ML20246J148
| ML20246J148 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 06/30/1989 |
| From: | Luke J STONE & WEBSTER ENGINEERING CORP. |
| To: | |
| Shared Package | |
| ML20246J098 | List: |
| References | |
| NUDOCS 8909050048 | |
| Download: ML20246J148 (237) | |
Text
{{#Wiki_filter:- _ _ _ _ _ - . 4L REACTOR COOLANT SYSTEM LEAKAGE DETECTION SYSTEMS NORTH ANNA UNITS 1 & 2 VIRGINIA POWER
/k 5- A JUNE 1989 Prepared by ad /# ,#b Reviewed by __
& </,l-- a c Approved by STONE & kIBSTER ENGINEERING CORPORATION BOSTON, MASSACHUSETTS 02107 9466-1493801-B1V 8909050048 890825 "
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l i h TABLE OF CONTENTS Page
- 1. INTRODUCTION . ........ . . .. .... . . .. . 1-1 1.1 Background . . . ..................... 1-1 1.2 Scope and Objectives . ... .... . .... . .. .. . 1-2 I 1.3 Acronyms & Abbreviations . . . . . ... . . .. . . . . . 1-2 l
- 2. DESCRIP7 ION OF LEAKAGE DETECTION REQUIREMENTS, SYSTEMS, INSTRUUt.NTATION & PROCEDURES. . . . . . . . . . . . ... 2-1 2.1 Presentation of Information. ... . . . . .. . . . . .. 2-1 2.2 Requirements . . . . . . . . . . . . . .. . .. . . ... 2-1 2.2.1 Regulatory Requirements. .. . . .. . .. . . . . . . . . 2-1 2.2.2 Standards and General Design Criteria. .. . ... . .. . 2-3 2.2.3 Technical Specifications . . . . . .. . . .. . .. . . . 2-3 2.2.4 General System Requirements. . . . . . . . . . . . . . . . 2-3 2.3 Leakage Detection Systems. . . . ... ...... . . . . 2-4 2.3.1 Identified Leakage . . . .. . . . . . . .. . . .... 2-4 2.3.2' Unidentified Leakage . . .... . .. . . . . . . . ... 2-4 2.3.3 Intersystem Leakage. . . . . . . . . . . . . . . . . . . . 2-5 2.4 Leakage Detection System Instrumentation . ... . . .. . 2-5 2.4.1 Identified Leakage Detection System Instrumentation. . .. 2-5 2.4.2 Unidentified Leakage Detection System Instrumentation. . . 2-6 2.4.3 Intersystem Leakage Detection System Instrumentation . . . 2-8 2.5 Instrument Loop Indication Accuracy Calculations . . ... 2-10 2.6 Containment Sump and Associated Instrumentation. . . .. . 2-11 2.6.1 Containment Sump . . . . . ... .. . . . . . . . ... . 2-11 2.6.2 Containment Sump Level Transmitters. . . . . . . . . . . . 2-11 2.6.3 Containment Sump Pumps . .. . .. . . ... .. . . .. . 2-13 2.6.4 Indicating Instrumentation . . . . . . . ... . . . ... 2-15 2.7 Procedures . ....................... 2-17 2.7.1 Alarm Response Procedures. . . . . . . . . . . . . . . .. 2-17 2.7.2 Periodic Test Procedures . . . . . . . . . . . . . . ... 2-18 2.7.3 Abnormal Operating Procedures. . . . . . . . . . . . .. . 2-19 2.7.4 Surveillance . . . . . . . . . . . . . .. . . . . . . . . 2-20
- 3.
SUMMARY
. . . . . . . . . . . . .. .. ..... .. .. . 3-1
- 4. ATTACHMENTS 4.1 References 4.2 Calculations 4.3 Technical Specifications 9466-1493801-B1V
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r [, , ?;, SECTION I p INTRODUCTION fN 1.1' Background N
. Virginia Power. su Criterion 4 (GDC-4)gtted a request in- compliance to Genera' Design-y for North Anna Units 1forand a licensing amendment to licenses NPF4 and NPF7 2, to exclude b effects of postulated main reactor protection against the dynamic na basis of primary system components / supports and piping. coolant pipe rup
? referred to as " Leak-Before-Break (LBB)."for the licensing nalyses ame L NRC . Generic Letter 84-04(2) provided a safety evaluationnghouse of the Westi u owners. group acceptable techn',submittal on Unresolved Safety Issue A-2 and concluded th cal basis exists i; ended pipe breaks in the Reactor Coolant System (RCS) primary i ..be met. - considered One in the design basis, provided that certain v.o t conditions leakage. condition is that an acceptable capability exists to detect RCS owners North Anna was not included in the plants reviewed under A-2 plant group, specifichowever, fracture individual mechanicsplants were encouraged to submit similar studies. May 12, 1986 GDC-4 was revised effective to utilize LBB technology to exclude dynamic effects of a postulated break of main reactor coolant loop. Use of LBB technology guidance used for evaluarequires that 'an assessment of leakage The . detection sys from NUREG 1061, Volume 3gg the leakage detection system is to be taken Regulatory Guide (R.G.) 1.45byhichemphasizesanassessmentwithrespectto An assessment of the North Anna Units I and 2 reaccor coolant leaka detection the system was prepared and submitted to the NRC as Attachment o p. p 1986 posed licensing amendment Serial No. 86-477A, dated November 6, with the The assessment compared the North Anna leakage detection system exceptionsnine givenRegulatory Guide in NRC Generic 1.4584-04. Letter design guidelines and considered the North Anna Unit 2, statements were made relative r
" detect a leakage rate of I gpia in four hours. o information The reviewer requested regarding leakage detection system instrumentation sensitivit , accuracy and leakages. y data and the methods currently employed in determining RCS i:
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1 o - L 2 m Scope 'and Objectives i 2 ,4 To: fulfill thel Reviewer's_ request and provide information on the RCS leahage [,** j detection systems, thc scope and objectives of this report include: (, , 1 Collating : int' ernal documents and' requirements related- to RCS- !
, ;1eakage detection systems. l y
q:r 12. - ' Identification of the systems and' related instrumentation 4
j
, -' . including .' accuracy and - sensitivity. data with emphasis on the )
,,7 ; containment sump. j w $ 3. Preparation ofm instrument _ loop : accuracy calculations for normal
. plant operating conditions.
- 14. ; Identification of , station procedures employed for identifying, quantifying, and. locating leakages.
c
- L3 ' Acronyms & Abbreviations
'. .AP' --
Abnormal Procedure i AR --
.A1 arm Response Procedure l CFR --
Code of Federal Regulations ! CONT - Containment
- CPM '- ' Counts per Minute CRO:- ' ' Control Room Operator 1
^
' CVCS'-' Chemical'and Volume Control System j GDC -
- General Design Criterion i GPM. - : Gallons per minute i IEEE - -Institute of Electrical and Electronics Engineers !
1Ri -~ . Insulation Resistance LBB .
, Leak-Before-Break LOCA~-- Loss of Coolant' Accident
- MSLB - Main Steam Line Break NRC - United States Nuclear Regulatory Commission PDTT - Primary Drain Transfer Tank
-'PORV - Power Operated Relief Valves t
PRT - Pressurizer Relief Tank PT - Periodic Test Procedure PWR- Pressurized Water Reactor PZR - Pressurizer i RAD - Radiation RCP - Reactor Coolant Pump RCPB - Reactor Coolant Pressure Boundary RCS -- Reactor Coolant System ! RG. - Regulatory Guide' RHR - Residual Heat Removal RWST - Refueling Water Storage Tank RX - Reactor SRSS - Square Root Sum of the Squares ; SSE - Safe Shutdown Earthquake !
>c. - SSOMI . Safety System Outage Modification Inspection SV -
Safety Valve
' UFSAR - Updated Final Safety Analysis Report VCT -
Volume Control Tank 9466-1493801-B1V 1-2 t
' " . 10 i' l.' - .____u____'_5 _.__m _..~m _ _ __
i SECTION 2 DESCRIPTION OF LEAKAGE DETECTION REQUIREMENTS, SYSTEMS, INSTRUMENTATION & PROCEDURES 2.1 Presentation of Information The following sections provide details on the leakage detection systems used at North Anna. First the regulatory requirements, general criteria and standards. have . been identified, followed by a general overview of each detection system. The instrumentation ured for each system to determine leakage is also described. In addition, calculations have been prepared for various parameters which provide specific -information on instrument loop accuracy and containment sump capacity. Finally, the station procedures that are used to detect, quantify, and locate leakages are described. 2.2 Requirements The regulatory requirements, applicable standards, general desigu criteria, technical specifications, and general system requirements are provided in the following sections. 2.2.1 Regulatory Requirements Generic Letter 84-04 & NUREG-1061, Vol. 3 NRC Generic Letter 84-04 and NUREG-1061, Vol. 3 state that, " Leakage detec-tion systems should be demonstrated to be sufficient to provide adequate margin to detect the leakage from the postulated through-wall flaw utilizing the gui: lance of Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems", with the exception that the seismic qualifica-
. tion of the airborne particulate radiation monitor is not necessary. In addition, at least one leakage detection system with a sensitivity capable of detecting 1 gallon per minute (gpm) in 4 hours be operable."
