with with Comments not closed indicating Fuel Load I&3 l
isolation 10CFR50.49 SQRT Criteria 10CFR50. App.E lights valve 4
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(Cat. 1)
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- 11
WM H. ZIMMER NUCLEAR POWER STATION
SUMMARY
INFORMATION FOR COMPLIANCE WITH R.C.1.97 REV. 2 TYPE D VARIABLES Table 3.4 Page 5 of 6 ENVIRONMENTAL SEISMIC QUALITY POWER CR TSC EOF VARIABLE QUALIFICATION QUALIFICATION ASSURANCE REDUNDANCY RANCE SUPPLY DISPLAY IDCATION IDCATION SCHEQULE C0tMENTS D-21 Cooling wate Will comply Will comply Will comply N/A 0-200*F teliable Indicated Yes Yes Insta'lled See Notes l&7 temperature with with with Non-IE to ESF systen 10CFR50.49 SQRT Criteria 10CFR50. App.B components (Cat. 2)
D-22 Cooling water Will comply Will comply Will comply N/A 0-6000 CPM Ig Indicated No No Installed See Notes flow to ESF with with with and on CRT
- 8. 13. & 15 system com-10CFR50.49 SQRT Criteria 10CFR50. App.B ponents (Cat. 2) l D-23 l
High radio-N/A tt/A N/A N/A Top to Botton Reliable Each tank on Yes Yes Installed activity Non-IE CRT liquid tank level (Cat. 3)
D-24 Emergency Will comply N/A Will comply N/A Open - closed IE Each damper Yes Yes Prior To See Notes 169 ventilation with with on CRT Fuel Load damper pos-10CER50.49 10CFR50. App.B ltion (Cat. 2) 1 i
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NOTES FOR
SUMMARY
INFORMATION FOR COMPLIANCE WITH R. G.1.97 REV. 2 TYPE A.B.C.D. AND E VARIABLES Note 1.
The currently installed equipment does not necessarily meet the environmental and/or seismic qualifications requirements.
Un-qualified components will be qualified, replaced, or justified for interim operation prior to fuel load.
Note 2.
Refer to FSAR Figure L.38-1 for reactor water level range cor-relation.
Note 3.
Redundant indication is not required on each redundant isolation valve. One set of open-closed indicating lights are provided for each remotely operated isolation valve.
Note 4.
The equipment has been specified to meet the environmental, seismic, and quality assurance requirements indicated.
Note 5.
RCS and/or sump samples requiring chloride content determination will be sent to an offsite facility for analysis.
Note 6.
Three (3) individual RHR flow loop indicators are provided.
~
Note 7.
The temperature indicated is for the service water system which is used for ESF system component cooling.
Note 8.
Indication is of service water flow to each RHR heat exchanger.
Note 9.
The indicating circuitry consists of qualified components and a IE power supply. Damper position indication is monitored via IE isolation devices on a non IE computer driven CRT.
Note 10. Status of the standby pcwer and other energy sources important to safety may be ascertained in the TSC and EOF by monitoring the position of the appropriate breakers. Since the breakers are l:
automatically controlled, a closed breaker indicates that adequate j
power is available.
Note 11. Flow to the main vent stack is terminated upon receipt of an ECCS signal and therefore this variable is not used for post accident 1.
monitoring.
I f
Note 12. Indicated range is obtained with 1000:1 dilution.
Note 13. The range indicated is greater than 110% of the system design flow.
1 i
Note 14. The currently installed equipment does not necessarily meet the seismic qualification requirement. Replacement equipment, however, will be qualified to the requirements indicated in accordance with
,I Sectior. 2 of this report.
I Note 15. The currently installed equipment does not necessarily meet the range, environmental and/or seismic qualification requirements.
Effected components will be qualified, replaced, or justified for interim operation prior to fuel load.
i.
P 4.0 JUSTIFICATIONS Justifications presented in the following sections, support the positions and deviations identified in section 3.0.
p 4
4.1 VARIABLE B1 - NEUTRON FLUX t'
Issue Definition: The measurement of neutron flux is specified as the key variable in monitoring the status of reactivity. Neutron flux is classified as a Type B variable, Category 1.
The specified range is 10 6 percent to 100 percent full power (SRM, APRH).
The stated purpose is function detection and accomplishment of mitigation.
Discussion:
Zimmer has four source range monitors (SRM) that have a range from 10-1 to 106 CPS and six average power range monitors ( APRM) i.
that have a range from 0-125% of reactor power.
The SRM receive power from two separate reliable 24 VDC buses and the APRM receive power from 120VAC supplies used for RPS power.
The SRM and APRM channels are monitored in the control room by indicators, recorders and CRT and are also available in the TSC and EOF facilities.
The neutren flux instrumentation is not classified as 1E and is not qualified to NUREG-0588 Cat. II requirements.
The SRM wiring, cables and connectors located within the drywell are 4
designed for continuous duty in the environmental conditions described in FSAR Table 3.11-1. The APRM is installed and operated in a control room environment as described in FSAR Table 3.11-1
==
Conclusion:==
The Zimmer neutron flux monitoring system is of a similar design as those used in most BWR'a. A Category I system that meets all RG 1.97 Rev. 2 requirements is an industry developmental item.
~
Zimmer will follow industry developmental activities and upgrade or replace the existing system when a fully qualified and proven neutron flux monitoring system becomes available.
4.2 VARIABLE B5 - BWR CORE 'HERM0 COUPLES CG&E's position for this variable has been provided in Appendix L, page L.31-2, of the Zimmer FSAR.
In summary, BWR core thermocouples will not be implemented pending their further development and consideration as requirements.
4-0
1 4.3 YARIABLE C1 RADI0 ACTIVITY CONCENTRATION OR RADIATION LEVEL IN CIRCULATING PRIMARY COOLANT Issue Definition:
Regulatory Guide 1.97 specifies that the status of the fuel cladding be monitored during and after an accident. The specified variable to accomplish thic monitoring is variable C1--radioactivity concentration or radiation level in circulating primary coolant. The range is given as "1/2 Tech Spec Limit to 100 Times Tech Spec Limit, R/hr." In Table 1 of RG 1.97, instrumentation for measuring variable C1 is designated as Category 1.
The purpose for monitoring this variable is given as " detection of breach,"
referring, in this case, to breach of fuel cladding.
Discussion: The usefulness of the information obtained by monitoring the radioactivity concentration or radiation level in the circulating primary coolant, in terms of helping the operator in his efforts to prevent and mitigate accidents, has not been substantiated.
The critical actions that must be taken to prevent and mitigate a gross breach of fuel cladding are (1) shut down the reactor and (2) maintain water level. Monitoring variable C1,as required in RG 1.97, will have i
no influence on either of these actions.
The purpose of this parameter falls in the category of "information that the barriers to release of radioactive material are being challenged" and "identi-fication of degraded conditions and their magnitude, so the operator can take actions that are available to mitigate the consequences."
i Additional operator actions to mitigate the consequences of fuel barriers being challenged, other than those based on Type A and B variables, have not been identified.
RG 1.97 specifies measurement of the radioactivity of the circulating l
primary coolant as the key variable in monitoring fuel cladding status during isolation of the NSSS.
The words " circulating primary coolant" are interpreted to mean coolant, or a representative sample of such coolant, that flows past the core.- A basic criterion for a valid measurement of the specified variable is that the coolant being monitored is coolant that is in active contact with the fuel, that is, flowing past the failed fuel. Monitoring the active coolant (or a sample thereof) is the dominant consideration.
The post-accident sampling system (PASAS) provides a representative sample which can be monitored.
The subject of concern in the RG 1.97 requirement is assumed to be an isolated NSSS that is shutdown.
This assumption is justified as l
current monitors in the condensor off-gas and main steam lines provide reliable and accurate information on the status of fuel cladding when the plant is not isolated. Further, the post-accident sampling system 4-1 l
(PASAS) will provide an accurate status of coolant radioactivity, and j
hence cladding status, once the PASAS is activated.
In the interim'*
between NSSS isolation and operation of the PASAS, monitoring of the i
primary con-tainment radiation and contal:went hydrogen will provide information on the status of the fuel cladding.
