ML19339A830
| ML19339A830 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 10/31/1980 |
| From: | JERSEY CENTRAL POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML19339A829 | List: |
| References | |
| TASK-03-12, TASK-3-12, TASK-RR NUDOCS 8011050425 | |
| Download: ML19339A830 (500) | |
Text
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ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED ELECTRICAL EQUIPMENT Oyster Creek Nuclear Generating Station Docket No. 50-219 October 31, 1980 Jersey Central Power and Light Company i
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bb@ ' (') Madison Avenue at Punch Bcwl Road Mornstown New Jersey 07960 (201)455-8200 October 28, 1980 Director, Nuclear Reactor Regulation Attn: Mr. Dennis M. Crutchfield, Chief Operating Roactors Branch No. 5 U. S. fluclear Regulatory Commission Washington, 3. C. 20555
Dear "r. Crutchfield:
The NRC l etter dated December 23, 1977 requested information concerning the SEP topic " Environmental Qualification of Safety-Related Equipment." The i n f ormation requested was the identification of the electrical equipment required to perform a safety function under the environmental condition resulting from each Design Basis Event for Oyster Creek Nuclear Generating Station, definition of the limiting service environmental conditions for
(~; operation of the equipment identifiec aDove and tne identification of the k- '
supporting documentation. Tais information was transmitted to you by a letter dated December 19, 1978.
Subsequently, en NRC letter dated February 15, 1980 transmitted NRC Guidelines For Evaluating Environmental Qualification of Class lE Electrical Ecuipment which recuesTed JCP&L to rev iew tne environmental cualification of the safety related equipment both inside anc outside containment.
On March 10 tnrougn 13, 1980 meetings .ere helc witn the NRC staff at Oyster Creek Nuclear Generating Station concerning the subject matter and a list of saf ety relatec systems essential to r itigate certain high energy line breaks was jointiy agreed on. Based on tnis Iist of systens, JCP&L developed a Iist of conconents in these systems.
The component list and an env ironmental profile for each oiece of eculpment
.ere transmittec to you by our letter dated May 7, 1980.
Your letter cated August 29, 1980 transmitted an order whicn requireo JCP&L to submit by Novemoer 1, 1980 all inf orrr.ation to support a saf ety eval uat ion of the safety re l atec electrical equipment exposed to a harsh environment, inside or outsice containment.
The attached report entitled "Env ircnmental Qualification of Safety-Related Electrical Equipment - Oyster Creek Nuclear Generating Station" is being transmitted to you as our resconse to your August 29,1980 order.
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Jersey Ce-tral Power & L ght Company is a Member of the General Pubhc Utat<es System
(G As it is indicated in the Chapter 7 of the report, those equipment which is without an adequate qual i f ication documentation and is considered essential to mitigate a postulated high energy line break will be either replaced or cualified by June, 1982 assuming that procurement aclays by equipment suppliers do not preclude replacement by that date. If you should have any further questions, please call James Knubel ( 201 -455-d 753 ) of my staff.
Very truly yours, f $11 4) '
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Ivan R. Fin rock Jr.
Vice Presi in t Ir STATE OF NEW JERSEY COL 7 TIT OF MORRIS Sworn and subscribed to before me this 3 day of O h , 1980
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ENVIRONMENTAL QUALIFICATION OF
{dl SAFETY-RELATED ELECTRICAL EQUIPVENT TAa,LE OF C01 TENT 5 Chapter 1 Dackground and Surmary Chapter 2 Containment Temperature and Pressur+1 Prof ile for Env ironmental Qual i f ication of Safety Related Equipment Chapter 3 Environmental Effects on Safety Grade Electrical Equipment Due to LOCA and High Energy Pipe Rupture Chaoter 4 Reactor Sullding Flood Level Analysis Chaotor 5 System / Equipment List Chapter 6 System Component Evaluation Work Sheet Chapter 7 J ust i f icat i on for Continued Operation Ca)
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/~3 O CHAPTER 1 BACXGROU.';D AND SUWARY On Decemcer 23 1977 Jersey Central Power & Light Company (JCP&L) received a letter f rom the commission which requested information concerning the SEP top ic "Env ironmental Qual i f ication of Safety-Related Equipment". The information requested by this letter was the identi f .le.ation of the electrical eculpment recuired to perform a safety function under the env ironmental conditions resulting from a design basis accident, definition of environmental service conditions at eculpment locations and the identification of the supporting documentation. In response to this recuest JCP&L submitted a letter dated December 10, 1978 which transmitted a list of safety related electrical eculpment inside containment, environmental conditions inside containment following a loss of coolant accident, and ref erence documentation.
Subsequently, JCP&L received an flRC letter dated February 15, 1960 which transm i tted NRC gui delines for Eval uating Env ironmental Qualification of Class IE Electrical Equipment.
For the inside containment environment, extensive plant specific analyses were performed using conservative assumptions for the determination of the containment transient response to a main steam line break. The resul ts of tne anal yses are given in Chapter 2 of this report.
The rtRC letter dated August 29, 1980 transmitted an order which required JCP&L to submit by November 1, 1980 all information to support a safety ev al uation of the saf ety related electrical equipment exposea to a harsh environment.
Chapter 3 of this report incl udes a list of safety related electrical equipment outside containment, outsice containment environmental profile, plant layouts showing the locations of the equipment and the resul ts of the radiation anal yses ins ide containment. The description of each analysis is also given in this chapter. Further anal ysis of the radiation l ev el s insice containment is still in orogress since the results given in this report do not consicer existing attentuators such as the biological shield wall w i th high density concrete. The results will be transmitted to you as soon as it becomes av ai l ab l e.
Following the meeting in March with the NRC staf f , an extensive search for cualification doc um en t a t i on was conducted by JCP&L and GPU personncl.
Numerous eculpment suppliers, NSSS vendors and other utilities were contactec to accomplish this task. Although we were able to obtain qualification cocumentation for some equipment, we have not been able to find the cocumentat ion f or many ecuipment. The main reason f or th is d i f f icu l ty is due to the fact tnat most of these equipment were instal led more than ten yeers ago.
The resul ts of this search are ref lectec on the Systam Componen? Eval uation _ acrk Sheet given in Chapter 6 of th is report.
JCP&L wiii either reoIace or qualify by June 1982 those components hq witnout adequate cual i f ication documentation that are considered essential to mitigate a postulated high energy line br eak assuming that procurement celays by
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(m) equipment suppliers do not preclude replacement by that date. The equipment I isted on Table 1 of Chapter 3 without an asterisk (*) are considered essential to mitigate a high energy line break (HEL3) and therefore, are subject to replacement or quoi!*1 cation if it lacks adequate qualifica* ion documentation.
To this end, several meetings have already been held with various equipment supoliers to discuss the replacement task. Since Oyster C eek Nuclear Generating Station will continue to utilize the "nonasterisked" equipment until they are replaced or quallfled, we have prov ided a justi f ication f or continued plant operation in Chapter 7. The asterisked items listed on Table 1 of Chapter 3, on the other hand, are those components not required to mitigate a HELB as explained in Chapter 7. Even if those asterisked equipment were to fail after a HEL3, the protection of the reactor is adequately prov ided by other systems and the "non-asterisked" equipment. Tnerefore tne thermal aging and radiation susceptibility characteristics of the component materials were eval uated. This evaluation revealed inat certain equipment included thermal agi.ng and radiation sensi sti, u materials. JCP&L will replace these component mater i al s w i th qualified ones by June 1982.
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OYSTER CREEK CONTAINMENT TEMPERATURE PROFILE j FOR ENVIRONMENTAL QUALIFICATION i OF EQUIPMENT
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TABLE OF ' CONTENTS Page 1.0 Introduction and Background 1-1 2.0 . Study Objectives 2-1 3.0 . Analysis Approach- 3-1 3.1 Background 3-1 3.2 Specific Discussion 3-2 4.0 Computer Codes and Models 4-1 4.1 Primary System Analysis 4-1 i .