Regulatory Guide 1.45 Early detection of leakage in components of the reactor coolant pressure boundary (RCPB) is necessary to identify deteriorating or failed components and minimize the release of fission products. Consequently, Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems," published May 1973, describes acceptable methods to select leakage systems for the RCPB and states that "the source of reactor coolant leakage should be identified to the extent practical. Reactor coolant pressure boundary leakage detection and collection systems should be selected and designed tc, include the following: Design Guideline 1
" Leakage to the primary reactor containment from identified sources should be collected or otherwise isolated so that:
9466-1493801-B14' 2-1 )
g- x, ( Ld i' f
'.a'. . the flow raites 'are monitored separately from unidentified leakage, and h
- b. ~the total' flow rate.can be' established and monitored."
* . Design Guideline 2
" Leakage to the primary reector containment from unidentified sources should
.be collected and the flow rate monitored with an accuracy of one gallon per minute or better."
Design Guideline 3 I
"At least three separate detection methods should be ' employed and two of these methods should be (1) sump level and flow monitoring, and (2) airborne
. particulate radioactivity monitoring. The third method may be selected from the following:
- a. monitoring of condensate flow rate' from air coolers
- b. monitoring of airborne gaseous activity.
Humidity, temperature, or pressure monitoring of the containment atmosphere should be considered as alarms or indirect indication of leakage to the containment." , ' ' Design Guideline 4-
" Provisions should be made to monitor systems connected to' the - Reactor Coolant Pressure Boundary (RCPB) for signs of intersystem leakage. Methods should include radioactivity monitoring and indicators to show abnormal-water levels or flow in the affected area."
Design Guideline 5 "The sensitivity and response time of each leakage detection system in-regulatory position 3 abcve employed for unidentified leakage should be adequate to detect a leakage rate, or its equivalent, of'1' gpm in less .than I hour." Design Guideline 6 "The- leakage detection systems should be capable of performing their
- functions following seismic events that do not require plant shutdown. The airborne ' particulate radioactivity monitoring system should remain func-tional when subjected to the safe shutdown earthquake (SSE)."
Design Guideline 7
" Indicators and alarms for each leakage detection system should be provided in the main control. room. Procedures for converting various indications to 1
a common leakage equivalent should be available to the operators. The calibration of the indicators should account for needed independent variables." 9466-1493801-B1V 2-2
Design Guideline 8 "The leakage detection systems should be equipped with provi icas to readily permit testing for operability and calibration during plant operation." Design Guideline 9 "The _ technical specifications should include the limiting ' conditions for identified and unidentified leakage and address the availability of various types of instruments to assure adequate cove'. age at all times." 2.2.2 Standards and General Design Criteria IEEE Standard 279-1971 I6) To assure the continued reliability of the leakage detection systems, the equipment should comply with Paragraph 4.10 of IEEE Std. 279-1971, " Criteria for Protection Systeas for Nuclear Power Generating Stations," for tests and calibration which states "that capability shall be provided for testing and calibrating channels and the devices used to derive the final system output signal from the various channel signals. For those parts of the system where the required interval between testing will be less than the normal time interval between generating station shutdowns, there shall be capability for testing during power operation." General Design Criterion 30f7) General Design Criterion 30, " Quality of Reactor Coolant Pressure Boundary," of Appendix A to 10 CFR Part 50, " General Design Criteria fer Fuclear Power Plants," requires that means be provided for detecting and, to the extent practical, identifying the location of the source of reactor coolant leakage. 2.2.3 Technical Specifications (s) The requirement for detecting and identifying the location of leakage ia implemented via plant technical specifications. The technical specifica-tions limit both unidentified and identified leakage from the reactor coolant system. Since equipment cannot be perfectly leaktight, s.lowance is made for identified leakage from valve packing, shaft seals, and other equipment. Thus, even during normal operation, there may be some leakage. The Technical Specification sections for limiting conditions of operation, mode applicability, action statements and surveillance requirements which pertain to RCS leakage detection systems, operational leakage, and primary to secondary leakage detection systems are provided in Section 4.3. 2.2.4 General System Requirements The RCS leakage detection systems shall comply with the following require-ments: l 1. The systems shall be capable of identifying to the extent practi-l cal, the source of the reactor coolant leakage. l l 9466-1493801-B1V 2-3 I l l
l 2. The systems shall be capable of separately monitoring and collecting leakage from both identifiable ar.d unidentifiable sources.
- 3. The systems shall be adequately equipped with indicators and alarms for each leakage detection system in the main control room, and readily permit qualitative interpretations of such indicators.
- 4. The systems shall provide for the monitoring of systems connected to the RCPB for signs of intersystem leakage.
2.3 Leakage Detection Systems Several leakege detection systems are installed at North Anna to identify, quantify, and locate RCS leakage sources. These systems are used to detect identified, unidentified, and intersystem leakages, and details of each are described below: 2.3.1 Identified leakage A limited amount of leakage is expected from the RCPB and from auxiliary systems within the containment such as from valve stem packing glands, reactor coolant pump shaft seals, and other equipment that cannot practically be made 100 percent leaktight. The reactor vessel closure seals and safety and relief valves should not leak significantly; however, if leakage occurs via these paths or via pump and valve seals, it is detectable and collectable and, to the extent practical isolated from the containment atmosphere so as not to mask any potentially serious leak should it occur. These leakages known as " identified leakage". are piped to tanks so that the flow rates and changes in level can be established and monitored during plant operation. Identified sources of primary system Leakage in the reactor at North Anna are collected in the Primary Drain Transfer Tank (PDTT) 1(2)-DG-TK-1 and Pressurizer Relief Tank (PRT) 1(2)-RC-TK-2. The instrumentation associated with the PDTT and PRT is described in Section 2.4.1, 2.3.2 Unidentified Leakage Uncollected leakage to the containment atmosphere from sources such as valve stem packing glands and other sources that are not collected increases the humidity of the containment. The moisture removed from the atmosphere by air coolers together with any associated lxquid leakage to the containment is known as " unidentified leakage" and is collected in a sump where the flow rate can be established and monitored during plant operation. Instruments-tion associated with the containment atmosphere is described in Section 2.4.2. Small amounts of unidentified leakage are impractical to eliminate, but are reduced to small flow rates, less than 1 gpm, to permit the leakage detection systems to detect positively and rapidly a small increase in flow rate. Thus a small unidentified leakage rate that is of concern will not be masked by a larger acceptable identified leakage rate. Unidentified sources of leakage in the reactor containment are collected in a common sump. The instrumentation associated with the containment sump is described in Sections 2.4.2 and 2.6. 9466-1493801-BIV 2-4
2.3.3 Intersystem Leakage Substantial intersystem leakage from the RCPB to other systems across passive ba rriers and valves is not expected. However, when such leakages occur, alarms and detection methods are employed. Steam generator leakage is monitored to detect tube - or tubesheet leaks. Provisions are made to monitor systems connected to the Reactor Coolant Pressure Boundaries (RCPB) for signs of intersystem leakage. Radioactivity monitoring is used at North Anna to detect intersystem leakage. The condenser air ej ector, steam generator blowdown, component cooling water, and N-16 monitors indicate primary system leakage into interconnected systems. The instrumentation associated with these monitors is described in Section 2.4.3. 2.4 Leakage Detection System Instrumentation The instrumentation used to detect, identify, and quantify leakages using existing station procedures is provided in the following sections. The instrumentation is grouped by system according to tSe following categories:
- Identified Leakage
- Unidentified Leakage
- Intersytem Leakage 2.4.1 Identified Leakage Detection System Instrumentation Identified sources of primary system leakage in the reactor are collected in the:
- Primary Drain Transfer Tank (PDTT) 1(2)-DG-TK-1
- Pressurizer Relief Tank (PRT) 1(2)-RC-;I-2 The PDTT which is located on the 219-foot elevation of the reactor contain-ment receives leak-off from the following:
- 1. Various primary valve packings
- 2. No. 2 and No. 3 reactor coolant pump seal leak-off
- 3. Reactor vessel flange leakage detection system level Transmitter LT-DG101 (LT-DG201) on the PDTT provides the following ovtputs:
- 1. Indication at Vertical Board Section 1-1 (2-1) LI-DG101 (LI-DG201)
Westinghouse VX-252, Scale: 0-100%.