==
Conclusion:==
The designation of instrumentation for measuring variable i
C1 should be Category 3, because no planned operator actions are identified and no operator actions are anticipated based on this variable serving as the key variable.
Existing Category 3 instrumentation is adequate for monitoring fuel cladding status and therefore no additional primary coolant radioactivity instrumentation will be provided.
4.4 VARIABLE B8 DRYWELL SUMP LEVEL VARIABLE C6 DRYWELL DRAIN SlW LEVEL Issue Definition: - Regulatory Guide 1.97 requires Category 1 in-strumentation to monitor drywell sump level (variable 88) and drywell drain-sumps leval (variable C6.).
These designations refer to the drywell equipment and floor drain sump levels.
Category 1 in-strumentation indicates that the variable being monitored is a key variabis.
In RG 1.97, a key variable is defined as "...that single variable (or minimum number of variables) that most directly indicates the accomplishment of a safety function....."
The following discussion supports the CG&E alternative position that drywell aump level and drywell drain-sumps levels should be classified as Category 3 instrumentation.
Discussion:
The BWR Mark II drywell has two drain sumps. One drain is the equipment drain sump, which collects identified leakage; the other is the floor drain sump, which collects unidentified leakage.
Although the level of the drain sumps can be a direct indication of breach of the reactor coolant system pressure boundary, the indication is' ambiguous, because there is water flowing into these sumps during normal operation.
There are other instrumentation required by RG 1.97 that would indicate leakage in the drywell:
1.
Drywell pressure--variable B7, Category 1 2.
Drywell temperature--variable D7, Category 2 3.
Primary containment area radiation--variable CS Category 3 0
4-2
t 9
The drywell-sump level signal neither automatically initiates safety-related systems nor alerts the operator to the need to take safety-related actions. Both sumps have level detectors that provide only the following nonsafety indications in compliance with RG 1.45:
1.
Continuous level indication 2.
Rate of rise indication 3.
High-level alarm (starts first sump pump) j 4.
High-nigh level alarm (starts second sump-pump)
Regulatory Guide 1.97 requires instrumentation to function during and after an accident.
The drywell sump systems are deliberately isolated at the primary containment penetration upon receipt of an accident signal to establish containment integrity.
This fact renders the drywell sump level signal. irrelevant. Therefore, by design, drywell sump level instrumentation serves no accident-monitoring function.
The Emergency Procedure Guidelines use the RPV level and the drywell' pressure as entry conditions for the Level Control Guideline. A small 4
line break will cause the drywell pressure to increase before a noticeable increase in the sump level. Therefore, the drywell sumps will provide a " lagging" versus "early" indication of a leak.
Concluaion: Based on.the above considerations, the drywell sump level and drywell drain sump level instrumentation should be classified as I
Category 3, "high-quality off-the-shelf instrumentation," and no ad-ditional Category I instrumentation is requirsd.
4.5 VARIABLE C14 RADIATION EXPOSURE RATE Issue Definition:
Variable C14 is defined in Table 1 of RG 1.97 as rollows:
Radiation exposure rate'(inside buildings or areas, e.g.,
auxiliary building, fuel handling building, secondary containment, I
which are in direct contact with primary containment where pene-trations and hatches are located").
The reason for monitoring variable C14 is given as " Indication of Breach."
O_iscussion:
The use of radiation exposure rate monitors to detect breach or leakage through primary containment penetrations is impractical and unfeasible.
In general, radiation exposure rate l
readings in the secondary centainment will be largely a function of radioactivity in primary containment and in the fluids flowing in ECCS piping, which will cause direct radiation shine on the area monitors.
Also, because of the amount of piping and the number of penetrations and hatches and their widely scattered locations, local radiation exposure rate monitors could give ambiguous indications.
The proper way to detect breech of containment is by using the plant noble gas effluent monitors.
j 4-3
==
Conclusion:==
Using radiation exposure rate monitors to detect primary containmenf breach is neither feasible nor practical. Other means of breach detection are better suited to this function (as described above), are available.
Therefore, it is CG&E's position that this parameter not be implemented.
4.6 VARIABLE D3 SUPPRESSION CHAMBER SPRAY FLOW VARIABLE 08 DRYWELL SPRAY FLOW Issue Definition:
Regulatory Guide 1.97 specifies flow measurements or suppression chamber spray (SCS) (variable 03) and drywell spray (variable 08) for monitoring the operation of the primary containment spray systems. Instrumentation for measuring these variables is designated Category 2, with a range of 0 to 110 percent of design flow.
These flow relate spray flow to controlling pressure and temperature of the drywell and suppression chamber.
Discussion:
The drywell sprays can be used to control the pressure and temperature of the drywell.
The residual heat removal (RHR) system flow channels are used to monitor drywell ' flow.
The suppression pool sprays can be used to control the pressure and temperature in the suppression chamber. Instrumentation is provided to initiate the spray system automatically.
The suppression chamber spray can be operated in parallel with the drywell spray. The flowpath is determined by the position of the valve that is in the branch line that feeds the wetwell spray.
These valve positions are indicated in the control room. The effectiveness of these flows can be verified by pressure and temperature changes of the drywell and suppression chamber.
==
Conclusion:==
The current plant instrumentation, in conjunction with operating practice, provide for operator information that is suf-ficient for determining the existence of spray flows to the drywell l'
and suppression chamber without the use of a dedicated flow-measuring instrunent.
4.7 VARIABLE 017 SLCS FLOW Issue Definition: Regulatory Guide 1.97 specifies flow measurement of the Standby Liquid Control System (SLCS). The purpose is for moni-toring the operation of individual safety systems. Instrumentation for measuring this variable is designated as Category 2; the range is specified as 0 to 110 percent of design flow. The variable is related to flow into the reactor pressure vessel.
4-4
~
a Discussion:
The SLC system is manually initiated. Flow-measuring devices were not provided for this system. The pump-discharge header pressure, which is indicated in the control room, will indicate SLC pump operation.
Besider the discharge header pressure observation, the operator can verify the proper functioning of the SLCS by man-itoring the following:
1.
The decrease in the level of the boric acid storage tank.
2.
The reactivity change in the reactor as measured by neutron flux, i
3.
The motor contactor indicating lights.
The use of these indications is believed to be a valid alternative to SLCS flow indication.
==
Conclusion:==
Since monitoring the SLCS can be adequately accomplished by measuring variables other than the flow, it is CG&E's position that this parameter not be implemented 4.8 VARIABLE D18 SLCS STORAGE. TANK LEVEL Issue Definition: Regulatory Guide 1.97 lists standby liquid-control system (SLCS) storage-tank level as a Type D variable with category 2 design and qualification criteria.
Discussion:
The symptomatic Emergency Procedure Guidelines (EPG),
Revision 1, as presently approved do not consider ATWS conditions; however, the EPG committee of the BWR Owners Group has been developing a draft reactivity control guideline in which procedures are described for raising the reactor water level based on the amount of boron injected into the vessel, as indicated by the SLCS tank level.
Additionally, the operator is required to trip the SLCS pumps before a low SLCS tank level is reached, thereby preventing damage to the. pumps i
that would render them useless for future injections during the scenario.
Regarding the instrumentation category requirement for variable D18, RG 1.97 indicates that it is a key variable in monitoring SLC system operation.
Regulatory Guide 1.97 also states that in general, key Type D variables be designed and qualified to Category 2 requirements.
In applying these requirements of the Guide to this instrumentation, the following are noted:
1.
The current design basis for the SLCS assumes a need for an al-ternative method of reactivity control without a concurrent loss-of-coolant accident or high-energy line break.
The en-vironment in which the SLCS instrumentation must work is therefore a " mild" environment for qualification purposes.
t 4-5
2.
The current design basis for the SLCS recognizes that the system han err importance to safety that is less than the importance U safety of the reactor protection system and the engineered safeguards systems. Therefore, in accordance with the graded approach to quality assurance specified in RG 1.97, it is un-necessary to apply a full quality-assurance' program to this instrumentation.
Based on a graded approach to safety, this varisble is more ap-propriately considered a Category 3 variable.