i 4.1.1 Break Model 4-1 4.1.2 Other Changes 4-2 l,
i 4.2 Containment 1 Analysis 4-2 5.0 Oyster Creek Main Steamline (MSL) 5-1 Break Analyses I
() 5.1 Pressure. Vessel 5-1 5.1. l' Scram Assumptions 5-2
- 5.2 Break Mass Flow Rate and Enthalpy 5-4 5.2.1 Background 5-4 5.2.2 Break Size Choices 5-5 ,
. 6.0 Containment Response to MSL Breaks 6-1 6.1 Model Information 6-1 6.1.1 Mass / Energy Input Data 6-1 6.1.2 Other Input Data 6-2 6.1.3 Evaporation - Condensation 6-2 6.1.4 Heat Sink Data 6-3 J
6.1.5 Containment Spray 6-4
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Page 6.2 Containment Analyses 6-8 6.2.1 Introduction 6-8 4
6.2.2 Containment Response - No Heat Sinks 6-9
- or Sprays 6.2.2.1 Analysis Results 6-9
- 6.2.2.2 Extent of Superheat 6-11 6.2.3 Containment Response with Heat Sinks - 6-21 J No Sprays 4 6.2.3.1 Heat Sink Heat Transfer 6-21 6.2.3.2 Heat Sink Condensate Treatment 6 6.2.3.3- Analysis Results with Heat Sinks 6-26 6.2.4 Containment Response with Heat Sinks and 6-34
{} Containmt.it Spray 7-1
) 7.0 Oyster Creek Temperature and Pressure 1 Prcfiles 8.0 References 8-1 1 9 n
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! '4-1 Model Nodalization Diagram 2
5-1 2.0 ft Feedwater & Break Flowrates 1
2 I 5-2 2.0 ft Break ~Enthalpy & Quality 2
1 5-3 0.75 ft Feedwater & Break Flowrates '
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i 5-4 'O.75 ft Break Enthalpy.& Quality 2
5-5 0.01 ft -Feedwater & Break Flowrates 2
5-6 0.01.ft Break Enthalpy & Quality i
! 6-1 Containment Spray Flow Diagram 2
6-2 2.0 ft Drywell Liquid-& Vapor' Temperature without Heat Sinks 6-3 2.0 ft2 Torus & Drywell Pressure without-Heat Sinks !
2 6-4 0.75 ft Drywell' Liquid & Vapor Temperature *
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2 6-5 0.75 ft Torus & Drywell Pressure without
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!~ 6-6 0.01.ft Drywell Liquid & Vapor Temperature v.ithout Heat' Sinks 2
j 6-7 0.01'ft Trous & Drywell Pressure without Heat Sinks 4
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6-11 2.0 ft' T^rus & Drywell Pressure with Heat Sinks 6-12 0. ' " ' f t 2 Torus.& Drywell Pressure with-Heat Sinks i
6-13 'O.01 ft2 Torus & Drywell Pressure with Heat Sinks 2
6-14 - 0.75 ft Drywell Vapor Temperature with Heat Sinks and Containment Sprays (0-800 sec) 6-15 0.75'ft 2 Drywell Liquid & Vapor Temperature with Heat Sinks and Containment Sprays (0-3200 sec.) .
- 6-16 0.75 ft Drywell Pressure with Heat Sinks and 4 Containment Sprays (0-800 sec.)
6-17 0.75 ft2 Torus and Drywell Pressure with Heat Sinks and containment sprays 7-1 Oyster Creek Drywell Temperature Profile (11 1
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', 6-2 Oyster Creek Passive Heat Sink Data Summary of. Containment Response I
j 6-3 Results (No Heat Sinks / Sprays) ,
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1.0 INTRODUCTION
AND' BACKGROUND The NRC " Guidelines for. Evaluating Environmental Qualifications of Class IE Electrical Equipment in Operating Reactors" indicates that Boiling Water Reactors are to use 340'F for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> as the basis for judging whether individual component qualifica-tions meet the guidelines. At a meeting with the NRC on February 21, 1980, some BWR licensees indicated that they might want to use plant specific analyses to justify less severe environmental conditions and thus to take exception to the guidelines. In this way, the technical issue of whether or not a plant specific analysis justifies less severe conditions than 340*F for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> has been raised.
GPU has undertaken the task of providing a plant
( specific containment temperature profile for the environmental qualification of equipment at Oyster Creek. This report provides a description of the analyses which were performed in fulfillment of this task. The analytical methodology which was used, the analysis approach, the plant specific input data and the analysis results are discussed herein.
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(') STUDY OBJECTIVES There are three major objectives associated with this study. These are as follows:
- 1. ' Review the applicability to Oyster Creek of the generic envelope for containment temperature and pressure provided in Appendix C of NUREG-05882,
- 2. Provide a plant specific analysis of the Oyster Creek time-dependent containment temperature and pressure profile for consideration in evaluating
, the environmental qualification of equipment inside containment.
- 3. Provide the "information necessary for Staff review of plant specific containment analyses" which was requested as Enclosure 3 to the March 28, 1980 letter from Dennis L. Ziemann to I. R.
Finfrock, Jr.
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(} 3.0 ANALYSIS APPROACH 3.1 Background In order to determine the containment temperature and pressure profile to be used for the environmental qualification of equipment, it is necessary to eval-uate two separate effects. The first is the primary system response to a loss-of-coolane condition. This is important in order to, determine the time-dependent mass and energy input to the containment structure.
The second consideration is then the containment response to the mass and energy input. This results in a time-dependent temperature and pressure pro-file. The severity of the containment response
, depends in large part on the magnit ade and nature of the mass and energy input. For example, if the pipe g break is below the core then a two phase mixture of Cd steam and water will enter containment. If the-break is above the vessel two phase mixture level then single phase steam will enter containment. Further, in addition to the phase characterization of the blowdown fluid, the rate of blowdown is also an im-portant consideration. Large breaks result in a t
rapid blowdown of large amounts of mass and energy but for only a short period of time. Small breaks produce the opposite effect. In addition, the nature of the blowdown can be. altered by various opera'or actions and by primary system automatic responses.
For the purposes of this study, it was desired to find a primary system break size, location and other l conditions which would maximize the time-dependent temperature and pressure profile of the containment l l
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tions are appropriate for Oyster Creek. If not, then the appropriate containment temperature and pressure profile for Oyster Creek could be determined. This.
. would meet the primary objectives of the study.
3.2 Specific Discussion Oyster Creek docket analysis of the DBA LOCA have shown that peak drywell temperatures do not exceed 285'F. This is for a double-ended rupture of a recirculation line with containment spray. This is confirmed by more recent analyses as well.4 . The blowdown is a two phase mixture of steam and water.
The peak ' containment pressure is about 38 psig which is the saturation pressure at the peak containment temperature. Smaller recirculation line breaks have gx also been previously evaluated using the CONTEMPT-LT/26 ccde. The peak temperatures for the smaller breaks are less than those of the large break. The profiles of temperature and pressure for small breaks are relatively flat at the elevated conditions for- a long period of time if containment sprays and the effects of containment heat sinks are ignored.
Otherwise, temperatures and pressures decrease substantially. This is also true of the DBA LOCA analysis as well.3 4 8 The Oyster Creek FCSAR shows the containment conditions following the DBA LOCA without containment sprays but with .the effect of containment heat sinks.5 In all of the above analyses, both small and large breaks, the contain-ment temperatures and pressures are substantially less severe than those provided in NUREG-0 588.
Temperatures as high as 340*F require that the s
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containment atmosphere be filled with superheated
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steam sinc ( the saturation pressure at a temperature of 340*F is 118 psia, which is far greater than the drywell design pressure. The maximum superheating in a containment atmosphere is obtained when the blow-doen is pure steam since the increased heat capacity of a droplet laden atmosphere reduces superheat con-ditions.
It can be concluded that the DBA LOCA provides the highest containment temperatures and pressures for breaks which occur below the core mixture level. The most severe long term t2sponse for all break sizes below the core mixture level results when containment sprays and containment heat sinks are ignored. When this assumption is made, temperatures and pressure inside containment are relatively flat at about 260*F
() and 20 psia, af ter an initial peak, for all breaks.3,4,5 On this basis, it has been decided that the analysis approach will be to consider only steam breaks above the core mixture level in deter-mining the containment temperature and pressure pro-file for environmental qualification of equipment.
4 A spectrum of breaks in the main steam line will be analyzed and their respective containment response will be evaluated.
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() 4.0 COMPUTER CODES AND MODELS 4.1 Pr imary' System Analysis The' basic Oyster Creek RETRAN model with feedwater and pressure regulator controls was used. The Oyster Creek RETRAN model was derived from the Exxon Nuclear large break WREM model described in XN-75-55.
The basic RELAP plant geometry was converted to l RETRAN format and control system models for feedwater !
and pressure regulation were incorporated.
The GPU version of RETRAN RETOl/ MOD 001 was used for these analyses.
Modifications were made to remove the hot assembly noding and to include additional changes for the main steamline break.
() Other changes made were'to incorporate a trip which allows MSIV closure on time and another unich permits a scram on time as well. All other trips were re-tained. A reduction in feedwater enthalpy following MSIV closure was not included. The revised noding diagram is.shown in Figure 4-1.
4.1.1 Break Model The break was assumed to be in one of the main steam-lines -(volume 32 in Figure 4-1) upstream of the MSIV in that line. The non-isolatable break was incor-porated as a junction (J50 ) with a valve. The valve-
! area represents the break area and can be adjusted accordingly. .The MSL' break valve is opened on time
( .01 seconds) using a RETRAN trip incorporated for
- that purpose. The break (valve junction) elevation 4-1
was arbitrarily set at an elevation _which established it at one half of the volume 32 height. This corres-ponds to a physical plant location of approximately the 60 feet elevation.