- 2. Computer input - Addressable point Y2015A Level switches LS-DG100 and LS-DG102 (LS-DG200 and LS-DG202) associated with level transmitter LT-DG101 (LT-DG201) iaitiate a PDTT Hi/Hi-Hi Level Alarm on Vertical Board Section 1-1 (2-1).
l Additional information on the instrumentation used for monitoring PDTT level j is provided in calculation no. 14938.01-C-9 in Secticn 4.2. l 9466-1493801-B1V 2-5
7, _,, -__ . , N
- The ' PRT$ 1ocated . on ~ the ' 252-foot elevation of the reactor. containment'
' receives. leakage from:
- 1. , Power Operated Relief Valves (PORVs).
- 2. Pressurizer Safety Valves (SVs)
~
- 3. Residual' Heat Removal,(RHR) System Relief' Valve
- 4. Chemical and Volume Control System (CVCS) Regenerative Heat Exchanger Relief Valves
! A level. transmitter LT-1470 (LT-2470) on the PRT provides the following outputs:
- 1. ' Indication at Vertical Board Section 1-1 (2-1) L1-1470 (LI-2470)-
Westinghouse VX-252, Scale: 0-100%
- 2. Computer input - Addressable point LO485A ,
-3. Annunciation of PRT High/ Low Level on Vertical Board ~ Section' l-1
~(2-1).
Additional information on PRT level instrumentation is provided in calcula-tion no. 14938.01-C-8 in Section 4.2. 2.4.2 -Unidentified Leakage Detection System Instrumentation _ Unidentified sources of system leakage in the' containment drate ty gravity to the' lowest level, and are collected in the containment sump which has a capacity of approximately 118 gallons. The containment sump is located just outside the reactor shield wall at elevation 216'-11" and receives drains from the following sources:
- 1. the Post-Accident Sampling Systems
- 2. the steam generator blowdown lines l
l 3. the reactor cavity
- 4. the containment drain trench
- 5. the in-core instrumentation room sump
- 6. the primary vent pot
- 7. overflow and drains from the neutron shield tank
- 8. the neutron shield tank heat exchangers 9466-3493801-B1V 2-6 l
4 4 - Level transmitter LT-DA110A (LT-DA210A)which monitors the sump provides the following outputs: 1. Indication at Vertical Board Section 1-3A - (2-3A) LI-DA110A.- (LI-DA210A) Westinghouse VX-252, Scale: 0-12".
'2 .
Annunciation. of RX Cont . Sump Hi Level on Vertical Board
.Section.1-3A (2-3A).
-3. On. increasing level, 'a containment sump pump will. start and provide a computer input signal _ (computer point - L2800D).
A andsecond provides: transmitter LT-DA110B (LT-DA210B) senses the' containment sump level
- 1. Indica' tion . at. Vertical Board Section 1-3B (2-3B) LI-DA110B
.(LI-DA210B) Westinghouse VX-252, Scale: 0-12".
2. Input to strip chart recorder LR-DA110B(LR-DA210B) located in the control room. A ' flow totalizer FTO-DA115 (FTO-DA215) is located in the containment sump. pump discharge line in the Auxiliary Building and is used'to determine the total volume of liquid pumped out of the sump. Additional information on the containment sump level monitoring instruments-
.Section tion is provided 4.2.
in calculation nos. 14938.01-C-10 and 14938.01-C-11 ini
- Instrumentation associated with the containment leakage monitoring system provides indirect indication of RCS leakage into the containment. This.
instrumentation includes temperature detectors at various locations within the containment. The temperature detectors sense containment ambient temperature and . provide indication in the main control room via ~ the plant computer. In addition, containment gas and particulate radiation monitors are provided to' detect respond rapidly to reactor coolant system leakage. The time required to. activ RCS leakage depends on the size of the break, reactor coolant UFSARg level, and containment background activity. Figure 5.2-16 in the contains the containment radiation gas and particulate monitor sensitivities. The plot depicts the time required for a given concentration-to reach a detector in relation to a specific leakage rate. This figute is documented in Calculation 'RP-11715-008-0 in Section 4.2. Additional calculations on radiation monitoring instrumentation are provided in Section 4.2. l' l
- l. 9466-1493801-B1V 2-7 l :.
2.4.3 Intersystem Leakage Detection System Instrumentation Intersystem leakage is detected by the following radiation monitors: Condenser Air Ejector Monitor The air discharged from the condenser by the air ejectors as they maintain a vacuum on the . condenser, is monitored for radioactivity. A high radio-activity would indicate a primary-to-secondary leak in the condensers or steam generators. Radiation monitor RM-SV-121 (RM-SV-221) continuously monitors the gaseous effluent from the condenser air ejectors. The monitor is located at the 286-foot elevation of the turbine building. Monitor RM-SV-121 (RM-SV-221) provides the following outputs:
- 1. Indication at radiation monitor cabinet in the Main Control Room RI-SV-121 (RI-SV-221) Westinghouse, range: 101 - 106 CPM 101 - 104 CPM 0 - 2.5 KVDC
- 2. Computer Input - Addressable point R0029A
- 3. Annunciation in the Main Control Room for Rad Monitor System Hi and Hi-Hi levels.
- 4. Input to chart recorder RR-100 (RR-200) at the radiation monitor cabinet.
Steam Generator Blowdown Monitor The b3owdown from each of the three steam generators is monitored for high radioactivity as as early warning signal that a primary-to-secondary leak has occurred in the steam generator. This monitoring actually occurs in Secondary Sampling System lines, which draw off a sample of the blowdown. Radiation Monitor RM-SS-122 (RM-SS-222) checks the liquid phase of steam generator 1-RC-E-IA (2-RC-E-1A) blowdown for primary to secondary system leakage. The monitor is located on the 292-foot elevation of the auxiliary building. Monitor RM-SS-122 (RM-SS-222) provides the following outputs:
- 1. Indication at radiation monitor cabinet in the Main Control Room RI-SS-122 (RI-SS-222) Westinghouse, range: 101- 106 CPM 101 - 10' CPM 0 - 2.5 KVDC
- 2. Computer Input - Addressable Point R0031A.
- 3. Annunciation in the Main Control Room for Rad Monitor System Hi and Hi-Hi levels.
- 4. Input to chart recorder RR-100 (RR-200) at the radiation monitor cabinet.
Radiation monitors RM-SS-123 (RM-SS-223) for steam generator 1-RC-E-1B (2-RC-E-1B) and RM-SS-124 (RM-SS-224) for steam generator 1-RC-E-IC (2-RC-E-1C) provide similar outputs as described above for monitor RR-SS-122 (RR-SS-222). 9466-1493801-B1V 2-8
,1t' fy Component Cooling Monitor y Cc:nponent . cooling ' water is monitored by taking ' samples from the component:
"I cooling water hest exchanger vent line'back to the head tank. The component cooling. water could be contaminated by in-leakage to ' the system from the l Reactor Coolant System.or'one of the other radioactive systems which give'up-their: heat to the . component Cooling System... Monitor'(RM-CC-120) is located at the 292-foot' elevation of the auxiliary building. Monitor (RM-CC-120) provides the following outputs:
- 1. ' In(ication at' the radiation monitor cabinet in the main control -
i ' room (RI-CC-120), Westinghouse, range: '101 - 106 CPM 101 - 104 CPM 0 - 2.5 KVDC
- 2. Annunciation in' the Main Control Room for Rad monitor system Hi and Hi-Hi levels.
- 3. Input to chart recorder (RR-150) at'the radiation monitor cabinet.
Additional information on the instrumentation used for monitoring inter-systera leakage - is provided in calculation nos. RP-11715-045-0, RP-11715-048-0, and RP-11715-050-0 in Section 4.2. N-16 Steam Line/ Steam Header Monitor The steamline from each of the three steam generators and the common steam header are monitored for radioactivity by. N-16 detectors. Detection of radioactivity ' by these monitors is indicative of a possible primary to secondary leak in the steam generators.- Monitor MS-RM-190 (MS-RM-290) for steam generator 1-RC-E-1A (2-RC-E-1A) is located -'on the 298-foot- elevation of the mechanical equipment roon. . Monitor MS-RM-190 (MS-RM-290) provides the following outputs:
- 1. Indication at the radiation monitor cabinet in the main control room RI-191 (RI-291), Merlin Gerin.
'2. Annunciation in the main control room for N-16 radiation detection and N-16 system trouble.
- 3. Input to chart recorder RM-RR-193 (RM-RR-293).