==
Conclusion:==
It is CG&E's position SLCS storage-tank-level instrumen-tation meet Category 3 design and qualification criteria.
4.9 VARIABLE E2 REACTOR BUILDING OR SECONDARY CONTAINMENT AREA RADIATION Issue Definition:
Regulatory Guide 1.97 specifies that " reactor building or secondary containment area radition" (variable E2) should be monitored over the range of 10 1 to 104 R/h for Mark II containments. The classification for Mark II containment is Category II.
t t
Discussion:
As discussed in the variable C14 position statement (4.5), Secondary Containment Area Radiation is an inappropriate parameter to use to detect or assess primary containment leakage.
==
Conclusion:==
It is CG&E's position that the specified reactor building area radiation monitors should not be required.
4.10 VARIABLE E3 RADIATION EXPOSURE RATE Issue Definition:
Regulatory Guide.1.97 specifies in Table 1, variable E3, tnat radiation exposure rate (inside buildings or areas where access is required to service equipment important to safety) be l'
monitored over t.he range of 10 1 to 104 R/hr for detection of l.
significant releases, for release assessment, and for long-term l
surveillance.
Discussion:
In general, access is not required to any area of the secondary containment in order to service equipment important to safety in a post-accident situation.
If and when accessibility is l
re-established in the long term, it will be done by a combination of l'
portable radiation survey instruments and post-accident sampling of the secondary containment atmosphere. The existing lower-range area radiation rronitors would be used only in those instances in which radiation levels were very mild.
==
Conclusion:==
The Zimmer Plant specific design does not require eccess to a harsh environment area to service safety-related equipment during an accident. It is CG&E's position that this parameter should be reclassified as Category 3 and that existing area radiation monitors with a lower range meet the intent of RG 1.97.
1 4-6 8
5.0 REFERENCES
1.
BWR Owners Group Position en NRC Regulatory Guide 1.97, Revision 2 July 1982.
2.
Wm. H. Zimmer Nuclear Power Station Final Safety Analysis Report Revision 95.
3.
Wm. H. Zimmer Nuclear Power Station, NUREG 0588, Rev. 1 En-vironmental Qualification Report 4.
NUREG-0528 Supplement No. 3, " Safety Evaluation Report Related to the Operation of Wm.m H. Zimmer Nuclear Power Station" t
i l
l I
l 5-0 l
i WM. H. ZIMMER NUCLEAR POWER STATION - UNIT 1 DOCKET NO.
50-358 SAFETY PARAMETER DISPLAY SYSTEM IMPLEMENTATION PLAN September, 1983 l
l The Cincinnati Gas & Electric Company Columbus & Southern Ohio Electric Company The Dayton Power and Light Company 3-~
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E121EAL_215LEIEI1ON The general design and plans for emergency response f acil ities, systems, and components to be incorporated into the Wm. H.
Zimmer Nuclear Power Station have been previously described in submittsis dated June, 1981 and August, 1982.
A computer driven technical data acquisition and display system (DADS) has been designed and installed to enhance the emergency response capability of the station.
The purpose of the OADS is several fold.
to facilitate the assessment of plant steady state operating conditions during normal and pretransient times periods.
provide data for determination of transient initiating conditions.
provide data for analysis of plant dynamic behavior during a transient or accident.
To accomplish this purpose the OA05 is centered around the c apab il it y to accumulate, store and recall process data.
As a minimum, data for the most recent 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period will be avail able f rom redundant bulk storage discs.
The stored data will have sufficient time resolution to provide data without significant loss of information during a transient.
Each ddta point will be recorded at approximately five second time inter.vals.
The time resolution of transient data immed i atel y surrounding a reactor trip event is greatly enhanceo.by the General Electric Transient Anal ysis Recorder System (CETARS).
The GETAR S data, stored at scan rates of approximately 1,000/ scans /sec, can be transferred to the OADS on nagnetic tape.
The DAOS is designed to archive desired data without interrupting live data acquisition.
The DADS is designed to provide alpha-numeric and graphic respresentations of plant data.
The data displays are presented on colorgraphic video terminals, an electrostatic printer / plotter or an alpha-numeric l ine printer.' The OADS display functions are designed to all ow a flexible selection of data files for display and various display formats.
Plant parameters can be displayed
, in tables, on special logs, or on as a time history plot j
predefined graphic displ ays.
One predefined application of the OA05 is the Safety Paremeter
[
Displ ay Sys tem (5P05).
The SPDS is a computer software system, operating on the OADS computer, which provides parameter val idation, applies limit enecks, calculates certain derived parameters, and provides predefined colorgraphic video outputs to the SPDS d i spl ay s in the main control room, Technical Support Center and Offsite Emergency REV.00
Operations Facil ity.
It is designed to enhance the information presented to the reactor operator in a concise, rapid and understandable form.
SPOS GENERA: DESCRIPTION Three types of CRT displays are used in the SPOS:
Primary Oi s pl ay, Secondary Display and Message Summary.
The primary displ ay consists of six bar charts and is continuously disol ayed on the primary screen.
The secondary displays provide more detailed information on specific plant parameters and are tailored to the characteristics of the specific parameter being monitored.
There are ten secondary displays which may be called to the screen by the operator as needed.
The message summary displ ays all messages which have been issued over a given time period.
These messages appear as a list along with their time of issue.
This display may be called by the operator.
The studies conducted in support of the NSAC 21 report determined that the functions important to the safety of the public are:
Containment of Radioactivity (Radioactivity Release i
Control)
Fission Product Barrier Integrity Heat Transport Reactivity Control Containment of Radioactivity refers to the safety function of assuring that radioactive material does not escape from the plant to the environment.
Barrier Integrity refers to the maintenance of secure barriers as provided in the plant design to prevent the release of radioactivity.
The safety functions, Heat Transport and Reactivity Control, are special functions which must be performed to assure that Barrier Integrity is maintained.
The overall function
(" Containment of Radioactivity") is to orevent the release of radioactivity f rom the pl ant.
Monitoring radioactivity in both gas and liquid discharges to the environs will permit evaluation of how well this function is being performed.
Further, the radioactivity being restrained from rel ease by the primary cool ant system and containment barriers shoul d be mon itored.- These two barriers, al though present at all times, are not normally restraining radioactivity as the fuel cl adding barrier performs this function.
It is vital therefore to know if the primary coolant system and containment barriers are being called upon to contain radioactivity because the information could indicote that normally safe operations shoulc be conducted under more stringent controls.
For example, primary containment purging would not be desireable if nigh containment radioactivity l evel s exi sted.
_2_
REV.00
v Success in containing radioactivity is dependent on the integrity of the barriers provided to prevent rel ea se.
This "Darrier Integrity" safety function can be monitored and can be divided into two parts.
First, it must be determined that integrity does or does not exist at a given time.
That is, does the barrier now have leak paths for radioactivity?
Second, are there significant threats to the integrity of the barriers even though they may currentl y have integrity?
Two threats of this type have been identified which are of such major importance as to be considered separate safety functions.
They are " Heat Transport" and " Reactivity Control".
Heat Transport, as a safety function, is the principle that f ailure to remove core power or decay heat within a barrier will eventuall y and certainl y destroy the barrier's integrity.
This applies to the fuel cladding, primary coolant system and primary containment.
Simil a rl y, Reactivity Control as a safety function, is the recognition that reactivity excursions, or the failure to control steady state power level are intrinsic threats to barriers.
Failure to control steady state power level is characterized by a
'l evel of power production which exceeds the capability of the system to transport energy.
For instance, closure of the main steam line isolation valves virtually stops energy transport from the primary containment.
Failure to reduce power level by reactivity control leads to energy being stored in the containment (via the safety rel ief valves) faster than it can be removed and may l ead to eventual failure of the containment as a barrier to radioactivity release.
The SPOS design philosophy is to display several parameters to assure the safety functions described above are monitored.
The i
parameters selected to be continuously displ ayed on the primary displ ay screen accomplish the task.
The secondary displays are designed to enhance the information presented in the primary display.
Figure 1.b is a pictorial representation of the SPDS display.
l Pl ant safety is the principal concern of the display system and is shown as the apex of the hierarchy chart.