The break discharge coefficient is set equal to unity. The critical flow models used are extended Henry-Fauske for subcooled flow and Moody for satu-rated flow conditions at the break junction.
4.1.2 Other Changes In order to be conservative with respect to the con-tainment response, it was necessary to maximize the steam which would exit from the bre'ak. For large breaks, the system level would swell during the de-pressurization and eventually result in the discharge of two phase fluid out of the break junction. In
( order to minimize liquid swell, a very large phase separation velocity was used in the upper downcomer bubble model. This maximizes the transport of pure steam out of the steam separators and into the steam ,
dome and steam lines. This, of course, is an ef fect which minimizes the level swell which would occur for large and intermediate size main steamline breaks during which the break quality would become two phase.
Additional model details will be addressed as re-guired during the analysis discussions in forthcoming sections herein.
4.2 containment Analysis The containment response to the main steamline break was determined using the GPU version of the O
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CONTEMPTLT/26 code developed by Energy, Inccr-(}
porated. This is a computer code for predicting the time dependent pressure and temperature response to r.. mass and energy input into the containment.
The standard Oyster Creek model is utilized with unity assumed for the compartment inter-region heat and mass transfer multipliers and uniform vent flow fractions. The initial wetwell and drywell tempera-tures are 120*F and 135*F, respectively. Initial pcessures of 15.5 psia were assumed.
The spray was modeled with an efficiency of 1.0 with a U-tube heat exchanger and a service water and spray flow rate of 4400 gpm. The overall heat transfer coefficient used was 287. This is conservative with respect to the worst case expected for the actual s performance of only one loop of the Oyster Creek containment spray heat exchangers. One loop consists of two spray, two service water pumps and two heat exchangers. The redundant loop was conservatively ignored.
The analysis also includes containment heat sinks for the drywell walls and floor, the torus shell and the biological shield. These sinks were used in a sen-sitivity study to determine their effect on contain-ment pressure and temperature. The details of this and other parts of the analysis will be discussed in Section 6.
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(]) -OYSTER CREEK MAIN STEAMLINE (MSL) BREAK ANALYSES
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Main Steamline (MSL) breaks ranging in size from 3.14 2 2 ft (area of steamline) to 0.01 ft were analyzed. This. range of breaks was needed in order i
to determine the sensitivity of the containment
- response to break size.
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the break mass flow rate and the enthalpy as a func-i t
. tion of time for each break to be considered. This is provided in section 5.2.
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In each of these analyses, the feedwater flow was maintained-in the automatic control mode. The reason 1
for this is that a failure of the feedwater would-be j non-conservative with respect to the objectives of l the study. A steamline break without feedwater would j
result in both high drywell pressure and double low vessel water level conditions.
1 The exact timing would depend on the break size, but i
a these two conditions would' result in the automatic initiation of the containment spray system. The
- initiation of the spray system would rapidly reduce the containment atmosphere temperature and pressure and thus minimize the containment response. Main-taining feedwater flow for breaks which have an in-tegrated steam break flow less than the 2222 lb-/sec feedwater system flow runout' capability results in the prevention of automatic containment spray actua-
, tion since -double low level is not reached in the 5-1 i
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{;' vessel. Both double low level and high drywell pres-sure are required for automatic spray initiation as indicated earlier.
As indicated earlier, a large phase separation velo--
city is used in the upper downcomer region to mini-mize level swell effects and maximize the break en-thalpy. This assumption also results in an increase in the minimum break size which would result in double low level initiation. The reason for this is that the break mass flow rate under high quality, large enthalpy conditions is much smaller than the break flow rate under two phase conditions. Thus, it requires a larger break size under steam break' flow conditions to remove the same integrated mass as a smaller break with lower quality break flow. This point will be referred to in later discussions here-in. For the present, it should be noted that this modeling technique is conservative in terms of maxi-mizing containment energy input from both a break size and break enthalpy s tandpoint.
5.1.1 Scram Assumptions Various f actors will result in a plant scram follow-ing a MSL break inside containment. First of all, a high drywell pressure will result from the break mass and energy input to containment. A scram will occur on 2 psig high drywell pressure. This will range from less than five seconds for a 3.14 ft2 break to about 80 seconds for a 0.01 f t2 break. In addi-tion,-there is an MSIV closure on high flow in each of the main steamlines. This is set atll20% of rated flow. For breaks on the order of 0.75 f t2 and greater, this condition is met very rapidly (less
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(} than 5 seconds) and a scram on 10% closure of the-MSIV would occur. Another means of getting an auto-matic MSIV closure and subsequent scram is on low system pressure (<825 psig) . The final method of obtaininy c scram signal following an MSL break is low level in the upper downcomer. With feedwater assumed to be available, this requires break sizes of about 0.5 ft 2 or greater and would occur in about 15 seconds or less. It should be remembered that the assumption herein of a large phase separation in the upper downcomer results in larger break sizes being 1 required to reach the low level scram setpoint. This is also true of the other scram setpoints as well.
In all cases, a scram signal would be produced. For breaks less than 0.2 ft2, the only signal would be high drywell pressure since the availability of feed-s water prevents low level and pressure from occurring and the break is not large enough to produce 120% of rated MSL flow.
The assumption of a scram occurring later than these time values is conservative for most breaks because this tends to aaxisize the break flow and enthalpy.
Following scram, for those breaks which are large enough to remove decay heat and for the non-isolated small breaks as well, the pressure drops rapidly.
This either reduces the break flow for single phase conditions or results in two phase break flow which has a much lower enthalpy. In either case, the con-tainment response is improved. For the small breaks which are not large enough to remove decay heat, the system pressurizes following an isolation. This maximizes break flow and results in a conservative containment response.
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' Break Mass Flow Rate and Enthalpy
{} 5.2
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5.2.1 Background The input data required for coupling the pressure vessel behavior to the containment' response analysis are the break mass' flow rate and enthalpy as a func-2 tion o f time. Break sizes of approximately 1.0 f t or greater result in reaching double low vessel level if no other scram occurs, even if feedwater is avail-able. If an earlier scram occurs however, double low
! level may not occur because the resulting depressuri-zation and isolation reduces total system steam losses. In most cases, level will increase dramati-cally af ter the scram since feedwater flow imme-diately following scram is much higher than steam flow. For large breaks such as the 2.0 ft case or s greater, double low level will-be reached under all circumstances. Thus, for containment analyses,-this break size' can be considered an upper bound since, for larger breaks, auto containment spray initiation
, will result within the first few minutes and rapidly reduce the containment temperature and pressure.-
The existence of two phase. flow conditions at the break junction is also an important consideration. A two phase blowdown results in a reduction in the containment atmosphere temperature and pressure con-dition. This results from the introduction of water vapor into the gaseous region which reduces the steam superheat.
The l.0 ft 2 break.results in two phase break flow-conditions following scram while the 0.5 f t2 break remains. single phase throughout. It can thus be O
5-4
concluded that the largest break which remaina single
{l phase lies between these two break sizes. An analy-2 sis of the 0.75 f t break shows that it remains single phase as well. This implies that the break of 2
interest is between 0.75 and 1.0 ft . This of course is based on the conservative use of a large phase separation in the upper downcomer which tends to increase the size of the minimum break which does not. result in two phase break conditions. j
. 5.2.2 Break Size Choices On the basis of the above discussion, it would be instructive to examine the containment respense to che 2.0 ft 2 break, which is theLupper bound case.
2 The 0.75 ft break represents an intermediate size break which is approximately the largest break with a 73 pure steam blowdown and should result in a severe
\-) containment response. Thus, it should be considered in the containment analyses. The smallest break analyzed is the 0.01 ft 2 case. This has been chosen to be the lower bound of the spectrum of breaks analyzed. The break mass flow rate and enthalpy data for these breaks is provided in Figures 5-1 through 5-6.
Now that the three breaks of interest have been iden-tified, and a rationale.for their choice developed, the containment response to the mass ard energy input from the MSL breaks can be investigated.
5-5 o
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i 6.0- CONTAINMENT RESPONSE'TO MSL BREAKS
' 6 .1 Model Informatio'n l
6.1.1 Mass / Energy Input Data The input data for each of the three break sizes analyzed is provided in Table 6-1. For each of the '
cases-analyzed, it is assumed that break flow will-eventually degrade to become decay heat steam flow.
No other means of decay heat removal is assumed. In all cases the time chosen for decay heat steam flow out of the break is conservative. Further, Oyster Creek has redundant, safety grade isolation con-densers which would rapidly reduce the pressure and steamflow out of the break. This system has not
, been accounted for in this analysis. In addition, as stated in section 4.1, a reduction in feedwater s ( enthalpy following MSIV closure was not included in
[ the RETRAN Oyster Creek model used for this analy-
) sis. Isolation normally results in a loss of feed-water heating. Thus, the feedwater is entering at a
.t
. heated enthalpy of 280 BTU /lbm. In reality, this would be unheated feedwater at much lower enthalpy.