N-16 radiation monitors MS-RM-191 (MS-RM-291) for steam generator 1-RC-E-1B (2-RC-E-1B), MS-RM-192 (MS-Rh-292) for steam generator 1-RC-E-1C (2-RC-E-1C), and MS-RM-193 (MS-RM-293) for the main steam header provide similar outputs as described above for monitor MS-RM-190 (MS-RM-290). 9466-1493801-B1V 2-9
2.5 Instrument Loop Indication Accuracy Calculations Instrument loop indication accuracy calculations have been prepared to provide information on the accuracy associated with the various parameters utilized in identifying and quantifying leakages. Calculations were prepared using component / instrument numbers from North Anna Unit 2. Similar instrwnentatiop1 if nstalled (SRSS) method i on Unit 1. The square root sum of the squares was used in preparing the instrument loop indication accuracy calculations. The accuracy calculations are included in Section 4.2. The specific parameters are identified below:
- RCS Temperature (wide range)
RCS Temperature (narrow range) Accumulator Tank Level Pressurizer Relief Tank (PRT) Level
- Primary Drains Transfer Tank (PDTT) Level Containment Sump Level (Narrow Range)
Volume Control Tank (VCT) Level Pressurizer (PZR) Level a RCS Pressure (wide range and narrow range) The instrument loop indication accuracy calculations were prepared using normalenvironmentalcondigiopsasspecifiedintheNorthAnnaEnvironmental Zone Description, Rev. I Loss-of-coolant accident (LOCA) and main steam line break (MSLB) conditions were not considered. The transmitter ranges (calibrated span data) were obtained from referenced Test Loop Diagrams. The span for level transmitters has been considered to have already accounted for specific gravity of the fluid at normal operating temperatures. In 1988 gdpftion, calculation no. 16345-IC(B)-071, Rev. I dated August 5, for Comanche Peak Steam Electric Station, documents error inaccuracy for instruments inside containment due to low cable insulation resistance (IR) during a MSLB or LOCA resulting in a cable leakage error equivalent of 10.12 percent of span or less. Similar cable and instruments-tion are installed at North Anna. Due to the calculated values of leakage error which are relatively low and that the calculations in Section 4.2 were prepared for normal operating conditions (N0-LOCA/MSLB), the cable leakage error equivalent and errors associated with terminal blocks, Conax seals, and penetrations are considered to be very low. Their impact on the instrument loop indication accuracy calculations is considered negligible and has not been accounted for. Instrument loop indication accuracy data are provided in the calculations in Section 4.2. The results of each calculation document the total accuracy of each parameter (i.e. level, temperature), for the associated instrument loop as indicated by an indicator , recorder, or computer point for a typical 18 month fuel cycle. 9466-1493801-B1V 2-10
y i
?
g Thelresults of the calculations are summarized in table below: TABLE I i
' l'ndication.
9 , Calculation Parameter. Indication Device Accuracy j RCS_ Temperature (Wide Range) Recorder .
; 2.20%D Computer: Point T@4@6A 12.14%
RCS Temperature (Narrow Range). Indicator 15.56% Computer Point T@4@@A -14.30% Accumulator Tanks Level. Indicator- 13.14%. i '2-SI-TK-1A,7-1B, -1C PRT Level Indicator 13.07%' 4 ;2-RC-TK-2 ' Computer Point L@485A- 11.44% PDDT Level' Indicator 12.82% 2-DG-TK-1 ~ Computer Point Y2915A 21.03% Reactor Containment Sump Indicator 15.59% Level (Narrow Range) Reactor Containment Sump. Indicator 15.59% Level (Narrow Range) Recorder- 14.36%. VCT Level' Indicator ~13.81% 2-CH-TK-2 Computer Point L@ll2A 12.37%
- PZR- Level ~ Indicator 18.38%- ,
2-RC-E-2 .- Computer Point Lp480A .'17.22% RCS Pressure (Wide Range) Indicator 12.81% RCS Pressure (Narrow Range) = Indicator _ 16.28% Computer Point AP-Pp498A 11.02% 2.6 Containment Sump and Associated Instrumentation 2.6.1 Containment Sump l The containment sump is located outside the reactor shield wall at elevation l 216'-11" and has a capacity of 117.7 gallons per ' calculation no. 14938.01-S-15 in Section 4.2. Two pipe wells are located in the lower section of the containment sump. The combined volume of these pipes wells is 0.99 fts which is equivalent to 7.4 gallons. Refer to Figure 2.6-1 for details. 2.6.2 Containment-Sump Level Transmitters Two level transmitters LT-DA110A and LT-DA110B (I:T-DA210A and LT-DA210B) supplied by GEMS Sensors Division of Transamerica Delaval Inc. are installed to monitor the containment sump narrow range level. Transmitter LT-DA110A
-(LT-DA210A) provides output to indicator LI-DA110A (LI-DA210A) on the main control board, sump pump start and stop signal interlocks, computer input signal indicating pump start /stop and annunciation in the control room 9466-1493801-B1V 2-11 l-i
[ p t a when sump' level is high. Transmitter LT-DA110B (LT-DA210B) provides' output; to indicator .LI-DA110B - (LI-DA210B) on: the main control board. ' annunciation in ~ the control' room when sump level is high, and ' continuous output to strip chart recorder LR-DA110B (LR-DA210B). The GEMS transmitters provide almost - instantaneous . response to increasing level in the sump, with : the' exception' of the. inertia of the float whieh is negligible, and:the 20.5. inch deadband
'of the instrument (which corresponds to- 4.75 gallons). The deadband-of the level transmitter is attributable to the- physical arrangement.' of the reed switches housed within the central stem of thea transmitter. The . magnet i equipped float moves as the liquid level changes and magnetically actuates adjacent reed switches. The deadband of the transmitter will continuously-equate to a maximum of 0.5'ir.ch at all levels. .Using the data from calcula-tion 14938.01-$-15 based on area, a .1.0 inch level in: the sump equals 9.50 ~
gallons.' The deadband equivalent in terms - of gallons' can be determined 'as - follows:
-Deadband x Number of Gallons 1.0 inch sump level ,pg, ,g.Deadband Equivalent Gallons 0.5 inches x 9.50 Gallons 1.0 2nch sump level- = 4.75 Gallons As the sump level increases, float movement is proportional. However, the-reed switches activate in a . step-like progression. A similar progression occurs as the sump level decreases due to the reed switch arrangement
- described . above. With the transmitter resting on the bottom of the sump, the float .on the transmitter will- respond to changes which exceed 3.25 inches .above sump bottom as shown on manufacturer's drawing no.12050-7.26-28.(2ge Figure 2.6-2' displays transmitter output ' including -
the effects of deadband,'with respect to actual sump level for increasing / decreasing level conditions. The. time response of the transmitter including deadband is as follows: Deadband Equivalent (Gallons) = Time (Minutes) Flow Rate (GPM) Assuming'an'inleakage flow of I gpm:
-4.75 1.0 GPM Gallons = 4.75 Minutes Therefore, a 1 gpm leak rate into the sump will be sensed by the transmitter in 4.75 minutes. The transmitter response time for various leak rates is
-shown in Figure 2.6-3.
9466-1493801-B1V 2-12
2.6.3' Containment Sump Pumps The reactor containment sump pumps DA-P-4A and DA-P-4B automatically operate to maintain sump level and deliver approximately 125 gpm per pump. The operation of each pump is determined by the selector switch position. Each-pump switch features "0FF," " AUTO," and " HAND" positions. During normal plant operation the control switch is placed in the "AUT0" position which will allow the pump to run within an operating range according to the pump i i start and stop level setpoints. Both pumps discharge to the high level or low level waste drain header depending on the position of diverting valves - as lined up by the operator. The cont.airment sump pumps are automatica1'ly stopped on low level in the sump. The setpoints for the transmitter are shown on Figure 2.6-1 and are as follows: Pump Stop 3.0 inches of float travel Pump Start 5.375 inches of float travel High Level Alarm 6.75 inches of float travel Pump stop is set at 3.0 inches of ficat travel and since the float does not sense the water level until 3.25 inches above the bottom of the sump, this equates to 6.25 inches of water level. The number of gallons remaitling in the sump at the pump stop setpoint is as follows: k'ater level at pump stop setpoint (inches) x No..of gc11ons/per inch of sump (Calc. 14938.01-5-15)
+ no of gallons in 2 pipe wells.