The second level shows the safety functions to be monitored to assure plant safety.
These safety functions are monitored by the six primary displ ay parameters as shown at the third tier.
f The secondary displays are shown on the fourth tier.
These j
displ ays further detail the primary parameters since the primary l
display only monitors some fraction of the safety function.
Cisol aying all the variables (signals) on the primary display l
would clutter the display.
Therefore, key variables have been l
selected to be monitored on the primary display in their order of importance to safety.
_3_
l REV.00 i
The primary display (Figure 2) consists of a title and a table divided into three columns.
The names of the parameters appear in the first column, the value appears in the second and the rate of change in the third.
The value and the rate of change are displayed both numerically and graphically:
the parameter value appears numerically with its units on top of a colored bar, and the rote of change direction is specified by a colored arrow with the numerical value and units to the right of the arrow.
If the rote of change is zero (or within some small band of zero) the word ZERO is displayed in the third column rather than an arrow and numerical value.
The color of both the bar and the rate of change arrow can be either green, yellow, or red.
Green signifies either " normal range" or " acceptable rate of change"; yellow means " caution";
and red signifies " danger".
Thus, during normal, steady operation, all of the bars would be green and the word "zero" would appear in all six rows of the third column.
This enables the operator to tell at a glance whether or not these parameters are at their expected values.
If there are detected instrument f ailures or conflictin) information that makes the selection of a value impossible or que s t ionabl e, the value bar will turn blue and be displ ayed near full scale or near downscale (deoending on the causes).
In addition, the nuaerical value will be repl aced by the word
" indeterminate" and the rate of change box will be cleared and repl aced by th'e word " indeterminate".
The primary display alerts the operator to messages generated in the signal selection process by illuminating a red rectangle in the corner of the box containing the parameter name with tne word " message" written in yellow inside tne r ec t angl e.
This is a signal for the operator to bring up a secondary display for more detailed information.
The SPOS secondary displays are described individually in the following sections.
Each secondary display providing information pertaining to a specific safety parameter, i.s described.
Egactor_ Status Reactivity Control is an important safaty function for overall plant safety.
The primary display monitors the reactor power obtained from the Average Power Range Monitor (APRM) System.
There are several reactor conditions where the APRM's may not relate sufficient information f or pl ant safety until conditions have degraded beyond the desi rabl e level.
The reactor status secondary display is intended to di spl ay the other information important for plant safety.
The essential part of this display is the information provided after a reactor shutdown or scram.
If a reactor sc ram occurs, the power is expected to resoond acco rd i ngl y. REV.00
This display would provide information in the event of a highly unl ikel y transient such as an Anticipated Transient Without Scram (ATWS).
The power display may be derived from the APRM system or from decay heat computed from ar analytical model.
Af ter a scram has occurred, the APRM readings are displayed without any diagnostic message until ten (10) seconds have elapsed.
This time delay allows for the decay of neutron popul ation to become effective.
If power has not sufficiently decreased after a scram, an ATWS condition may be present and the operator is alerted.
However, if a successful scram and reactor shutdown have occured as indicated by the APRM's, the Source Range Fonitors (SR*) are displayed to assure that local criticality does not exist.
Algorithm logic is present to evaluate the SRM reading and position following tne scram.
The algorithm uses a time delay of 2 minutes to determine if the reactor has been safely shut down.
After the time del ay, the SRM reading and SRM position diagnostic information are displ ayed.
In conjunction with reactor power, reactor core flow is displayed to assist the operator in assessing heet removal status.
The jet pump core flow measurement or the core grid delta pressure measurement correlated to flow, provide the source for the reactor core flow disolay.
To evaluate the adequacy of core cooling when the reactor water level decreases below the active fuel region when the core spray systein(s) are aveilable, General Electric has conducted numerous tests to demonstrate that adequate fuel bundle cooling is maintained from the High pressure Core Spray (HPCS) or the Low Pressure Core Spray (LPCS) systems.
The reactor status secondary display provides the core scray flow rates (s).
The reactor status displ ay (Figure 3) is in the form of a three column tabl e.
The first column contains the name of the parameter being monitored, the second column displays the parameter's current value, and the third column indicates the l
measurement units.
The parameters appearing in this table are:
Time of Reactor scram, current decay heat level (or power level if there is no scram),
core flow, high pressure core spray flow, low pressure core spray fl ow and the SRM readings.
There is, in addition, a region at the bottom of the disolay that is reserved for messages ger.erated in the l
algorithms that pertain to these parameters.
j Reactor Water Level Maintaining proper reactor water level assures adequate cooling of the reactor core and monitoring tnis essential parameter is vital for reactor safety (Figure 4).
Reactor l
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water level history provides valuable information to track water make-up systems being initiated during a transient event.
Al though some water level transients will occur to rapidly to permit operator action, the operator should be aware of any trends in water l evel.
Observation of trends will assist the operator in implementing appropriate action to control water level.
Review of the level history will assist the operator in verifying operation of automatic controls.
Any diagnositic messages appearing on this figure would automatically initiate the alert " message" indicator on the primary dicmlay corresponding to the reactor water level parameter.
The display features 10 minute, one hour and ten hour history displays.
This display consists of a title, a schematic figure of the reactor vessel and a graph of the water l evel history.
The lower portion of the display is dedicated to the diagnostic messages.
The vessel schematic includes the reactor core, steam se'parators and steam dryers.
The actual water level is shown in cyan ano the level of the water can thus be seen directly in rel ation to the major reactor vessel regions.
The water level graph is located directly to the right of the vessel schematic, so that the point on the graph which represents the pres *ent level is at the same neight on the screen as the top of the water in the schematic.
The groch shows the reactor water level history over a previous time period with the present l evel to the far right of the graph.
The history can be plotted for a ten minute, one hour or ten hour time period.
The graph is divided into five colored regions:
The green region maps the normal operating water level (low to high alarm level).
a l
Two yellow regions:
One from above the high al arm limit region to Le. vel a trip elevation, and one from low l ev el alarm limit region to the too of active fuel.
These indicate the caution Zone.
Two red regions:
one from the l evel 8 trip el ev at ion to the top of the vessel dome, and one from the top of the active fuel to the bottom of the core.
These indicate the danger zones.
l The abscissa on the left indicates the top of the core.
On the right, different level elevations have been indiccited.
l 1 0.EV.00 i
These levels correspond to the following system initiation el evat ions :
LEV. 1
= Automatic Depressurization (ADS) and LPCS LEV. 2
= Reactor Core Isol ation Cooling (RCIC),
HDCS and Main Steam Isolation Valve (MSIV)
LEV. 3
= Pea,ctor scram and Recirculation pump runback LEV. S
= Trip RCIC, HPCS, and Reactor Feeapumps and close main turbine stop valves.
The ordinate indicates the time in HR : MIN:S EC.
Cn the far right the present time of the display (the last time data was 3
avail abl e to the displ ay), and to the l ef t, the past times are shown.
The time-axis label also indicates which history is being displayed (10 min, I hour or 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> history).
The reactor pressure vessel figure shows different portions inside the vessel, mainly the core and the steam lines.
This is designed to show the current level of the water corresponding to the current tine, and it is linkea to the graph with a dashed line.
Tne loavsc cashed line is at the top of the core and the upper dashed line shows the main steam line l evel.
The water level in the reactor vessel shown in cyan changes dynamicall y with tne next measured water level.
In case of instrument f ailure or other events rel ated to the water level, a diagnostic message will appear in the dedicated region at the bottom of the d ispl ay.
If there is no diagnostic message, the screen indicates "NO MESSAGE" to indicate the system is functional thus precluding the possibility of misinterpretation of a system failure.
Uc to 3 lines of message space is available.
If there are more than three lines of messages, the operator can "page" to the next set of messages by depressing a designated keyboard switch.
l Should the operator continue paging after the last page of j
messages, the first page automaticall y reappears.
On the left uoper side of the message region, the number of pages and the l
di spl ayed message page number is shown so the operator knows the exact number of message pages, and which one is on the screen.
P Raac _to r _Ve ssel pressure The reactor vessel pressure display (Figure 5) presents the current reactor vessel pressure along ith the previous pressure nistory.