, The saturation pressure at this elevated eathalpy is 80 psia. The system cannot depressurize be.ow this value since the incoming feedwater will flasa to steam and maintain the pressure at this .value. This of course will also result-in an artificial break flow. This is an artificial effect as a result of a model deficiency, but it is conservative since it increases the system energy which enters the con-tainment.
6-l'
- _ _ . . _ . . _. - _ . . . _ . . . . . . _ . _ . _ _ . ~ . _ ..
e 6.1.2 Other Input Data Other pertinent containment analysis input para-meters are as follows:
DRYWELL/WETWELL INPUT DATA INPUT ITEM DRYWELL WETWELL Net Free Volume (FT 3) 180,000 213,300 Initial Pool Water Volume (FT 3) 0.0 87,30d Initial Temp of ATM ('F) 135. 120.
Initial Temp of Pool (*F) 135. 120.
Initial Pressure (PSIA) 15.5 15.5
.' Relative Humidity (0 to 1) 1.0 1.0 Pool Surface Area (FT 2) 1373.9 9219.
'] HTC Multiplier 0.0 1.0 Mass Transfer Multiplier 0.0 1.0 O
%)
6.1.3 Evaooration - Condensation The CONTEMPT LT/26 code accounts for heat transfer from a compartment pool surface. Heat transfer in the form of sensibie heat transferred by a tempera-ture gradient is accounted for as well as the latent heat of mass transferred by a concentration gradient in the vapor.
When the mole fraction of vapor in the bulk atmos-phere space exceeds the mole fraction of vapor at the pool-atmosphere boundary, then condensation occurs. If the opposite if true then evaporation occurs.
(~S
%.)
6-2
1 l
l
()
l This can~be seen from the following equation 6.1:
Q=C y Hb - (T y -Tb) + [C 2 K b M (i ev + I )e (:X y - X b )]/X am (6.
Q = surface heat flux Hb = Sensible heat transfer. coefficient at interface Cy = input heat transfer multiplier C2 = input mass transfer multiplier kb = mass transfer coefficient M = molecular weight of water e = specific internal enthalpy of fluid l
transferred i
gy = latent heat of vaporization X y, Xb = m le fraction of vapor in bulk and at boundary X
am = 1 garithmic mean mole fraction of air.
This model was utilized in the suppression pool compartment but was not utilized in the drywell.
The reason for this is that the model tends to cause ,
an approach to saturation conditions in the compart-ment atmosphere. The purpose of this analysis is to ce vevatively estimate the drywell vapor tempera-tuct . within reason. Since the highest temperatures are obtained for a pure steam atmosphere under superheat conditions, the use of the evaporation-condensation model would be nonconservative since it would increase the vapor mass in the atmosphere and reduce the extent of superheat.
6.1.4 Heat Sink Data Analyses were performed with and without heat sinks in containment in order to determine the relative O
6-3
t 1
1
() change in the drywell vapor temperature profile which can be attributed to heat sink effects. The I details of the heat sink treatment are presented in a later section. The Oyster Creek heat sinks utilized in the analyses were limited to the walls and floor of the drywell and the walls of the biolo-gical shield. This provides a significant amount ~ of surface area for heat transfer to the structures and
- for condensation of steam on the surfaces of the heat sinks. The pertinent passive heat sink infor-mation is ta! ulated in- Table 6-2. All additional o
drywell equipment and piping whi.ch would be avail-
- able for sensible heat transfer and steam conden-sation has been conservatively ignored. The drywell fan cooling system has also been ignored.
6.1.5 Containment spray O The model for containment spray and containment spray cooling is shown in Figure 6-1. Water is taken from the suppression pool, passed through two heat exchangers with an area of 6200 ft2 each and a pre-specified overall heat transfer coefficient.
In CONTEMPT, this is modeled as one heat exchanger with a surface area of 12,400 ft 2 . 95 percent of the cooled water is sprayed in the drywell atmos-phere and 5% in the torus atmosphere. The spray water is collected in the drywell pool and returned to the torus by gravity. This is a closed loop system. The sprays are assumed to be 100% efficient in the analysis. Thus, the spray droplets absorb energy until their specific enthalpy equals that of the vapor region, i
l CE) 6-4 4
4
. - - , ,_w -
+ , , - , , , . - .
As mentioned above, drywell spray collects in the
.( }
drywell pool and drains back to che wetwell pool via
~
-the vents. CONTEMPT does not simulate this behavior via an internal model. Spray water merely continues
.to fill the drywell. In order to overcome.this shortcoming, an alternate spray system was used to 4
obtain water from the drywell pool at a rate equal ~
to the drywell spray flow rate and spray it all back to the wetwell witho'ut passing it through a heat exchanger and with a zero spray efficiency. This serves to merely transfer liquid from the drywell
- pool to the wetwell pool without energy exchange in the process. This simulates the vent draining pro-cess and overcomes'this deficiency in the CONTEMPT
, code.
Additional details regarding the Oyster Creek con-gg tainment spray heat exchangers and the appropriate
\l CONTEMPT modeling of these systems have been docu-I mented,4 and will not be repeated here.
- The following parameters are used for the Oyster Creek containment spray heat exchanger modeled in this analysis:
Heat Exchanger Type: U-tube Overall HTC : 287 BTU /lbm*F hr Spray Flow Rate : 4400 gpm Heat Exchanger Area: 12,400 ft 2 Service Water Flow : 4400 gpm Service Water Temp : 85*F The heat removal rate and spray flow temperature for I l
a given suppression pool water temperature which
. \
6-5 i
l l
results from these heat exchanger model assumptions is conservative' with respect to the actual single loop operation of the Oyster Creek containment spray system under worst case f ailure and performance assumptions as discussed in section 4.2. The effect of the contairement spray system on the containment response following an MSL break will be discussed in a later sec-ion.
l
- O 1
6-6
l
.1 l
l TABLE 6-1 l O Containment Analysis Input Data Breaksize Time Mass Input Enthalpy (ft ) (Sec.) (LBM/Sec.) (BTU /LBM)'
2.0 0 0 0 0.5 3375 1173 8.5 2614 1139 11.5 5802 663 42 3086 648 80 724 1181 124 360 1175 Decay Heat at 400 Seconds (1.2 ANS) 400 43 1200 600 38.6 1200 1000 33.8 1200 2000 28.7 1200
-4000 23.4 1200 8000 19.2 1200
. 0.75 0 0 0 12 1379 1194
, 80 637 1201 160 354 1198 l 240 264 1196 Decay Heat at 600 Seconds (1. 2 ANS) 600 38.6 1200 1000 33.8 1200 2000 28.7 1200 4000 23.4 1200 8000 19.2 1200 0.01 0 0 0 0.5 11.39 1192 6.0 22.54 1196 700 21.11* 1197 3600 '
21.11* 1197
. 8000 19.25 1197
- Break energy at 700 seconds = decay heat'at 1 hr. <
Assume constant 'till 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, then 1.2 ANS l 6-7 O l 1
1
- - - - . - . .--- - r ,4 c ,,m , -- , , . , , , .
eN
() 6.2 containment Analyses 6.2.1 Introduction i Oyster Creek LOCA containment analyses for peak pressure and temperature determination have histori-cally.been analyzed without heat sinks or contain-ment sprays. The reason for this is that the LOCA considered has always been the DBA LOCA. This large recirculation line break results in reaching peak temperature and pressure conditions within 10 seconds. This is too rapid for heat sinks or sprays to be of any benefit.
When heat sinks and/or sprays are used for the DBA LOCA, their effect on the peak conditions is negligible. A benefit is seen in the long term
-however.3,4,5
(}
The approach which has been decided upon for the work herein is to perform a series of parametric analyses. Initially, the containment response will be evaluated without containment heat sinks or sprays. This.will be followed by a parametric analysis which will include heat sinks, but not sprays. Finally, the effect of both heat sinks and sprays will be evaluated. It is the nature of the MSL breaks considered.in this study to be long term in their containment response since the blowdown for j most cases is pure steam, eventually becoming only the steam from decay heat. On this basis it'is expected that benefit will be derived from the use of containment heat sinks, especially if the environment is superheated. Since the containment ,
will ' ave a steam environment, there will be an
(~s) s obvio1s benefit when sprays are used. Since sprays 6-8
i
/~
are not initiated automatically as a result of th.e
(-)/-
assumption of continued feedwater, operator action will be required for spray initiation.
6.2.2 containment Response'- No Heat Sinks or Sorays 6.2.2.1 Analysis Results The containment response to the 2.0, 0.75 and 0.01.