6.25 inches x 9.50 gal + 7.4 gal = 66.77 gallons existing in sump at 1.0 inch pump stop setpoint Using the same formula (above) the number of gal'lons is the sump at the pump start and high level alarm setpoints is determined: Setpoint Gallons in Sump Pump Start 89.33 Pump Stop 66.77 High Level Alarm 102.39 Containment sump pump run time can be determined by the pump start and stop setpoints, pump output, and known inleakage into the sump. The pumps run in an operating band equivalent to 22.56' gallons (pump start-pump stop). Pump run time for a 1 gpm leak rate and single pump operation is as follows: No. of Gallons in Operating Band = Pump Run Time (minutes) IPump output (gpm)] - [inleakage (gpm)] 1 22.56 gal = .18 Minutes of Pump Run Time 125 gpm - 1.00 gpm The run time for single pump (P-4A or P-4B) operation and the time to fill sump between pump setpoints based on sump inleakage is shown in T ible 2. 9466-1493801-B1V 2-13
L p TABLE 2 L Time to Fill Sump Sump (Within Pump Inleakge Pump P-4A or P-4B Operating Limits) (gpm) Run Time (minutes) (minutes ) 0.1 0.18 225.60 0.2 0.18 112.80 L 0.3 0.18 75.20 0.4 0.18 56.40 0.5 0.18 45.12 0.6 0.18 37.60 0.7 0.18 32.23 0.8 0.18 28.20 0.9 0.18 25.07 1.0 0.18 22.56 1,5 0.18 15.04 2.0 0.18 11.28 2.5 0.18 9.02 3.0- 0.19 7.52 3.5 0.19 6.44 4.0 0.19 5.64 4.5 0.19 5.01 5.0 0.19 4.51 6.0 0.19 3.76 7.0 0.19 3.22 8.0 0.19 2.82 9.0 0.19 2.51 10.0 0.20 2.26 15.0 0.21 1.50 20.0 0.22 1.13 25.0 0.23 .90 30.0 0.24 .75 40.0 0.27 .56 50.0 0.30 .45 9466-1493801-B1V 2-14
s 2.6.4. Indicating Instrumentation The level indicating instruments used to monitor containment sump level are level indicators L1-DA110A (LI-DA210A), LI-DA110B (L1-DA210B), and-level recorder LR-DA110B (LR-DA210B). Using the ' data from calculation 14938.01-S-15, the actual sump gallons, the indicator and recorder reading can. be obtained. Calculation nos. 14938.01-C-9 and 14938.01-C-10 provide instrument indication accuracy data for monitoring sump level. The results are p;esented in Table 3. In addition, flow totalizer FTO-DA115 (FTO-DA215) is located in the contain-ment sump pump discharge line to determine total sump discharge flow. 9466-1493801-B1V 2-15
1
' TABLE 3 Actual Sump A tual audicator Indicator Recorder Level Sump Reading Accuracy Reading _ Recorder (inches) Gallons (inches) (inches) (inches) Accuracy 0 7.4 0 0 0 0-1 16.9 0 0 0 0 2 26.4 0 0 0 0 3 35.9 0 0 0 0 4 45.4 0.75 1.04 0.75 1.03 5 54.9 1.75 1.09 1.75 1.07-6 64.4 2.75 1.15 2.75 1.11 7 73.9 3.75 1.20 3.75 1.16 8 63.4 4.75 1.26 4.75 1.20 9 92.9 5.75 1.32 5.75 1.25 10 102.4 6.75 1.37 6.75 1.29 11 111.9 7.75 1.43 7.75 1.33 11.64 117.7 8.75 .48 8.75 1.38 i
9466-1493801-B1V 2-16
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(.. 1 l' 2.7 Procedures lP l The _ procedures for. detecting, locating'and. quantifying leaks are described-
. .f ,
in:the following section. 2.711 Alarm Response Procedures - y
' ?Alare response procedures are utilized at the. station to assist operations l by providing specific system information , such as probable cause and anti-cipated operator action. Because it.is essential that- leakage ~ detection systems have the- capability to detect significant RCPB degradation as soon after occurrence as practical to minimize the potential' for gross boundary l
failure, system alarms provide early ' warning signals . to permit s proper ' evaluation of . RCS . leakage. The alarms that provide. indication of RCS' leakage are listed below according to parameter: Level.
- RX cont sump Hi level
*- Incore Inst Room sump Hi/Hi-Hi level
- Pressurizer relief tank high-low level
- RCP 1A standpipe high level
- RCP IB standpipe high level.
-* RCP IC standpipe high level
- Primary drain transfer tank Hi/Hi-Hi level Temperature
- Pressurizer safety valve line high temperature
*- Pressurizer power relief line high temperature Pressure
- Containment partial pressure + 0.25 psi Ch. I-II
- Containment partial pressure + 0.1 psi Ch. I-II' Containment inter Hi-Hi Cont. Press. Ch. II, III, IV
- Containment Hi-Hi press Ch. I, II, III, IV Radiation
- Mn Stm A, B, C Aux Stm Loop Rad Mon Hi Rad
- Ma~Stm A, B, C Aux Stm Loop Rad Mon Rad Alert
- Rad Monitor System Hi Rad Level
- Rad Monitor System Hi-Hi Rad Level
- N-16 Rad Detection Flow
- RCP 1A-B-C Seal Leak High Flow l
l l' 9466-1493801-B1V 2-17
2.7.2 Periodic Test Procedures The periodic test procedures that are used to quantify and locate RCS leakages are:
- 1(2)-PT-46.21 " Components Affected by RCS Leakage" }
- 1(2)-PT-48, " Visual I of Reactor Coolant Pressure BoundaryComponents."(1gspection
- 1(2)-P " Reactor Coolant System Leak Rate" (band calcula-tion) g .2,
- 1(2)-PT-52.2 Coolant System Leak Rate" (computer calculation)g" Reactor The Periodic test procedures utilized to determine leak rates are:
1(2)-PT-52.2 " Reactor Coolant System Leak Rate" (Hand Calculation) and 1(2)-PT-52.2A " Reactor Coolant System Leak Rate" (Computer Calcula-
-tion)"
The purpose of 1(2)-PT-52.2 and 1(2)-PT-52.2A is to verify that reactor coolant system leakage is within limits set by technical specification Section 3./ 6.2 by performing a reactor coolant system water inventory balance in accordance with technical specification Section 4.4.6.2.1.d. 1(2)-PT-52.2 . is a hand calculation which utilizes data from addressable computer points and surveillance logs to verify that identified leakage is less than 10 gpm and unidentified leakage is less than I gpm. The test consists of performing a reactor coolant water inventory balance under steady-state conditions for a minimum time period of one hour between initial and final readings. For the test, pressurizer, volume control tank, and primary drains transfer tank levels are monitored. In addition, pressurizer relief tank and accumulator leakage is determined, and a RCS volume / temperature correction factor is applied. The identified and un-identified leak rate is calculated. Because these tests are conducted under steady-state conditions for specified time intervals, the accuracy of the instrumentation used to provide input data to the plant computer, indicators, and recorders will remain relatively constant. The instrument loop indication accuracy calculations in Section 4.2 provide accuracy values for a period of approxi-mately 18 months. The calculated accuracy of eat.h parameter represents a conservative value. With the accuracy of the instruments remaining constant, errors encountered during the performance of 1(2)-PT-52.2 or 1(2)-PT-52.2A would most likely be attributable to human error in reading the data. Prior station operating experience in determining the leak rates using 1(2)-PT-52.2 or 1(2)-PT-52.2A has proven to be very accurate. The accuracy of a one-hour leak rate test is estimated to be 1.2 gpm 95 percent of the time per 1(2)-PT-52.2 or 1(2)-PT-52.2A. Throughout the test period for 1(2)-PT-52.2 and 1(2)-PT-52.2A, the reactor coolant system must be maintained in steady state operation. Steady state operation shall be defined as: 9466-1493 DOI-B1V 2-18
l '.y 5 a l < 4 .
, 3
- 1. - Power L:tanges less' than- 1 percent of rated thermal power.
g 2. RCS pressure changes less than 15 psi when RCS pressure is greater p than 1000 psig. At1 low temperatures- and. pressures,. water compression is negligible; however, at 580'F and 2235 psig, a 110 psi, change in RCS pressure.would result in a 0.14 gpm error.in; a 60 minute leak rate, test. i- :3. RCS. temperature changes less than 12'F.
- 4. No changes in11etdown/ makeup systems (i.e., no blender makeups, charging pump starts,. ion exchangers placed into/ removed o from .
17 . service, etc). Y 5. Pressurizer level changes less than 2 percent.
- 6. No RCP starts or stops.
- 7. VCT H 2 blanket pressure must be maintained greater than or equal l' "
to 23 psig throughout this test to ensure the. RWST does not leak into the VCT. This maintains the leak rate determination con-servative.
- 8. Maintain Boron Recovery System Gas Stripper pressure stable betwiten 0-3 psig.
' Operators execute periodic test procedure 1(2)-PT-52.2 RCS leak rate (nand calculation) . at ' a frequency of at least ence every 21 days during operating modes'1, 2,;3, and 4 and periodic test procedure'1(2)-PT-52.2A RCS. leak rate (computer calculation) at a frequency of at least once ' every 72 hours. If the calculated unidcatified leak rate has increased more than .4 gpm 'since the last time 1(2)-PT-46.21 was . perfomed, then a walkdown of containment
~
i: is " initiated and 1(2)-PT-46.21 " Components; Affected by' RCS Leakage" is performed to . determine: and quantify thei source of any RCS leakage. The acceptance criteria for both ' 1(2)-PT-52.2 and 1(2)-PT-52.2A are such that the limits of 10 gpm identified leakage from the RCS and 1 gpm unidentified leakage have not been exceeded. With the RCS leakage greater than any,one of the limits describ:J above, appropriate action in - accordance with technical specification Section 3.4.6.2 must be taken.