Although some cressure transients will occur too rapidly to permit operator action, the operator 1
l should be aware of any trends in reactor pressure.
'bservation of the trends will assist the operator in I
l l
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implementing appropriate action to control pressure.
- Also, review of the pressure history will allow the coerator to verify that automatic controls are properly controlling reactor pressure.
The reactor vessel pressure display consists of a title, a single graph with three color coded regions.
The green region extends from zero psig through the normal operating range to the lowest relief valve pressure setting.
The yell ow range extends from the lowest relief valve setting up to the highest safety valve pressure setting.
The red region extends above that pressure.
The lower portien of the display is dedicated to diagnostic messages.
The lef t hand abscissa indicates the reactor pressure from 0 to 1500 psig.
The right side abscissa presents the various pressure trips and from top to bottom are:
a.
High safety valve pressure setting b.
Low relief valve pressure setting c.
Main Steam Isolation Valve closure d.
LPCI/LPCS injection pressure The ordinate indicates the time in HR: MIN: SEC.
The present time is shown on the far right of the display (the last time data was available to the display) and to the lef t, the past times are snown.
The time-axis label also indicates wnich history is being displ ayed (10 min, I hour or 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> history).
On each display a block curve plot of reactor pressure versus time shows the trend of the reactor pressure witnin the ti me interval.
This plot is updated and changed dynamically as new data is available to the data base, (depending on the time scale selected).
The plot will only display valid data.
A discontinuity will occur in the pl ot when indeterminate data exists.
In case of instrument failure or other events rel ated to the reactor pressure, a diagnostic message will appear at the dedicated region at the bottom end of the d ispl ay.
If there are no messages, "NO MESSAGE" will appear to indicate the system is functional, thus precluding the possibility of misinterpretation of system failure.
If there are more than four lines of messages, the ooerator can "page" to the next set of messages by depressing a designated keyboard switch.
Should the operator continue Daning after the last page of messages the first page reappears automatically.
On the left upper side of the message portion, the ndmber of pages, and the displ ayed message page number is shown so the operator knows the exact number of message pages, and which one is on the screen.
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Esas12r_Y111sl_Esi_G.221ana_E12u The net coolant flow display (Figure 6) illustrates the mass flow to and from the reactor vessel.
This allows the operator to quickly determine which s ystems and components are operating correctly and which are not functioning.
A quick review of this display will give the operator information as follows:
Main steam line isolation 5/RV operation Leakage to the drywell sump Emergency Core Cooling System operation Operation of other water supply systems (Feedwater, RCIC and CRD cool ing).
This display is not intended to show the operator the reactor vessel water l evel.
Its purpose is to provide the operator with the information needed to oetermine the cause of changes in water level.
The display consists of a diagram of the reactor vessel with the core and steam separator shown and with the water in the vessel indicated Since this display is intended to illustrate flows to and from the vessel onl y, the water level indicated on this display d211_E21_SDAG22*
On the right ha n'd s i de o f the CRT, seven arrows pointing towards the vessel represent the systems from which water fl ows into the vessel; mass flow into the vessel from the system is shown inside the corresponding arrow.
The arrows are labeled top to bottom with the appropriate system name as follows:
1.
Feedwater 2.
HPCS - High Pressure Core Spray 3.
RCIC - Reactor Core Isolation Cooling 4.
Low Pressure Coolant Injection 5.
LPCS - Low Pressure Core Spray 6.
Standby Liquid Control System 7.
CRD Cooling On the left side of the CRT four arrows pointing away from the vessel represent the systems for which water fl ow s from i
i
_o_
?. E V. 0 0
v the vessel; mass' fl ow from the vessel is shown inside each arrow.
The arrows are labeled top to bottom with the appropriate system name, as follows:
1.
Safety re'ief valves 3.
RHR 4.
Containment sump pumping rate.
All flowrate values are di splayed in units of pounds per hour (L3/HR.)
Flowrates which are normally presented as GPM (such as the ECCS flows) are converted to LB/HR for the display.
This allows the operator to easily and accurately compare inflows and outflows (even if some are steam and some water) to evaluate net coolant flow.
The color of each arrow changes depending on the specific s i gn al and its status.
For all signals, an indeterminate value will result in the arrow turning blue and the printing of " indeterminate" in the arrow.
The arrow for all flows into the vessel are green when fl ow is present and red when no flow exists.
The arrow representing main steam line fl ow is green when fl ow is present and red when no flow exists.
However, for the S/RV flow and RHR outflow, the arrows are green when there is no flow and will turn red when fl ow is present.
When the drydell drain sump level is less than a to be specified level, cool ant system leakage is considered normal,
the leakage arrow is green and the word "icentified" appears in tne arrow.
Should the drywell drain sump level exceed the setpoint, the leakage arrow will c"ange to red, and the word
" unidentified" will appear.
The arrow will remain red until the sump level drops below the setpoint.
For other system arrows, the fl owrate will be orinted in the arrow when flow is present.
If there is no flow, the words "no flow" or "off" will be printed in the arrow.
Reactor Blowdown Prgdictor BWR plants take advantage of the large thermal capacitance of the suppression pool during pl ant transients requiring rel ief val ve actuation.
The discharge of each relief valve is piped to the suppression pool, where the steam is condensed.
resulting in a local temperature increase of the pool water, but a negl igible increase in the bulk containment pressure.
Most transients that result in relief valve actuations are of very short duration and have a small effect on the suppression REV.00
pool temperature.
However, three events present the potential for substantial energy releases to the suppression pool that could resul t in undesirably high pool temperature if timely correc tive action is not taken.
These events are:
(1) isol ation of the plant from the main condenser, (2) stuck open relief valve, and (3) Automatic Depressurization System (ADS) operation.
El eva ted suppression pool temperatures during extended relief vcive operation at high reactor pressures has become a major concern at BWR plants.
Under these conditions severe, continuous structural vibrations nignt be encountered.
The possibility of encountering the above condition is unlikely due to rigid technical specifications on the pool temperature during power operation and the large capacity for heat absorption.
This is supported by the fact that such an occurrence has never happened at a domestic BWR site.
However, since the possibility does exist when assumming limiting situations of peak service water temperature, technical specification pool temperatures, etc., it is important that potential situations leading to this phenomenon be recognized and procedural controls and instrumentation be utilized to avoid it.
There are two fundamental phenomena associated with rel ie f valve operation which can cause structural loadings.
They are the air clearing transient and the steam Jet condensation.
The air clearing transient phenomenon is associated with relief valve opening which causes a rapid compression uf the air initially in the discharge pipe.
This air is compressed until the water in the discharge pipe is expelled by the increasing pressure.
The highly compressed air is discharged into the suppression pool dater as an air bubble.
The bubole oscill ates and migrates to the woter free surface by the bu t ancy forces exerted on the bubble.
The pressure forces of tn s scillating Dubble are tranmitted to the structure throughout tne duration of the disturbance.
The pressure oscillations occur with every rel ief vave actuation and are independent of the suppression pool temperature.
The high temperature condensation vibration phenomenon is associated with the condensation process of the steam discharge jet into high temperature water.
This process occurs during steady state steam discharge.
The high temperature condensation vibration phenomenon occurs only with l
concurrent high pool temperature and high mass flux discharge.
l The high temperature condensation creates impul ses which are responsible far the uncesirable structural loads.
The structural loads are produced by the sudden recurring rapid condensation when a not layer of water surrounding the steam bubbles breaks up and is replaced by colder water (i.e.,
steam l
bubble collapse).
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f The majority of events resulting in S/R valve operation results in transitory relief valve actuation.
Tnese cause very mild pool temperature increases (less than 10 Deg F bulk rise).
Only threc types of occurrences have the potential for continuous energy dump to the pool via the relief valves and are of concern as previously stated.
Once the potential events of concern are recognized, any possible problems can be avoided by timel y, correct operator action without undue limitations on plant operation.
The Energency Procedure Guidelines (EPG's) provides guidelines for the timely and correct operator action.
However, the EPG's do not give guidance to tne operator when the procedural limit is exceeded.