2 ft breat 3 without heat sinks or containment sprays was evaluated. The.results are provided in Figures 6-2 through 6-7. The mass and energy input data was provided in Table 6-1. As discussed in section 6.1.1, th s input was chosen to be conserva-tive. A comparison of Figures 6-2, 6-4 and 6-6 shows that the most severe containment temperature response from a peak temperature standpoint occurs for the 0.01 ft 2 break. The temperature and pres-sure results for these cases are summarized in Table 6-3. The cor.tainment response to the 2.0 ft2 break is tempered to a large extent by the presence of two phase fluid during the first 80 seconds as seen in Table 6-1. This greatly 51nders the tem-perature rise during the first 1000 seconds by introducing water vapor into the atmosphere and reducing the extent of superheat. As the blowdown returns to pure steam, the temperature increases. I It requires a long time however, to build up a con-siderable superheat and the.overall response is less severe than either the 0.75 or 0.01 ft2 cases.
The pressure response characteristtes for each of the three cases are quite similar. An initial rapid increase in pressure results and is then followed by a monotonically increasing pressure. Tne rate of
()
6-9
1 I
i 1
i
( the' initial pressurization is break size dependent as can be seen in Figures 6-3, 6-5 and 6-7 and Table 6-3, but there is only a 2.0 psi difference at 8000 seconds between the 2.0 and 0.01 ft breaks.
The m.ny oscillations which are observed in the drywell vapor temperature and pressure response are the result of repeated vent clearing and reset which occurs very frequently throughout the entire problem duration. The drywell conditions are very sensitive to this vent clearing process as can be seen in the plots. If the pressure differential between the drywell and wetwell vapor regions is less than the hydrostatic pressure associated with the vent submergence, then the vents are closed. The vent clears if the compartment pressure differential exceeds the submergence hydrostatic pressure. The
(} vents are continually relieving the pressure in the drywell by providing a flow path to the suppression pool. The pressures of the suppression chamber and drywell will thus rise together in this oscillatory manner.
2 A review of the 0.01 ft temperature profile reveals that the NUREG-0588 peak temperature of 340*F is quite reasonable when heat sinks and sprays are ignored. The dual peak is not seen, however, and neither is the very rapid rise time. The 0.75 2
ft break does show a mild dual peak behavior, but the magnitude and timing are not in accord with the NUREG-0 588 profile. It can thus be concluded that, i if containment heat sinks and sprays are not con- !
sidered, the peak temperature of 340'F is appro-priate. It is also conservativel appropriate to 6-10 l
i
1
(~3 maintain this high teniperature for a period of time N) until the operator acts to initiate sprays. The exact temperatures for the .01 ft 2 break are pro-vided in Figure 6-6. .The time required to reach the peak temperature for Oyster Creek will be greater than that provided in .NUREG-0 588. This is true of all breaks considered.
The next section will evaluate the containment re-sponse with consideration given to containment heat sinks in order to determine the relative change in the drywell vapor temperature profile.
6.2.2.2 Extent of Sucarheat A review of Table 6-3 shows that the peak tempera-tures are approximately 330 F for all cases, with corresponding pressures of approximately 38 raia sJ (somewhat lower for the 0.01 f t 2 case). The
- saturation pressure for 330*F is about 103 psia.
Since the actual pressure is substantially less than this, a significant superheat exists. The satura-tion temperature at 38 psia is about 264'F. Thus, the extent of superheating is the difference bets een this value and the calculated temperature, or 66*F (somewat higher for the 0.01 f t 2 case). A review of the pressures and temperatures for each case in Figures 6-2 through 6-7 reveals that superheat conditions exist throughout the entire analysis interval for each of the cases analyzed when containment heat sinks 'and sprays are ignored. '
6-11 i
)
l P
i
' TABLE 6-2 ,
l 1
h- ~ OYSTER CREEK PASSIVE HEAT SINK DATA Area Thickness 2 Material Inch ft Concrete 48 886. Cyl.
- 1. Biological Shield - Lower.
Concrete 60. 370. Cyl.
- 2. Biological Shield - Middle Biological Sheile '- Upper Steel 0.25 2088. Cyl.
3.
Concrete 29.4 Steel 0.31 Drywell - Floor Concrete 11.25 1374. Slab 4.
Drywell Sphere - Lower Steel 1.154 T 12. Spher.
5.
Insulation 2.5 Concrete 78.
Drywell Sphere - Middle Steel 0.770 4220. Spher.
6.
Insulation 2.75 Concrete 78.
stee' o 722 3,85. Spher.
O 7- orrwe11 sphere - upper Insulation 2.75 Concrete 78.
- 8. Drywell Transition Steel 2.56 1433. Spher.
Insulation 2.5 Concrete 78.
- 9. Drywell Cylinder Steel 0.640 2574. Cyl.
Insulation 2.5 Concrete 78.
~
- 10. Drywell Head Steel 1.188_ 428. Spher.
Steel 0.385 14952. Cy1.
- 11. Torus MATERIAL PROPERTIES ,
Thermal Conductivity Volumetric Heat Capacity (BTU /hr-ft2 op)
(BTU /FT3-o p )
1 Material 22.62 1
Concrete 0.92
- 27. 58.8-Steel FIRE-BAR 0.02 3.74 (Astestos fiber -
magnesitecement) 6-12 l
OYSTER CREEK MSLB CONTAINMENT ANALYSIS CONTAINMENT SPRAY FLOW DRT.lELL FLCW FRACTICII
= 0.95_ l DRYWELL SPRAYS
,rYl ih Yn !
jl Jk y a TORUS 4 6 d A i COOLING FLOW hy VATER FLCW
/ "t /l
/'1 FRACTICN
- 4+00 c * ~
- 85 i 4
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a g i CCNTADIMENT SP RA'l HX ~
U = 287 A = 6 200 FT,2 '
HX O- CCNTAOLCIT
. REZ 57 RAY FLOW "ORUS
- 0?
- h A
//n u f
DR M E SP m TORUS
- BREAK FLCW n RETUMIS TO TORUS ffe.=
-7 Q _
y J '
6-13
"h I TABLE 6-3 Summary Results Containment Response - No Heat Sinks / Sprays-Temp Character Break Peak - at 8000 of Peak Size Temp Time Sec Temp Pressure Time (ft2) (*F) (sec) (*F) ,_
Curve psia -(Sec) 2.0 328 7994 328 1. Flat After 2000 sec' 38* 8000
- 2. 300*F by 600-sec
- 3. 320*F by 1200 sec 0.75 328- 6740 326 1. 300*F by 250 sec 37.5** 8000
- 2. 320*F by.1500 sec
- 3. Flat after.2000 sec
- 0.01 333 1081 319 1. Flat a f ter peak 36*** 8000
- 2. 310*F by 400 sec
- 35 psia at 40 sec, then relatively flat
- 35 psia at 150 sec, then relatively flat
- 30 psia at 800 sec, then linear to peak i
3 6-l' l
__ . . _ . _ _ _. _ .~ .. . . _ _ _ . . , . ,
l SI-9 O\
ORYVELL VAP. TEMP.F 336 i
211 23S 261 286 311 l i i i i l i i DRYWELL LIO. TEMP.F 175 215 255 295 335 p5 ,
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) 6.2.3 Containment Response With Heat Sinks - No Sprays 6.2.3.1 Heat Sink Heat Transfer The geometric heat sink data was introduced in sec-tion 6.1.4. This section presented the general information regarding the conductor geometry, mesh spacing, material identification, and material ther-mal properties used in the CONTEMPT analysis. An-other important consideration in the treatment of heat sinks in this analysis is the manner in which heat transfer from the containment enviror. ment to the heat structures is being calculated.
A discussion of the methodology which should be used for heat s;nk heat transfer in the determination of the containment environmental response to MSL break jg is presented in Appendix B of NUREG 0588.2 It is stated that the Uchida heat transfer correlation should be used for MSL break accidents while in the condensing mode. A natural convection heat transder coefficient should be used at all other times when not in the condensing heat transfer mode. The Uchida correlation is applied as a condensing heat transfer coefficient in the Newton rate equation with the temperature difference defined by the steam saturation temperature at the containment pressure and wall surface temperature of the heat sink.
Convective heat transfer is applied with the same equation and the use of a natural convection coeffi-cient. The temperature difference for natural con-vection heat transfer should be the bulk compartment vapor temperature and the wall surface temperature of the heat sink.
(7
/
xj l
6-21
,_. . .- -________ -_ - =______ __- - __ __-
P
~- The CONTEMPT code Uchida heat transfer option has been utilized for all heat sinks in this analysis.
The heat transfer rate for this option is calculated from:
q,,
g_ = h u (Ts - Tw) (6.2) or
. =2 (Ty - Tw) (6.3)
Whichever is greater.
f"=Surfaceheatflux h
u = Uchida condensing heat transfer coefficient Ts = Satur: :_on temperature based upon compartment
()' pressure Ty = Bulk temperature of compartment atmosphere T y = Heat sink wall temperature A = Surface area associated with heat transfer
. 2 = Lower limit heat transfer coefficient For saturated steam conditions, equation J.2 would be used. For superheat conditions, it would be expected that a switch to equation 6.3 would occur )
as the heat sink wall temperature approached the saturatiori temperature at the containment pressure.