- 2.7.'3 Abnormal Operating Procedures The abnormal procedures associated with RCS leakage include:
- 1(2)-AP-5.0 " Radiation Monitoring System" (29)
- 1(2)-AP-16 " Increasing Primary Plant Leakage"(20)
- 1(2)-AP-18 " Increasing Containment Pressuren (21)
- 1(2)-AP-24 " Minor Steam Generator Tube Leak"(22) l 1(2)-AP-5.0 " Radiation Monitoring System" is an abnormal procedure which provides indications, probable causes, immediate operator actions, and long-term ope rator actions for radiation monitoring alarms. Alarms 9466-1493801-B1V 2-19 l
1
c.- , .
. indicative of primary RCS leakage in'the containment are associated with the following monitors:
*' . Containment gas
*- Containment particulate Alarms that indicate intersystem leakage are associated with the following monitors:
4
- Condenser air ejector
.* Steam generator blowdown
*- Component cooling
- N-16 steam line/ steam header 1(2)-AP-16 " Increasing' Primary Plant leakage" provides a systematic approach to' locating, ' quantifying, and mitigating increasing primary plant' leakage.
This procedure will be executed when any of the following indications is
-present:
- Increasing charging flow
- Increasing frequency of VCT makeups
- Increasing frequency of pumping the PDTT
- More frequent operation of the-containment sump pumps
- Increasing area radiation
- Leak Rate 1(2)-PT-52.2'and 1(2)-PT-52.2A results increasing 1(2)-AP-18 " Increasing Containment Pressure" provides the indications of, probable causes for, and immediate and long-term actions to be tsken during an increase in containment pressure. Probable cause for the increase in pressure could be a primary coolant leak inside containment.
1(2)-AP-24 " Minor Steam Generator Tube Leak" provides indications of, and actions required to be taken for primary to secondary leakages. 2.7.4 Surveillance The Control ' Room Operator (CRO) surveillance sheets 1-LOG-4 (2-LOG-4)(23) contain handwritten data documenting the plant status. The surveillance sheets are filled out on a per shift basis every 8 hours. The applicable various parameters that may indicate RCS leakage include:
- 1. Rx vessel flange leakage temperature
- 2. PRT pressure
- 3. PRT level
- 4. PRT temperature
- 5. PRZ safety valve A line temperature
- 6. PR2 safety valve B line temperature
- 7. PRZ safety valve C line temperature
- 8. PRZ PORV line temperature
- 9. VCT level
- 10. Charging line flow
- 11. PDTT level
- 12. RCP A seal leakoff flow
- 13. RCP B seal leakoff flow 9466-1493801-B1V 2-20
h
^'
- 14. RCP C seal leakoff flow
- 15. Containment pressure
- 16. Containment temperature
- 17. Containment sump level
- 18. A SI accumulator level
- 19. -B SI accumulator level
- 20. C SI accumulator level
- 21. Radiation monitoring Auxiliary Building. surveillance sheets 1-LOG-6D (2-LOG-6D)(24) are also filled out once per shif t. The flow totalizer FTO-DA115 (FTO-DA215) in the containment sumo pump discharge line is read and the total flow is determined from the previous shift reading, i
l-l l l l~ 9466-1493801-B1V 2-21 c_- - _ -_
. ( .;; ;>. i p
g-- l (. 4 l5, , SECTION 3' q SUt91ARY Reactor coolant system leakage at North ' Anna Units 1 & 2 is detected using various' methods. Diverse' instrumentation provides the capability to detect - a 1 gpm leak within 4 hours, meeting the intent of R.G.1.45, with the i: : exceptions identified in Generic Letter 84-04.' The-. reactor coolant , system leak rate- periodic tests 1(2)+PT-$2.2 and 1(2)-PT-52.2A have proven to be reliable and accurate in determining identified and unidentified leak rates. A threshold value has 'been established to ensure that the results of the leak rate tests have not-increased more than 0.4 gpm since the last performance of 1(2)-PT-46.21
" Components Affected By.RCS Leakage." If the threshold has been exceeded, a walkdown is initiated to locate the source of leakage.
Indication of leakages are. sensed by. various systems, which~ provide detection for identified, unidentified, and intersystem leakage. The radiation monitoring system and increased operation of the containment sump pumps will provide operations with indication of increasing RCS leakage. Indication of excessive leakage will direct an operatorL to execute
- 1(2)-PT-52.2 or 1(2)-PT-52.2A to determine the quantity of leakage. The leak rate results may also. require that 1(2)-PT-46.21 be performed to determine the source of leakage.
System instrumentation in- conjunction with the station periodic test, surveillance, and abnormal operating procedures provide operations with. means to identify, locate, and quantify leakage. 9466-1493801-B1V 3-1 I
I l i SECTION 4 'I l i ATTACIDENTS I I 4,1 References J i 4.2 Calculations ) i 4.3 Technical Spet".fications l l
, j j
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1 I l I 1 i i 1 i l i l l 1 1 9467-1493801-B1V 4-1 l l l
4 , 4.1 ' References f - 1.
- j. Criterion 4 - Environmental and Missile Design Bases," Code of L
Federal Regulations, 10' Energy,' Part 50,- Appendix A,' General Services Administration, Washington, D.C. m
- 2. Eisenhut, D. G., " Safety Evaluation of Westinghouse Topical Reports Dealing with Elimination 'of Postulated Pipe Breaks in PWR Primary Main Loops (Generic Letter 84-04)," February, 1984, United States Nuclear Regulatory Commission, Washington, D.C.
- l. 3. NUREG-1061, Volume 3. Report of the United States Nuclear-Regulatory- Commission : Piping Review Committee; Evaluation of Potential For Pipe Breaks, November, 1984.
.4. " Reactor Coolant Pressure- Boundary Leakage ' Detection Systems,"'
Regulatory. Guide 1.45, May 1973, United States- Regn1 story Commission, Washington, D.C.
- 5. Virginia Power Letter to' Nuclear Regulatory Couunission, Proposed License Amendment - GDC4, Serial ' No. 86-477A, dated November 6, 1986.
- 6. IEEE Std. 279-1971, " Criteria ' for Protection Systems For Nuclear Power Generating Stations"
- 7. " Criterion 30 - Quality of Reactor Coolant Pressure Boundary," Code of Federal . Regulations, a0 Energy, Part 50, Appendix A, General Services Administration, Washington, D.C.
- 8. North Anna Technical Specifications, " Reactor Coolant System Leakage"
- 9. North Anna UFSAR Section 5.2.4 " Reactor Coolant Pressure Boundary Leakage Detection Systems."
- 10. Technical Paper "The Application of Statistical Methods in Evaluating the Accuracy of Analog Instruments and Systems," by C. S. Zalkind and F. G. Shinskey of the Foxboro Company.
- 11. North Anna Environmental Zone Description, Rev. 1.
- 12. Comanche Peak Steam Electric Station, Calculation No. 16345-IC(B)-071, dated August 5, 1988.
- 13. Manufacturer's Drawing No. 11715 (12050)-7 26-28A, GEMS Level Transmitter
- 14. North Anna Power Station Alara Response Procedures.
- 15. North Anna Power Station Periodic Test Procedure,1(2)-PT-46.21,
" Components Affected by RCS Leakage"
- 16. North Anna Power Station Periodic Test Procedure 1(2)-PT-48,
" Visual Inspection of Reactor Coolant Pressure Boundary Components."
9467-1493801-B1V 4-2 x < __._..-l-A.--_.-.---_.-__.._--_--..-_-._-_--.- - _ _ . _ - _ _ _ - _ - - - - - . . . _ . . - _ . _ . - _ . _ _ . - _ _ - _ _ - - _ -
b;; e ,
- 17. North Anna ~ Power Station Periodic Test ' Procedure 1(2)-PT-52.2,
" Reactor Coolant System Leak Rate."-(Hand calculation) -
'18. North ' Anna Power Station Periodic Test Procedure 1(2)-PT 52.2A,
'" Reactor Coolant System Leak Rate." (Computer calculatica)
- 19. North Anna Power _ Station Abnormal Procedure 1(2)-AP-5.0 " Radiation Monitoring System."
- 20. North Anna Power Station Abnormal' Procedure 1(2)-AP-16 " Excessive
' Primary Plant Leakage."
. 21. North Anna Powtr Station Abnormal Procedure 1(2)-AP-18" Increasing Containment Pressure."
- 22. ' North Anna Power Station ' Abnormal Procedure 1(2)-AP-24 " Minor Steam Generator Tube Leak."
- 23. North Anna Power Station Control Room Operator Surveillance Sheets 1-LOG-4 (2-LOG-4).
- 24. North Anna Power Station Auxiliary Building Surveillance Sheets, 1-LOG-6D (2-LOG-6D).