This secondary display is intended to provide the operator additional guidance by illustrating the potential reactor vessel depressurization path for successful operation once the procedural limit is exceeded.
The reactor blowdown predictor display (Figure 7) incorporates a model to, indicate the increase in suppression pool temperature for a given number of safety / relief valves opened for a depressurization.
The display consists of a title, a graph, end a message area.
The graph itself is divided into three colored regions.
The red " danger" region is a zone in which there is a risk of unstable steam condensation and excessive containment oscill at ion loading.
The yellow region represents a transgression of the procedural l imits, Dut no imminent risk of damaging the containment.
The green region is the normal ooerating zone.
The message area at the bottom of the plot is used to describe the progress and/or results of the credictor.
The predictor model uses signal outputs from the algorithms to initialize the conditions for its blowdown simulation.
The predictor sill not begin unless there is a scram demand, a successful isolation, and onscale signals for both reactor pressure and suppression pool temperature.
If these conditions have been met, a bl ack cross will be drawn on the screen to indicote the current operating location and a black curve will be plotted on the colored background to represent the predicted blowdown path oy opening one saf ety/rel ief val ve.
If the path avoids the red region, a message will appear 6t the bottom of tne screen saying that containment risk has been avoided and the estimated blowdown time will appear on the graph.
The predictor then scans the pressura and temperature s ignal s, searching for a significant change in initial conditions before repeating its ca l c ul at i ons.
i -
PEv.00
If, on the other hand, the blowdown path enters the danger zone, the predictor will re-initialize itself and calculate another bl owdown path, opening another safety /rel ief valve.
The predictor model will continue this iterative process until a bl owdown method is calculated to remain outside the danger zone.
Descriptive messages are supplied to the operator during program operation in the message area.
The predictor will wait for a significant change in initial conditions before repeating its calculations, as long as it was able to avoid the danger zone on.its previous set of cal cul at ions.
Once, the initial conditions change, the process will be repeated starting with one S/RV open.
Orywell Pressure and TemEerature The drywell pressure and temperature display (Figure 8) allows the operator to Quickly check the drywell conditions.
The pressure and temperature history allows the operator to see gradual changes which may occur from a small primary coolant leak that the drywell coolers cannot compensate for.
A loss of drywell coolers would cause the drywell pressure and temperature to change which the operator can take appropriate action before limits are exceeded.
One safety concern is a high drywell temperature in the vicinity of tne reference legs of the reactor water level indication.
Such a condition can degrade the water l evel indication to read the water level erroneously.
There is an Emergency Procedure Guideline to assist the operator in this situation.
The display consists of a diagram of the containment (drywell and suppression chamber) on the left side of the screen.
To the rignt of this diagram i s a graph of the drywell pressure and temperature.
Both the pressure and temperature are plotted against time on the graph.'
The horizontal scale of the graph i s time and is common to both tne cressure and temperature plott.
Separate scales are given to the left of the graph for the two functions.
The pressure scale is marked in green and covers the range f rom
(-) 5 psig to 60 psig.
The pressure data is plotted on the graph as a green l ine.
The temperature scal e is marked in cyan, and the temperature is plotted as a cyan line.
The temperoture scale
' covers the range from 40 Deg F to 440 Deg F.
The pressure and temperature l imi ts are identified by yellow and red lines across the graph.
For the pressure, the yellow positive limit is set at 2 osig and the red l imit at 40 osig.
These li its are identified on the right side of the graph by green charac ters.
Al though negative l imits exi st for drywell REV.00 r-w
--v
- - - +
pressure, they are not marxed on the graoh.
Both the yellow and red limits would be so close to zero psig that they would clutter the graph and interfere with reading of the disol ay.
The temperature yellow limit is set at 150 Deg F,
and the red limit is set at 340 Deg F.
The limits are identified on the right side of the graph by cyan characters.
For the time history display either a ten minute, one hou r or ten hour history prior can be selecteo.
As with the other time history displays, the update rate varies with history period.
ElBV S$33gs_and_lM22ress ion Pool Tempera _tures The display (Figure 9) provides the operator with an illustration of the suppression pool temperature distribution in event of safety / relief valve steam discharge.
The local heating of the suppression pool resulting from S/RV steam discharge could cause condensation oscill ation without tne bulk suppression pool temperoture being at the critical temperature limit.
Therefore, this information can assist the operator to take action sur*' at initiating the Residual Hedt Removal syste.n to cool or mix the suppression pool water to mi tigate the local effects.
The operator can also take manual centrol and discharge other S/RV's to. provide equal distribution of steam addition throughout the suppression pool.
This display qual itativel y illustrates the temperature distribution around the suppression cool and the status of the sa fety/rel ief valves.
To the left of the screen is a top view-of the suppression pool showing the location of the eight,een thermocouples and thirteen quenchers.
The diagran also show the north and west directions.
The right hand part of the disolay contains two bar graphs which show the distribution of temperatures around the uuter and inner walls of tne pool.
These temperatures are displayed in order, clockwise fror west on each graph.
Each thermocouple and quencher is labeled with its own identification number.
The colors of the bars and thermocouales are either green, yellow or red, depending on the value of the local temperature.
There is a one to one correspondence between the thermocouple locations (represented as colored circles)
.a n d the bars tu the right.
The colors of both a bar and the thermocouple to which it corresponds are always the same. thus f acili tating identific ation of the lucation of local pool neating.
The quenchers, which are outl ined in wnite against the black background of the pool, fill with red when the corresocnding safety /rel ief valve is detected open.
The cuenchers return to their white.on black outl ine when tho valve is closed. 4EV.00
14Ry_(1((_S g gy Safety / relief valve lif t summary provides the operator information on which S/RV's lift and when they open and close.
Plant operation has shown that S/RV's performance can de]rade with open/close cycling.
Sventually the valve may resoond by sticking open.
The S/RV lift summary (Figure 10) keeps the operator informed of the safety /rel ief valve actuation over the prececing 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period.
This display also allows the ooerator to review the reactor pressure corresponding to S/RV opening and closing as compared to the specific S/AV relief setpoint.
The operator can readily observe a S/RV stuck open since the table will show that the valve is still oDen in the column for closure time.
Another possible use of this display is to observe the duration of the S/RV opening, if fuel failure nas occurred.
1he operator receives an indication of the steam mass added to the suppression pool.
He can correlate the suppression pool radioactivity with the S/RV discharges to qualitatively assess the possible degradation of the fuel cl ad barrier.
The display consists of a table with six columns.
From left to right the table headings are:
Valve identification number Oaening time Closing time Opening pressure Closing pressure R.elief oressure setting All S/RV lifts for the preceding 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> will bo listed in the table.
If no lifts nave occurred in that oeriod. a message will so state.
The l i.'ts will be listeJ in order based on the time of opening, with the most recently opened valve at the top of the list.
Valves which nave opened and closed will h3ve all six columns completed.
Valves which have not yet closed will nave the words "still open" disulayed in red instead of closing time and pressure.
Containment _ Status There are several containment barriers to protect against radioactive release.
The final barrier is tne containment itsel f.
For this reason, a display is needed to summariza all the important parameters to protect the containment.
d REv.00
Included in.the display are parameters which generally do not lead to individual parameter display or are more maaningful when presented in groups.
For example, containment isol ation and radiation throughout the containment.
For this raason, a tabular display was selected.
The containment status displ ay (Figure 12), like the reactor status display, is organized into three columns containing the parameter name, its value, and its units, respectively.
All of the parameters in this table remain displayed at all times.
They are:
RPV and containment isol ation status, drywell pressure, drywell temperature, drywell hydrogen concentration, drywell oxygen concentration, suppression pool water level, suppression pool bulk temperature anc offgas pre-treatment radiation.
All parameters are monitored and updated at the frequency required for the display.
The isol ation status colunn will either display the time of day at which isolation occurred. or the words "not i sol ated".
The readout for offgas radiation will also change to reflect the isolation status of the containment.
During normal operation, the value of radiation will be displayed.
When there is an isolation, the word
" isolated" will appear in parenthesis beside the value in the second column.
All other parameters are either displayec or noted as indeterminate.