The use of equation 6.2, which is. the Newton rate equation with a Uchida condensing heat transfer coefficient is in complete accordance with the NUREG-0588 quidelines. The switch to equation 6.3 when its surf ace heat flux exceeds equation 6.2 is
{} nonconservative in the sense that continued use of l 6-22
{} equation 6.2 would result in a1 smaller heat flux.
.However, equation 6.3 is in accordance with NUREG-0588_ which suggest the use of an equation of the form of 6.3 but- with a natural convection heat transfer coefficient. That is, the NUREG guidelines suggest that the following equation should be used at all times when not in the condensing heat trans-fer mode:
q/A = h (6.4) c IIv - T,)
where h e = convection" heat transfer coefficient.
Equations 6.4 and 6.3 are identical if h c =2.
l The CONTEMPT treatment in this regard is non-
, conservative. However, the mode of heat transfer throughout most of the 8000 second run time is con-( densing heat transfer. For most heat sinks, the mode of heat transfer is condensing throughout the entire run time. Thus, the nonconservative contri-bution is very small.
9 NUREG/CF-1511 , " Containment Main Steamline Break Analysis-Equipment Qualification," reviews the heat sink analysis heat transfer coefficients and bulk temperature assumptions used in CONTEMPT for MSL break analyses. .With; respect to the use of the Uchida condensing heat transfer coefficient, it is
" recommended that this approach be continued pending the development of a more satisfactory approach to the evaluation of condensing heat -transfer. . ." With~
respect to the heat transfer temperature difference, it is suggested that the temperature difference used l
in obtaining the original Uchida test data should be l
()-
6-23
i l
used in analyses. . It is surmised that this tempera-(]) ture difference is most probably.the bulk-to-wall l
! value. Further justificati'on for this is also pro-vided based on the containment heat -transfer boun-l dary layer. " Heat icansfer is dominated by the l air-steam boundary layer and therefore is f ar from a steam condensing dominated process. As-a result, the use of the saturation temperature, except for LOCA conditions where it does not matter, is shown
< to be inappropriate." bensitivity studies using (T sat - T,311) versus (T bulk - Twall) show that the latter results in 'close r predictions of the CVTR test data. On the basis of the above argu-ments, the bulk-to-wall temperature difference is-4 more appropriate. Since the bulk-to-wall assumption results in significantly higher heat transfer rates than the saturation-to-wall temperature dif ference used by CONTEMPT, it can be concluded that the CONTEMPT analysis results in conservative results with respect to this parameter.
6.2.3.2 Heat Sink' Condensate Treatment NUREG-0588 specifies that a maximum of 8 percent of the condensate formed on the heat sinks may be i assumed to remain in the vapor region when tne !
atmosphere is superheated. It further specifies
' {
that *.ne amount of condensed mass should be cal- j j
culated'from the following equation:
M =X
- g/h y -
h (6.5) cond Mcond = Mass condensation rate !
X =. Mass condensation fraction (0.9 2 recommended) g = surface heat transfer rate O
\ 6-24
, , ,, --a-e ., .-e.- , . - - - --
hy
(]} =~ enthalpy of the superheated steam hg = enthalpy of liquid condensate entering the sump region In CONTEMPT, the rate of mass tr,ansfer from the vapor region to the liquid region as a result of heat sink condensation, when the Uchida condensing
~
' heat transfer option is used, is the same as equa-tion 6.5. For superheated conditions for which the heat structure wall surface temperature is greater than the saturation temperature, no mass transfer.is calculated.
9 NUREG/CR-1511 points out that relatively small variations in the calculation of the mass and energy of the steam in the compartment atmosphere can pro-duce large changes in the extent of superheat.
O'- If a model overpredicts the condensation mass removal rate, it will decrease the steam mass and increase the steam specifi; internal energy re-sulting in conservatively.high temperatures.
9 NUREG/CF-1511 specifically addresses the conden-sation mass removal model in CONTEMPT and concludes that it results in an overestimation of mass removal by not accounting for the energy necessary to cool the air in the boundary layer at the wall surf ace j
and the energy to cool the steam that does not con- H dense. A modified model is proposed and a series of sensitivity studies are performed. The results show that the CONTEMPT condensation mass removal model results in the highest calculated temperatures with respect to the CVTR test data and is thus conser-vative.
O 6-25
( ). It should be noted that the use of a mass condensate fraction .of 0.9 2 reduces the CONTEMPT condensate mass removal by 8 percent and results in tempera-tures which more closely approximate CVTR test results.
9 In the NrJREG/CR-1511 studies, it was found that the CONTEMPT mass removal method resulted in an overprediction of the CVTR measured data by approxi-mately 60 *F. Thus, a condensate mass removal frac-i tion of 0.92 is still conservative.
6.2.3.3 Analysis Results With Heat Sinks The effect of the heat sinks on the drywell vapor temperature response for the 2.0, 0.75 and 0.01 2
ft breaks is substantial as can be seen in
() Figures 6-8 through-6-10. These figures provide overlay plots of the drywell vapor temperature with and without heat sinks for each of the cases analyzed. Throaghout most of the 8000 seconds con-sidered, the temperatures are within the 275*F to 300*F range. When heat sinks are ignored, the dry-well vapor temperatures are within the 320'F to 335'F range throughout most of the problem time.
The greatest reduction in temperature is found to 2
occur for the 0.01 ft break. This is expected l since the blowdown mass flow rate is quite small for this case and the effect of a given amount of energy and mass removal by the heat sinks would be more pronounced than for larger break sizes. Heat sink heat removal and its associated mass condensation rate is a temperature difference dependent phenome-non. The .01 ft 2 break results in higher bulk temperatures _than the larger breaks. In this way, U<s 6-26
~ J 4
l one would expect a large benefit from' the heat sinks i during the early part of the analysis, with a some-what diminished benefit later as the bulk tempera-ture is reduced by the heat sinks. 2 The 2.0 ft break would provide the worst results except for the fact that the blowdown goes tso phase for a small period of time. This diminishes its effects. Thus, the most severe break with heat sinks considered is the 0.75 ft 2 break. This is the largest break which does not result in two phase blowdown. The peak temperature during this break is 311 F at 480 seconds as can be seen in Figure 6-9 . The tempera-ture remains relatively flat in the 305 F to 298*F range until about 1200 seconds after which a rapid reduction to about 280 F occurs. This is followed
- by a relatively linear increase to 295 F at 8000 seconds. The effe9ts of pressure suppression vent-ing ihich occurs throughout the problem can be seen as small temperature oscillations with an occasional large oscillation throughout the entire analysis
, duration.
The drywell and torus pressures for each of the breaks is provided in Figures 6-11 through 6-13. A review of these figures shows ttat the 2.0 ft2 break results in pressures of 3b to 37 psia through-out the entire analysis interval. This is basically unchanged from the results without heat sinks. The same is true of the 0.75 ft 2 break. Apart from a milder pressurization in the first 1500 seconds, the 0.01 ft break containment pressure is also basi-cally unchanged from the case when heat sinks are ignored. 1 1 [] v 6-27
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- O i- 6.2.4 CONTAINMENT-RESPONSE WITH-HEAT SINKS AND ,
l' CONTAINMENT SPRAY The containment spray model information was dis-cussedi'n section.6.1.5. The.MSL break with the most severe containment response was identified 2 in section 6.2.3.3 to be the 0.75 ft MSL break. j when containment heat sinks are considered. The containment transient severity for this.and all other break sizes is substantially reduced when the containment' spray system is initiated. In light of tie rapid vessel depressurization, scram and subsequent drywell pressure and temperature 2 response to the 0.75 ft MSL break, it is assumed that the operator will manually activate the containment spray system within a period of ten minutes. The resultant transient is illustrated [} in' Figures 6-14 through 6-17 Figures.6-14 and 6-15 show the. pressure and' temperature transient for the first 800 seconds. Figures 6-16'and 6-17 provide the long term response. As can be seen, the initiation of sprays results in a rapid reduction in containment temperatures and pressures. The drywell temperature will almost immediately drop to saturation conditions and then proceed to gradu-ally approach the original 135 F temperature. In the' absence.of shutdown heat exchangers, all decay heat is removed by boiling.to containment. This is a conservative assumption and results in about 9 hours for the containment temperature and pressures-to be reduced by the spray sys'.em to their initial values. ()' 6-34
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fg (_/ 7.0 OYSTER CREEK TEMPERATURE AND PRESSURE PROFILES The Oyster Creek containment temperature and pressure profile to be used for the environmental qualification of electrical equipment inside con-tainment is derived from the most severe MSL' break response with heat sinks and containment spray 2 considered. This is the 0.75 ft MSL break analysis discussed in the previous section (6.2.4). The results for this case are repeated in Figure 7-1. This plant specific analysis represents a significant reduction from the 346 F for 6 hours recommended in . NUREG-0588. The major reasons for the departure from the NUREG-0588 generic profile are the con-sideration of containment heat sinks and the initiation of containment spray. The use of the shutdown heat exchangers would greatly reduce the ()
' time required to reach the initial containment temperatures and pressure conditions. The shutdown cooling system can be initiated after the reactor ccolant temperature is below 350 F and pressure has been reduced to below 150 psig. The use of this system has not been considered in the Figure 7-1 temperature profile.