9467-1493801-B1V 4-3
)
i ! 4.2 Calculations Calculation No. Calculation Title ll.938.01-C Instrument loop' indication accuracy ' for Reactor Coolant System' Wide- Range Temperature TE-2410, TE-2420, TE-2430, TE-2413, TE-2423, TE-2433 14938.01-C-6 Instrument loop indication accuracy for Reactor Coolant System Narrow Range Temperature TE-2412B, TE-2412D, TE-2422B. TE-2422D, TE-2432B, TE-2432D 14938.01-C-7 Instrument loop indication accuracy for Accumulator Tanto @ $P-TK-1 A , 2-SI-TK-1B, and 2-SI-TK-1C. 14938.01 <-8 Instruser.t loop indication accuracy for Pressurizer Relief Tank 2-RC-TK-2 Level 14938.01-C-9 Instrument loop indication accuracy for Primary Drains Transfer Tank 2-DG-TK-1 Level 14938.01-C-10 Instrument loop indication accuracy for Reactor Containment Sump Level as sensed by Transmitter LT-DA210A. 14938.01-C-11 Instrument loop indication accuracy for Reactor Containment Sump Level as sensed by Transmitter LT-DA210B. 14938.01-C-12 Instrument loop indication accuracy for Volume Control Tank 2-CH-TK-2 Level. 14938.01-C-13 Instrument loop indication accuracy for Pressurizer 2-RC-E-2 Level - Channel I. 9467-1493801-B1V 4-4
Q 1culation No. Calculation Title 1493a.01-C-14 ~ Instrument loop indication . accr. racy for Reactor Coolant System Pressure (Wide range and . narrow range) loop 1. 14933.01-S-15 . Containment Sump Volume RP-11715-008-0 Containment Gas and Particulate Response Time as a Function of Leak Rate. RP-11715-045-0 - Steam Generator Blowdown Sample Monitor Setpoints. RP-11715-048-0 Condenser Air Ejector Monitor Setpoint. RP-11715-050-0 Component Cooling Water Monitor Setpoint. RP-11715-055-0 Containment Gas'and Particulate (normal operation) Setpoint. 9467-1493801-B1V 4-5
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Calculation No.
, s Calculation Title
-14938.01-C-5 Instrument: . loop indication accuracy- for Reactor Coolant System Wide Range : Temperature TE-2410,
,i :
. TE-2420, TE-2430, TE-2413,~TE-2423, TE-2433 I a.
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t 9467-1493801-B1V
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q STONE & UEBSTER ENGINEERING C07.PORATI N CALCULATION TITLE PAGE 3
*SEE INSTRUCTIONS ON REVERSE SIDE
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SCOPEOFW6RK WEED HOT LES RTD WIDE RANGE ACCURACY CALCULATION '
. . NORTH ANNA SWEC will prepare a calculation which provides the loop error associated with the referenced het leg RTD's (TE 1413, 1423, 1433, 2413, 2423, 2433, Weed model N90030-2B). The calculation will ir.clude the loop from the RTD's to the Validyne DAS. The error associated with the loop from the Validyne DA$ to the ICCM microprocessor is the responsibility of Westinghouse.
The error of the Ys11 dyne DAS itself was provided to Vir ~ Westinghouse via SWEC calculation 14938.10-C-6. Rev. 0,'ginia Power and That calculation should be plainly referenced in the new RTD calculation todated Decem ensure effect. that Westinghouse can easily combine those two pieces of the total loep The-new calculation should be in the same general format as the Va11 dyne calculation referenced above and the SWEC calculation Dav.1, performed for the Wide Ranga Pressure Transmitters last munLh.14938.10-C-7 The calculation must be provided to Virginia Power and Westinghouse (via appropriate cover letter) by November 30, 1987. After discussions with our Environmental Qualification group, Virginia Power hereby authorizes RTD's during accident SWEC to'use + 2.5'F as the inaccuracy band'for the Weed conditions. TE12 value was obtained from data in Weed Test Report SWRI 06-8680-003, Rev. I and from discussions with Mr. Ricky Gotcher of Weed on November 18, 1987. This value consists of a + 1.75'F error [~ associated with an RTO in a process fluid whose temperature is g7 eater than 650*F (which is possible during accident conditions) plus a +0.4'F error due te end-of-life aging and LOCA affects. of the two referenced temperature bands.A margin of about 16% was added to the sum : The normal environmental inaccuracy band to be used by SWEC is +1.5'F. This value consists of a +0.3*F trror associated with normal calibration, hysteresis, n'nd repeatability errors plus a +0.1'F/ year error associated with drift over time. (per QDR-N-8.24) so 9,5 x +0.1*F = +.95'F is the value used for ovara case drift. A margin of 20I was thew added to the sum of the te referenced; temperature bands, h(.rrwwP v$er A i CALC 14-9 5 7.o I- C , e II/n117 \ p { 135-WTD-1588
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L WITI __ Budget / Schedule Impacted Proj. Engr. by above: Concurrences
*cca (Required for any change in scope,
*Tes No schedule, requests for Document
*CE0'Neil-14 Tursover, technical bid evaluation.)
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- TLBardi Job Book (s)
- Vf* Mit/DSON October 1987
~
+
- ^ " - - - - - - ~ - - - - - _ _ _ _ . _ _ _ ___
NOTES OF TTLEFRONE CONVERSATION NORTP ANNA POWER STATION - UNITS 1 & 2 STONE & UERSTER ENGINEERING CORPORATION SURRYN- UNITS-1-4, PACE I OF i CALL DATE 4.7.99 ii O.HO. j 49 3 7.O l TIME,,, 4. , i o pm _ NP NO. ~
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$12.) 2. S F-7o 43 TASK l'"EM TITLE L EAsc SEFo/te BREAK
SUBJECT:
RC$ TEMP llo r d Cour; LE G RTO M et>c t_ duMSEAS
SUMMARY
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SCOPEOFW64K m WEED HOT 1.EG RTD WIDE RANGE ACCURACY CALCULATION '
. NORTH ANNA associated with the referenced het leg RTD's (TESWEC will prep 2433 Weed model h90030-2B). 1413. 1423, 1433, 2413, 2423 ,
RTD's to the Validyne DAS. The calculation will !r.clude the loop from the The error associated with the loop from the
- The error of the Validyne OAS itself was orovtded to V Westinghouse via SWEC calculation 14938.10-C-6. Rev. O, dated December 8 , 1986.
That calculation should be plainly referenced in the new RTD calculation to effect. ensure that Westinghouse can easily combine those two piects of the tota Valid The new calculation should be in the same general format as the Rev. yne calculation referenced above and the SWEC calculation 1, performed for the Wide Range Pressure Transmitters lanLLis mus:14938.10-C-7,
. The calculatten must be provided to Virginia appropriate cover letter) by November 30, 1987. Power and Westinghouse (via After discussions with our Environmental Qualification group, Virginia Power hereby RTD's during accidentauthorizes conditions. SWEC to'use + 2.5'F as the inaccuracy band for Test Report SWRI 06-8680-003, TEis value was obtained from data in Weed Gotcher of Weed on November 18Rev. I and from discussions with Mr. Ricky
[ 1987. associated with an RTD in a pro, cess fluid whose temperature is grea 650*F (which end-of-life is possible aging and during accident conditions) plus a +0.4'F error due LOCA effects. , a of the. two refarenced temperature bands.A margin of about 16% was added to the sum This value consists of a +0.3*F trror associated with nor hysteresis, a'nd repeatability errors plus a +0.1*F/ year error associated with drif t over time. (perdrift. case CDR-N-8.24), so 9.5 x +0.1'F = +.95'F is the value used A margin of 20Y was thei added to the sum of the two referenced; temperature bands. mW h r rM w"T A* i i u rwJ C ALC 14 9 3fr.o T- C- G
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STONF. E CEBSTER ENGINEERING COIPOR ATION CALCULATION TITLE PAGE
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STONE L DEBSTER ENGINEERING CORPO2ATION CALCULATION TITLE PAGE
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~
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. Calculation No.' - Calculation
Title:
14938.01-C-14 Instrument ' loop.' indication - accuracy for Reactor
- Coolant System Pressure'.(Wide range and narrow
(- range) loop 1. o.
,9467-1493801-B1V
I STONE & CESSTER ENGINEERli:G CORPORATION
-/ CALCULATION TITLE PAGE
*SEE INSTRUCTIONS ON REVERSE SIDE tso,oam nown
"^
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7 VI R Q tNI A POWER' MoRTa ANN A 'TcTAL PAG e4 Il C ALCULATION TITLE (Indicotive of the Objective)5 0 A CATEGORY (d
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)4938 o1 T %C- M aa.OI-c-14 "[A- 8[4
- APP ROVA LS - SIGN ATUR E & D ATE REV.NO. SUPERSEDES CONFIRM ATION OR NEW
- C ALC. NO.