$2ElliO310t_Atmo12h2rg_(ogirgl The containment atmosphere control display (Figure 13) provides the operator with a pictorial display of the containment oxygen and hydrogen concentrations in relation to the flammability limit curve.
This allows the operator to rapidl y visualize any potentially dangerous changes in containment atmosphere ;ak e-up, and to compensate for these changes by hydrogen recombination, drywell purging or drywell venting.
This display is a plot of the oxygen / hydrogen flammacility curve with the oxygen concentration increasing f rom lef t to right on the bottom axis, and hydrogen concentration increasing f rom bottom to top on the lef t axis.
The oxygen scale includes the region of zero to ten percent oxygen concentration, and the hydrogen scale includes zero to thi rty percent hydrogen.
(The instrument limits).
This results in the flammability curve enclosing most of the right side of the graph.
The graph is divided into green, yellow and red regions.
The green region covers the left half of the graph, from zero to five percent oxygen and zero to tnirty percent hydrogen.
Al tnough high levels of nydrogen are pot?nt iall y dangerou s, in this region the nydrogen can be reduced cirectl y without entering or approaching dangerously near tne flammable recicn. D.EV.00
In the right half of the graph, the region surrounding the fl ammability curve is the yellow zone.
This is the region in which a relatively.small change in hydrogen or oxygen concentration might result in entering the flamnability region.
The region encompassed by the flammatility curve and graph boundries is the red zone.
In this region, ignition of the containment atmosphere is possible.
The bottom portion of the displ ay is dedicated to the diagnostic messages.
The current containment oxygen / hydrogen concentration appears on the display as a olue cross.
If either or both of the signal s are of f scale (indeterminate) the blue cross will not be displayed, instead on the lower part of the green zone, the word " INDETERMINATE" followed by the signal name (oxygen and/or hydrogen) will be displayed.
The display is updated every 10 seconds and if there are no changes in the concentrations, the clue cross will stand still.
If there is a change in the concantration, the bl ue cross will show the current concentrations and an extracolation of this change will be displayed as a dashed olue l ine to its intersection boundary witn a blue pound sign
(#) with the current rate of change, and this crediction is displayed in the lower portion of green or yellow zone depending on the position of the cross.
This has the form
"... HOURS TO REACH 0" or "... MINUTES TO REACH THE FLAMMABLE ZONE" etc...
E2113gg_lummary The intent of this display is twofold.
First, it provides the operator a summary of messages wnich most likely indicate sensor failures, so he can be aware of the singl e and mul t ipl e failures existing.
Secondly, it provides maintenance personnel a summary output to scnedul e repair of the f ailures on the next scheduled outage.
The messages will exist as an alarm state on each secondary display for a time period of two hours.
It would De a nuisance to have a continuous alarm indication on the secondary display due to a failed instrument.
To avoid such a nuisance, the alarm messages are automatically transferreo to the message summary oy the software system of the SPOS.
It has intelligence to recognize the end of a day, month and year to transfer the message after two hours.
The time ceriod chosen is two hours to clear the secondary display alarm and transfer the alarm message to the summary table.
The messages are organized and displayed in a taole form (Cigure 14).
A maximum of eight messages c an be disol ayed on DEV.00
the screen at one given time.
On the top right corner of the screen, the total number of pages of messages and the current displayed page number are shown.
Each individual message is preceded by a time and date tag which compiles the exact initial starting time of the message, and the tag is formated as follows:
Time Hr: Min:Sec:
Date Po: Day:yr The color of the messages will be conventional yellow and the tabl e is updated every 60 seconds which will provide the opertor sufficient time to page through all the pages.
It is important to note that the operator can stay on a page as long as ne wants; that is, update occurs only when the 60 second interval is reached and the operator is finished examining one page of messages and decide to go to the next one.
Another thing to note is the initial starting time is not the time when the message is being tranferred, it is exactly two hours before the transfer time.
The first page of messages will be displayed automatically when the table is referenced.
In the event that no f ailure/ warning indication has existed for more than,two hours, "no messages" will be displayed in the table.
In cases when there are more than eight messages to oe displayed, the operator has the option to page to the next page of messages by selecting from the menu providing that he does not exceed the 60 second time limit for next update.
The first page would again be put up on tne screen wnen the operator decides to page after examining the last pace.
The opertor should be well aware of the total number of pages and the page he is examining as indicated on tne top right corner of the screen.
If there is a scram demand (and the ADRMs are reading less i
l than li power the top line of the taol e will indicate the time l
of day at which scram was. initiated and the second line will
(
disolay the current decaf heat l evel (a calcul ated value).
At l
the bottom of the plot, the SRM readings will be individaally di spl ayed.
If there is no scram (or the APRMs indicate more than 1*.. power) the top line is erased and the second line will di sol ay the current power level as indicated by the APRMs.
The SRM readings will al so be erased and the table will become physicall y smaller.
All of the other parameters on this l
display remain in the table at all times and are updated at s
l the update frequency of the display.
VALIDYNE DATA AC7UISITkON SYSTEM The OADS utilizes a Validyne Engineers.ng Corp.
HD-310 Hich Speec Data. Acquisition System to acquire the plant sensor beta being di spl ayed.
The 90-310 is A _distri0uted data acquisiti~on,
REV.00
['
sys tem (Figure 15).
The system, as configured at Zimmer, consists of 26 model MC370AD remote multiplexers, 4 model MX311 sub tultiplexers, 3 model DB327 signal splitters, and a model MR312 master receiver.
The remote multiplexers are fully 1E cualified devices, and therefore can be used to obtain data from safety related circuits.
The remote multiplexers are connected to the rest of the-system using IEEE 383 grade fire-retardant fiber optic cables.
The fiber optic cables serve to isolate the IE portions of the system from the non-1E portions of the system, as well as providing increased immunity from signal noise.
do to 25 signals are fed to each remote mul tipl exer di rec tl y from the plant sensors.
These signals are then digitized and multiplexed into a serial data stream.
The system consists of up to 16 links each of which may contain up to 8 sub-links.
Each remote multiplexer in the system is uniquel y identified by a i
combination of a link and sub-link, which serves as the harcware address of the remote mul tiplexer.
The address of any remote multiplexer is set by a unique pattern of dip switch settings on the mul tipl exer.
This information is transmitted al ong w i th the digitized signal to allow the upstream components to know what data they are receiving.
From the remote multiplexer, the data is received by a sub-mul tiplexer.
This device accepts inputs from up to 3 remote mul tipl exers and mul tipl exes them into a singl e data stream.
Each remote multiplexer attached to the sub-mul tiplexer must have the same link number and.different sub-link numbers.
From the sub-multiplexer, the data are routed to a signal splitter.
Here the signal is duplicated to allow use by more than one master receiver.
A cable f rom each signal splitter is ther connected to the appropriate connectors on the master receiver.
Here the data is checked for correct addresses, and the data is stored in randon access. memory.
This refresh crocess occurs a mininum of 300 times /second.
When the DADS computer wants to input some process data, it makes a request of tne master receiver.
The master receiver than transfers the latest data that it has in RAW to the computer.
If the user requests data from a mul tiplexer which does not currentl y exist on the system, or which is not functioning properly, the master receiver will set an error bit in any data transferred for the mul ti pl exer.
DAOS COMPUTER HARQfARE Many factors were taken into consiceration when tne choice of a computer vendor was made.
These include the history of the vendor in suoporting the equipment they sold, the upward compatability of all software developed for the system, the speed and flexibility of the system, the availabil ity of database management tools, and compatability with computer systems al ready installed at Zinmer.
PRIME Computer Inc was chosen as the vendor who orovided the best system for the recuirements at the Zimmer station.
The computer CPU being utilized is a model 750
~
minicomputer which has a soeed_of aoorox imatel y 1 million instructions per second.
The system was purchased with 4 Mbytes REV.00
of main memory, as well as 3 -- 300 Mbyte hard disks, for on-line storage of programs and data and a 75 ips 800/1600 bpi tape drive for archival storage of data on magnetic tape.
Output from the system is available either on a colorgrephics CRT or from one of two hardccpy devices.