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i REFERENCES 8.0 A
\l 1. Guidelines for' Evaluating Environmental Equipment Quali-fication of Class lE Electrical Equipment in Operating Reactors - Enclosure I to NRC Letter to Licensees dated February 15, 1980.
- 2. NU REG-0 5 8 8. Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment.
Published December, 1979.
- 3. Oyster Creek, FDSAR, Amendment No. 32, Figures 3-1, 3-2.
- 4. Performance Evaluation of the Oyster Creek Containment Spray Heat Exchangers, GPU TDR 165, May 1980.
- 5. Oyster Creek, FDSAR, Amendment No. 5, Figures III-1-1, III-1-2.
() 6. EPRI CCM-5, RETRAN - A Program for One-Dimensional Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems. ]
- 7. G. F. Niederauer. CONTEMPT-EI; A Computer Program for Predicting Containment Pressure - Temperature Transients.
EI-77-18, December 1977.
- 8. McAdams, W. H. Heat Transmission. McGraw-Hill, 1965, Chapter 7.
- 9. NUREG/CR-1511 (LA-8305-MS), Containment Ma n Steam Line Break Analysis for Equipment Qualification Informal Report, June 1980.
- 10. P. S. Smith, Graphical Compilation of Oyster Creek RETRAN MSLB Analyses for Environmental Qualification of Equipment, Supplement 1 to TDR #180, GPU Service Corp., October, 1980.
O 8'-l
REFERENCES (Continued) O l
- 11. XN-75-55 (Rev. 2); The Exxon Nuclear Company; WREM-Based NJP-BWR ECCS Evaluation Model and Application to Oyster Creek,' April 1977.
O i l O 8-2
O CIIAPTER 3 OYSTER CREEK NUCLEAR GENERATING STATION ENVIRONMENTAL QUALIFICATION OF ELECTRICAL EQUIPMENT Environmental Effects on Safety Grade Electrical Equipment Due to LOCA and High Energy Pipe Rupture O l l l l !v 1 l l _
l Report No. 02-0370-1045 Revision 0 l l TABLE OF CONTENTS
.E!!ge.L
- 1. 0 Executive Semay 1 2.0 Thermal-Hydraulic Analysis 4
< 2.1 Introduction 4 2.2 Building Descriptions 4 2.3 Selection of Line Breaks 5 2.4 Reactor and Turbine Building Model 6 Construction 2.5 Key Analytical Assumptions 7 2.6 Summary Results 9 4 2.7 References 9 i O 3 .0 Radiological Analysis 10 . 3.1 Areas Outside of Containment 10 3.1.1 Introduction 10 3.1.2 Analytical Methodology 10 3.1. 3 Analysis Results and Discussion 13 3.1.4 References 14 3.2 Areas Inside Containment 14 3.2.1 Introduction 14 3.2.2 Analytical Methodology 14 3.2.3 Analytical Results and Discussion 16 i 3.2.4 References 17 Tables 1 through 5 Figures 1 through 58 Appendix A EDS FLOW Computer Code Abstract Appendix B Recommendations for Consideration by BWR Owners Group O
i 4 Page 1
- Report No. 02-0370-1045 l Revision 0 s
i
- 1. 0 Executive Summary The Nuclear Regulatory Commission (NRC) has recently man-4 dated in letters to utilitias that the environmental qualification of electrical equipment required to ihnction subsequent to 4 certain poshdatad accidents be reviewed. The postulated ac-cidents include a LOCA inside containmaat or a High Energy Line Break (HELB) inside or outside of containment.
The qualification conditions to be considered include the following:
- 1. Post accident pressure, temperature, and humidity
, conditions
- 1 l 2. Post accident raritation exposure O 3. Exposure to post accident chemical spray i
- 4. Submergence i
i i 4 At a March 17, 1980 meeting, the Jersey Central Power and
- Light Company (JCP&L) requested EDS to execute the analyses necessary to adequately define the post accident service con-dition profiles for electrical equipment whose designation and plant location were specified by JCP&L. The EDS scope of work included the following tasks:
J. Develop pressure and temperature time histories for plant areas outside of containment based upon postulated line breaks in the following systems specified by JCP&L: Main Steam System Main Feedwater dystem Reactor Cleanup System
1 Page 2 Report No. 02-0370-1045 Revision 0 l l 1
- Emergency Condenser System l 1
- 2. Calculate radiation levels at component locations selec-
- ted by JCP&L in plant areas outside of containment.
- 3. Calculate radiation levels inside containmant.
i j The thermal-hydraulic analysis to develop the pressure and
, temperature service condition profiles was performed utilizing l the EDS proprietary computer code EDS FLN. A capsule description of EDS FLN is included as Appendix A.
The thermal-hydraulic results are summarized on Table 1 and supporting Figures 1 through 18, which indicate the various temperature time history profiles. A temperature profile was generated in cases where compartment temperature exceeded ^ O tooor-
; Radiological analysis was conducted utilizing source terms 1
furnished by JCP&L, the computer code QAD-5PA, and in some cases hand calculations. Table 1 indicates the total in- ! I tegrated radiation exposure one year subsequent to a LOCA at a number of component " targets" specified by JCP&L. Table 2 summarizes analysis executed to define the total one year integrated radiation exposure to equipment located inside containment subsequent to a LOCA. The esposure is divided into the following contributory components: i , 1. Reactor Vessel Streaming I
- 2. Containment Airborne Activity
; Drywell Sump Activity The above components are algebraically summed in total or partially dependent upon the equipment specific location with-in containment.
O
.y._, -- - _y,., .
,.- , _ ..,_ ---, _. . , . . , , , - -.----r a-*w'<
4 O >==e a s Report No. 02-0370-1045 i Revision 0 l The results of the above discussed analyses, as summarized on Tables 1 and 2 and Figures 1 through 18, provide JCP&L with an accurate prediction of environmental conditions exist-ing in plant areas subsequent to certain postulated accidents. This information can be used to assess the capability of exist-ing equipment in terms of qualification, and/or as the source document for preparing an equipment c,ualification specifica-tion. e-i O O
l
>== 4 O- 1 s Report No. 02-0370-1045 i Revision 0 l J
2.O Thermal-Hydraulic Analysis 2.1 Introdmetion The NRC requirea that the capability of safety related equip-ment to function in harsh environments, which may exist following postuisted pipe break accidents,' be evaluated. A part of this evaluation includes defining the post accident pressure and temperature profiles. This section details the analytical methodology employed by EDS to develop temper-ature and pressure profiles for the Oyster Creek Nuclear
- Generating Station in plant areas and compartments located l outside of containmant.
., The EDS proprietary computer code EDS FLOW wr.s utilized as the primary tool for this analysis. An abstract describing the EDS FLOW code is included as Appendix A. t The following paragraphs in this section will describe in de-l tail the methods, analysis, and results of the thermal-hydraulic effort. l 2.2 Building The reactor building is essentially a six floor structure a Descriptions ranging from the " corner rooms" at elevation -19' to a full l floor at elevation 119'. The building structures are shown on - i Figures 19 through 25, and include locetions of the electrical equipment specified by JCP&L. The floors are connected by stairwells, HVAC duct shafts, pipe shafts, cable shafts, and floor openings for equipment i installation and removal. These connections allow free communication of steam between floors and compartments located on any individual floor. The turbine building is basically a three level structure in-ciuding the basement floor which houses the reactor feed pump room, a mezzanine floor, and the operating floor where the turbine-generator is located. The turbine building struc-
- tures are shown on Figures 26 through 31, and include loca-
~ .h
- tions of electrical equiprwat identified by JCP&L in the reactor feed pump room and in,the area of the main steam turbine admission valves.
- r. . - - -_n--, , - _ - . ..m-.,-- .,, , , -,.7 v- ,,.__ m-- e w----- .