- R EQUIRED M INDEPENDENT PREPARER (S)/DATE(S) REVIEWER (S)/DATE (S) REVIEW ER(S)/D ATE (S) C ALC NO. O R R EV. N O. YES NO
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^ CALCULATION
SUMMARY
,,,,73, ,, , ,.,,,,, a e,, ,,,,,,, ,,,,
me a i 49 3 6.o 1 - C o - or
''"'#"*'dT VIRGIMIA PcWeRf NORM AM A 70WEk STWnoMu cat 30av/ con clam
. sect /mLa TNsrRV mew 7 AHD chat 48 eL. A cc u a Acy p er;L RwTeK CootAM T' 59 5 TEM w t D C- RA 44 E Pgc-ssoRE (Loop.1}
OSJECTIVE OF CaLcWLafl05 To PRovi D E ToTn t Lo oP AccuRM_y (gen, g o am s @tur ofeRFnsc. ccH D
- Tiod S) Fo( Loop-l WIOe-2A M I
PResso a.e , wo xwe c.ame eaes s u c.c. calculation WETN00/aSSUMPfl0NS tieTHOD: T1-t E TOTAL Leo? e R eor Is sq V ARE Roof 0F SOM OF SQ U A R 6 G " C F A \ L C-RQO R. S . hU M l# rH G _Tu ni c Aro R 52 e AD A6m T7 tw aoR. w n t GC ALG e6 R A I crit L] A o ce_D r o Terat L o o P eE R o% (' ( sounces or oaTa/sevations REFER ro Re pc-Re# ce-5 e9 ' PAG E 4, y goQin AL Ghlvi G.0 N M G4 7 c.oM 01Tg eQ pe( Rcs PC2MN4 L.,f-$
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CALCULATION IDENTIFICATION NUMBER
< _ . J.O. OR W.O. NO. DIVISION G GROUP - CALCUL ATION NO.' OPTIONALTASK CODE PAGE 3 14 9 5 8 J o 4 ' I k C- 144SS Ol-C-M M/4 '
1-8 TABLE of CONTst4TS L'D EscR.W Tiet4 'PAos-
-' . TlYLE PAGE l-
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.I TA B LE OF CONTE NTS 3 43 REFeRENc.e5 4-5
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- 29 CoMPoTGA IN PuT DATA 11 f4 Cr c A A.0 ER Ro K CHC ) 12-f$ U MMING AMP Li fic-R c'N 5 A 3 L ATA 13
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STC::E G CESSTER ENGINEECl!.;G CCCPOR ATION CALCULATION SHEET
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STCIE OCE85TER ENGl EECING COCP30 ATIDO CALCULATION SHEET. A spic as CALCULATION IDENTIFICATION NUMBER J.D. O R W.O. NO. OlVISION & GROUP . CALCUL ATION NO.. OPTIONAL TASK CODE PAGE O e i 4 ct 3 8 . o t- T. t, C. 14818 6 ol-C-f 4- M/A 8 TR ANS MITTGR ' A ccO RA C.V '. (\/47 ) 5 vA r = 1 o 25 */. OF cA-u G A A reo sPMJ.
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H STCNE C.CESSTER ENGINEERI':S CCCPORATION s CALCULATION SHEET a mis m . CALCULATION IDENTIFICATION NUMBER J.O. O R W.O. NO. DIVISION & GROUP CALCUL ATION NO.' OPTIONAL TASK CODE PAGE LQ. - x 14 Cl '6 L 01~ I(c. 14 9 s s. ol -c-It t4/A 2 TRAM S M iTT6 R. CV e-R.P R e S s U R G. E F F E.CT ,(OPE d i-
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. STCNE G CESSTER ENGINEERING CORPORATION CALCULATION SHEET CALCULATION IDENTIFICATION NUMSER J.O. O R W.O. NO. OlVISION & GROUP CALCUL ATION NO. OPTIONALTASK CODE PA'E G IO 14 9 3 8. o t 2(c M938.ot- e-14 8lA LO o P Powl6R Su pp t i/ IsoL ATOR 3)AT A
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MANU Fa-cru RE:R : viesnac Roose Six Le / tuoDet. : 2.8 's7 Atz.Go 2
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- STCNE O CEBSTcR ENGINEE;; LNG COCPOR ATION CALCULATION SHEET a nno. l CALCULATION IDENTIFICATION NUMBER J.0. 0 R W.O. NO. OlVISION O GROUP CALCUL ATION NO. OPTIONAL TASK CODE PAGE Ib 14cl 3 8 . o t' I ( c. 149 3 8.ot -e-H s/g i
3 CAtt BR AT i NG Equi PMer4T En RoR . (C E) 2 o . 5 */, of Speu usordD
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't SU M M I N G ~ AM P LI F1 & R ( t4 5 A ) D ATA MF4 R : v4 C-S TI N G HO O.5 E ( REF
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r . STOCE O CESSTER ENGl%EECl%G CORP 7,OATIDCJ L CALCULATION SHEET b a nno es C ALCULATION IDENTIFICATION NUMBER I L J.O. O R W.O. NO. OlVl% ION & GROUP CALCUL ATION NO. OPTIONAL TASK CODE PAGE I4
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STOZE G CEBSTER E!!GINEECl3G CORPO2ATION CALCULATION SHEET s w. . CALCULATION IDENTIFICATION NUMBER J.O. O R W.O. NO. DIVISION D GROUP CALCUL ATION NO. OPTIONALTASK CODE PAGE M L i 49 38. e t .I ( c. 14.al a s.ot e-t 4 s/A 8 r , s u TO TA L LcoP em et (VAT )t + r(Te ) + ( PsET 7
)4 (D v4 8 DE RANGE
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+ (oper )t+(t-9)t+(ce[+(N5 )L A
+(IEA)c _
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= I ( b 5_) +
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35 i,_c f._ '/' , 2 -[,
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h STONE G CESSTER ENGINEEftl%G C^RP?,OATION CALCULATION SHEET a nato es CALCULATION IDENTIFICATION NUMBER !. J.O. O R W.O. NO. DIVISION & GROUP CALCUL ATION NO. OPTIONAL TASK CODE PAGE 1 s ]49 56.0% J- ( C,, 149 36. Ol-C-14 N/A r , 2 y 1]&oj y*L$* 9=1{v^xf+ CTe h(Pseb N )' - r t z .: 'r~
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+
= ._
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t. 1 W D IC /t-T O kC-A-DASIL,lT 7 ER0o 2 { o r.- m* t, M AusFs7to55t0V siNcc $ M a u.c- sT 'Divis s e d T5 - *T."L. M 00"E
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4 STO::E O CEBSTER EC68 EERi%G CORP 3':ATIO N CALCULATION SHEET, a mne s CALCULATION IDENTIFICATION NUMBER J.0, O R W.O. NO. DIVISION G GROUP CALCUL ATION NO. OPTIONAL TASK CODE PAGE 17 14936.o1 .I A c. 14938 ol-C M N/A i t
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46 If 18
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- 2. /
ET IS 29 30 35 38 53 34 35 36 - 37 38 39 40 l el 48
. \ 43 44 43 46
i Calculation No. ' Calculation Title 14938.01-S-15 Containment Suarp Volume 9467-1493801-B1V
I
' CALCULATION TITLE PAGE ***'y ' '"',, 4 A.ND 50MMARY- 'j'fk *s#[,'[."*$**.* 1 c3 l Q ^ C^T 55Ca' M
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- CALC.NO
- REcui#ED I A _
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. PREPARER (S)/DATE LS) R EVIEWER(S)/C ATE (5) ',No gP t gvitw 5 /c ATE (51 C ALC NO- OR R EV. NO. vg3 l No jm A )cPedi Gbecule.Cfrru &%k:C@rres -
f.a,scanio :.c. we.es. c.c.To ates. o 9/A 4 /11181 4 24/57 4/M/83
.. . u n. . ., .. s .. s., . . .
To DeTs.euias TOTAL Volume. GPActTY ol= ESAC. TOR. Coo TA iu ms tor GutA P @. E L ztr_P -it' 9EE bocy og: C ALc,,
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kF PvJGS. t)' Pv]4 inII - F(- l'o A - G 2.) DW A ti7if -FV -M co cousi..:
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llti5 ,', No" '05 w/s- l' N 5 d , J. t m 2~*y2 l viits.'n.h 3vd,r - R 3Ls1% l# l 1
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TABLE OF CONTENTS CALCULATION. NUMBER PAGE REVISION DESCRIPTIO N REMARKS N O. NO. DATE i TITLE FAGE O Q[2fl81 2 TABLE CF CONTENTS p q y g 3- G BoPT #F (Alc.ulAJtOH O 4 tl f f} 6 haosx o F ^TTAditME.MT O. 4 hA I?9 L 1 s i
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14938.ot - S- 15 o 3 I' PetsmeER/SAtt AEVIEWER / CNECRER 2 DATE INDEPENDENT REVIEWER /DATE r' A .S cmd /+-21-r'f
. dCTofEE5 / 6 24 53 CCToEE E5 /4 20.39 x au***cte riE - v E PCO - N oat e AusA umt t ) t 2,/RF4cTart.
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