The colorgraphics CRT's used are Ramtek model 6LOOA Col orgraphics Computer Terminal s.
There are currently 14 CRT's planned for the system.
Four of these.are to be installed in the Main Control Room, three are to be installed in the near site Emergency Operations Facility EOF, six are to be installed in the TSC Technical Support Center, and one is to be installed in the plant technical staff offices.
A Printronix 300 1pm line printer is used to provide hard copy of tabular outputs, and a Versatec Elec trostatic printer / Plotter is used to provide hard copy of graphical outputs.
DAOS SOFTWARE DATA FLGW DES (RIPTION The OAOS software consists of three sub-systems or tasks, 1) data acquisition, 2) data display, and 3)
SPOS' parameter validation and disolay.
Both 2 and 3 are indeoendent of each other, but are dependent upon task 1.
The following sections describe in more detail these sections.
The task of data acquisition is performed by two rather tightly coupl ed processes and a rather l oosely coupled process.
The first of these is responsible for coordinating requests for data from the Validyne data acquisition hardware and placing it in appropriate shared memory locations.
The second of these is responsible for converting the data from raw counts into engineering units, as well as performing alarm checking.
Tne converted and alarmed data is then placed in a common block, (or buffer), in main memory containing the last 20 minutes worth of data.
This data is then available to any other sub-system by simpl y using this common block.
It is from this common block that both the data display and SPOS validation & display tasks obtain their data.
This fact will be utilized to simplify the hardware testing of the 5005.
The last process is responsible for transfering the converted and alarmed data from memory to a database on disk for longer-term storage.
The data display task is responsible for cresenting all aata, excluding the validated SPOS parameters, to tne operator.
It is a series of programs which recalls the data previously stored by the data acquisition task and displ ays it to the user upon recuast.
This display can be in the form of either present or past time graph and tables, or present time system mimics.
However, in the case of the present time graphs and tables, historical data from the in-core common block is disolayed to allow the user to utilize the display in a more efficient fashion. REV.00
The SPOS validation C display task is responsible for producing validated parameter values using logical comparisons between various redundant signals in the plant, as well as displaying them to the user.
This is accomplished by a combination of programs.
The first performs the actual parameter validation.
It extracts the appropriate data from the internal common block of converted and alarmed signals previously developed by the data acouisition task.
This data is then processed through the validation algorithms to produce validated parameter values.
These parameter value are then stored in their own, unique common block in memory, along with message pointers.
These message pointers are used to allow the various output programs to inform the user of any abnormal conditions in the input data.
The actual SPOS parameter display is performed by a series of output programs, each of which retrieve all their information from the SPDS information common block in memory.
In this way, it is assured that all users of a display from the SDOS sub-system will see the same information.
The basis for selection of pla6C parameters, tabulation of the process instrumentation monitored, the methods of data combination and the methods of signal validation employed in the SPCS are described in the SIGNAL JUSTIFICATION AND ALGCRITHMS manual which is supplied for your review.
VERISICATION 05 HARDWARE AND SOFTWARE CONSTRUCTIO3 The SDOS hardware will be tested as part of the Data AcqJisition and Sisplay System during the preoperational testing phase of the Zim..ar Nuclear Dower Station.
The preoperational testing program begins with the turnover of individual components and systems from the constructor to the Cincinnati Gas & Electric Company f or tes ting.
As part of the construction program certain testing requirements are accomplished.
Preoperation11 test procedures contain prerequisite requirements that construction tests have been completed.
Construction test which are applicabl e to the O ADS system include.
a)
Verification of correct caole routing pulling, solicing and termination;, proper polarity E grounding.
b)
Megger and high potential testing.
c)
Checking control E interlock functions of instruments.
relays and control devices.
d)
Instrument and rel ay c al ibr a't ion.
e)
Initial equipment energization.
Plant process instruments are turned over for preoperational testing as part of the systems which they monitor or control. OEV.00
~
For the purposes of the SPOS system preoperational test the pl ant process sensors are assumed to be calibrated.
The subject instruments are verified to be calibrated prior to preoperational testing of it's associated system.
Once the preoperational test is completed the instruments are maintained in calibration as required by the technical specifications or the established periodic calibration frequency.
In this manner the OA05/SPOS is provided wi th properl y cal ibrated signal s.
The OADS/SPDS hardware will be preoperationall y tested in two parts.
1)
Point to point verification of each data point from the Validyne H0310 Master Receiver through the OADS computer system to a display device.
2)
Point to point verification of each process signal from the source instrument, through the data acquisition system to the master receiver.
As previously described, the validyne data acquisition system f eatures the fl exibil ity of ass ignabl e l ink and subl i nk names to the remote multiplexers and submultiplexers through the use of manual " dip" switches.
This fl exibility can be used to test the data stream from the master receiver through the computer processor to an output device for each data point using a test panel containing a remote multiplexer configured as a locally installed remote mul tipl exer with its appropriate link and sublink identification.
Using this test arrangement each data point will be driven with test instrumentation and tested for proper range output, correct units, correct alarm settings and general processing.
The logging function of the OAOS software will be utilized to observe proper outDut.
As part of instrument loop calibrations and preoperational tes ting o f plant systems associated with the instrument, each data. point will be verified for. wiring continuity and were l
appl icabl e, correct polarity.
This verification will be performed through observation of instrument channel response to know variations in process parameters on the OADS display device.
In this way the SPOS/ DADS data display system can be useo to support tes ting of pl ant systems and true response to system l
changes can be observed.
The total continuity of throughput from sensor to display is verified.
l SPDS VAL!DATION Validation of the Zimmer Nuclear' Power Station SPDS software will be implemented under the direction of the Independent Safety l
Review Group (ISRG) Supervisor.
Prior to initiation of tne sof tware validation program the SP05/0A05 software will be REV.00
varified by demanding each function from a users console ano documenting proper output response.
To ensure the validity of planned test cases, once tne validation program begins, no changes may be made to the SPDS software without the permission of the ISRG Supervisor.
Validation of the SPOS system will be accomplished in mudules as described below.
1)
The parameters selected as inputs to the SPDS will be reviewed for adequacy.
As part of tnis review the process instraments empl oyed in estaolishing the p3rameters are evaluated for aoecuacy and appropriateness.
2)
The SPCS primary display parameters will be evaluated against the entry points of the emergency operating procedures and emergency procedure guidelines to determine compatabil ity.
3)
The algorithms used in the SPDS for validation of input signals will be reviewed.
The review shall ensure that the selected value. where multiple process signal s are empl oyed, is a value tnat would not lead to a non-conservative operatcr action being taken.
The messages displayed to the operator from the validation algorithms will be evaluated f or cl ari ty, correctness and adequacy.
4)
The methodology used to generate the SPOS algorithm constant will be reviewed for adequacy.
The constants will be evaluated for reasonability.
5)
Each parameter val idation algorithm will be individually exercised.
Predetermined process signals will be supplied and the SPDS displ ays monitored to verify correct outputs and messages to the ocerator.
6)
Test cases will be run to ensure proper rate of change calculations by the SPOS for each parameter hich is accompanied by rate of change dispays.
7)
Test cases will be run to ensure that HI/ LOW alarm and HI/ LOW danger limits are properly detected and displayed.
Color changes will be observed on the SPOS displays when appropriate.
8)
All SPOS displays will be reviewed for clarity and the ease of understanding.
9)
The SPOS users manual will be reviewed for clarity, ease of use, and adequacy. REV.00
10 The SPDS and the users manual will be tested in a real time integrated fashion to identify those area which may be improved to provide smooth and effective use of the systems.
Following the SPOS validation program a comprehensive report will be generated which compiles the test results and documents areas of potential improvement.
Disposition of the recommended changes will be documented and approved in accordance with preoperational testing program administrative control s.
The Cincinnati Gas and Electric Company is presently participating in a BWR owners group committee to access the scope C feasibility of integrating SPDS displays with the BWR symptom based emergency procedure guidelines.
The results of this committee's work is not expected to be available within the next 12 months.
However, our partic ipation and the final committee recommendations will be evaluated for applicaoility and incorporated into the design of the Zimmer Nuclear Dower Station SPDS as deemed appropriate at that time.
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