O Page 5 s Report No. 02-0370-1045 Revision 0 As for the reactor building, the three turbine building floors are connected by stairwells, HVAC duct shafts, pipe shafts, cable shafts and floor openings for equipment installation and removal. These connections allow free communication of steam between floors and compartments located on any in-aividual floor. A plant visitation was mada by EDS engineers to verify that th3 structures shown on the Oyster Creek General Arrange-ment drawings accurately reflected the actual plant configura-tion such that precise reactor and turbine building computer models could be constructed. 2.0 Selection of as specified by JCP&L, line breaks were considered in the Line Breaks following four systems:
- 1. Main Steam System
- 2. Reactor Feedwater System
- 3. Emergency Ccndenser System
- 4. Cleanup and Demineralizer System The piping systems were located on the general arrangement drawings utilizing piping isometric and arrangement drawings received from JCP&L. A total of five double ended guillotine piping breaks were assumed at the worst locations from an equipment qualHication viewpoint; i.e. , breaks were located in areas where the environmental response was most severe.
Piping stress levels were not considered as an input,to bmak i locatior.- l Figures 32 through 36 show the piping systam arrangements and the selected break locations. As shown on Figures 52 and 36, two main steam line breaks were evaluated. The break indicated on Figure 32 is located within the steam tunnel at elevation 23'-6", and was selected to evaluate the effect of pressure and temperatum on the steam tunnel, torus room, and connecting corner rooms at the -19' elevation.
Page 6 Report No. 02-0370-1045 j Revision 0 The main steam line break was selected in lieu of the main feed line break in this area based upca energy release con-siderations. The main steam line break indicated on Figure 36 was selected to evaluate the pressure and temperature effects on the IB-10 temperature switches located at the turbine side , of the steam tunnel. ' Figure 33 indicates the location of the worst case Cleanup System line break at elevation 51-3". It should be noted that only the high energy sections of the system were considered in the analysis. For the Cleanup System, the high energy portions include piping segments from the reactor vessel to the non-regenerative heat exchangers, and from the non-Q regenerative heat exchangers back to the vessel. Figure 34 indicates the location of the worst case Emergency ! l Condenser line break at elevation 75'-3". This break and the I Cleanup System break produce elevated pressures and tem- i peratures throughout most of the reactor building compart-ments. Figure 35 shows the reactor feed line break at the feed pump discharge. This break was evaluated to assess the environ-mental effects on the RV-23 series pressure switches located within the feed pump room. 3.4 Reactor Building An EDS FLOW computer model was constructed for the and Turbine reactor building and was utilized to evaluate the Cleanup Building System, Emergency Condenser, and Main Steam line breaks Computer Model located within the reactor huildig. All regions of the Construction reactor building were nodalized as volumes, and pertinent portals such as stairwells, equipment hatches, doorways, etc. were modeled as connecting pathways between volumes. Blowout panels were modeled to release steam to the turbine buGding and atmosphem at the 23'-6" elevation steam tunnel and at the 119'-3" elevation respectively. The panels were
l 1 l Page 7 s Report No. 02-0370-1045 Revision 0 l l set to initiate steam relief at a differential pressure of .25 psi (Reference 1). Figures 37 to 43 show the reactor building volume nodalization as indicated by broken lines, and Figure 44 shows the junction flow pathways between 4 volumes.
'In addition to the reactor building model, two models were con-structed for the turbine building to evaluate the FMn Steam j Line and Reactor Feed Line breaks. Separate models were constructed for each break to minimize computer time, and to mawimime the respective environmental responses.
The turbine building was nodalized utilizing the same methM-l ology as for the reactor building with volumes, junctions, e - 3 O 81e e=* 9anels. The turbine b 11 dine vol me neaalization for the main steam line break is shown on Figures 45 through 50, with Figure 51 indicating junction flow pathways between vol-umes. In the same fashion, the turbine building was nodalized for the reactor feed line break as indicated on Figures 52 through 57 with junction flow pathways between volumes shown on Figure 58. These models comprised the Amdamental luput to the EDS FLOW 4 computer code. Pertinent input assumptions are discussed in the following section. 2.5 Key Analytical The following assumptions and approximations were used in this Assumptions analysis to produce the final results:
- 1. Initial reactor and turbine building pressure, tempera-ture, and humidity were assumed to be 14.7 psia, 700F, and 35 percent, respectively.
- 2. All lines were assumed to be at their maximum operating pressures and temperatures.
. O.
l l l l Page 8 s Report No. 02-0370-1045 Revision 0
- 3. Maximum main steam line break flow was limited to 200 percent of the 7.15 x 106 LB/HR mein stemming rate by flow restrictors (Reference 2). The break was assumed to isolate at 11.5 seconds. Main staam 4
line isolation was assussd to occur at 10.5 seconds (Reference 1), and an additional second was added to
- account for reverse flow from the turbine.
- 4. The Emergency Condenser and Cleanup System line i brer.k flows were calculated n*ilt ing the Henry Fausk5 Critical Flow Tables, and were assumed to isolate at 60 seconds (Reference 2).
- 5. The reactor feed pump line break flow was approxi-mated utilizing the Henry Fauske Critical Flow Tables at saturated liquid conditions as described in paragraph
- O E.2 of Aust 58.2 (Reference 3).
j The break was assumed to be fed by all three feed i onmps for a period of 25 seconds. Subsequent to the 25 second period, the break fluid was assumed to be
; low energy subcooled water not contributing signifi-cantly to any ihrther compartment pressurization or temperature increase. This assumption is based upon the fact that the feedwater is heated from the
- main condanner hotwell temperature to 3150F at the high pressure heater outlet exclusively by turbine extraction steam. Subsequent to a feed line break, 4
MSIV closure occurs at 25 seconds terminating ex-traction steam flow and any significant feedwater heating (Reference 2).
- 6. All doors were assumed to remain closed if compart-ment pressurization applied force in the direction of the doar jamb. All other doors were assumed to open -
~a t a diferential pressure of 4 psi.
- 7. Forward and reverse flow shock losses were account-
!~ 9d for where junction flow areas were reduced or ex-i panded in relation to the adjacent volume flow areas.
- , ~_. _ _ . _ _ . , _ , . _ _ _..
Page 9 -l s Report No. 02-0370-1045 Revision 0 2.6 kmma ry The EDS FLCW Computer Code was executed for the five se-Results lected line breaks. The sunmary results are shown on Table 1 under the columas entitled " Peak Temperature" and " Peak Pressure". Ar hidicated, many of the compartments within the reactor building and turbine building are exposed to ele-vated temperatures. No appreci@le increase in pressure is evidenced except for a 21 psia pressure increase in the steam hmnal and a 28 psia pressure spike in the area of the main i steam turbine admission valves. Temperature time history profiles to 600 seconds were de-veloped in cases where electrical equipment was exposed to a compartment temperature above 1000F. These temperature profiles are shown on Figures 1 through 18 and correspond to the Figure numbers listed below the " Temperature Profile" O cot === or r ate 1-2.7 References 1. Record of Conversation dated March 21,1980, J. Knubel to M. Ballard and J. Moran, Oyster Creek Nuclear Plant Subcompartment Analysis.
- 2. Record of Conversation dated April 7,1980, J. Knubel to M. Ballard, Oyster Creek Nuclear Plant Sub-compartment Analysis.
- 3. AIGI 58.2, dated November 1978, Design Basis for Protection of Nuclear Power Plants Against Effects of Postulated Pipe Rupture.
O l l I L
Page 10 Report No. 02-0370-1045
, Revision 0 3.0 Radiological Analysis 3.1 Areas Outside , of Containment
] 3.1.1 . Introduction EDS has calculated post-accident integrated radiation ex-posures at specified vital equipment locations in plant areas , outside of containmant- . These calculations were performed by using the computer program QAD-PSA and simpitfied i manual techniques. In all instances, accident case source
. terms provided by JCP&L were utilized.
! 3.1.2 Analytical Post accident radiation exposures were calculated at the Methodology vital equipment target locations specified by JCP&L as shown in Figures 19 through 25 The area surrrumdimr each target location was reviewed in order to determine Q applicable radiation source components (piping, heat ex-changers, . . .) and shielding. The total target radiation exposure is then equal to the sum of exposures due to all applicable radioactive sources. ) In order to calculate exposures due to radioactive piping, the 1 following procedure was utilized. P&ID's for systems poten-ially contaminated after an accident were marked up to in-dicate radioactive piping. These systems were supplied by JCP&L and are listed in Table No. 3. Non-flowing lines connected to contaminatad lines were considered contaminatad up to the first closed or block valve. For those systems which would be isolated after an accident, piping was consid-
- ered contaminated up to the first isolation valve.
After the P&ID's were marked-up, the contamina+ad piping information was transferred to the Oyster Creek piping . drawings in order to provide physicallocations of radio-4 active piping. These drawings were used as the basis for modelling computerized radiation exposure rate calculations for piping sources subject to the following assumptions. Radioactive piping with diameters of three inches or less was not modelled unless it was determined that such piping s was a major exposure contributor. Contaminated piping
Page 11 Report No. 02-0370-1043 Revision 0 l with diameters greater than three inches were not modelled j if located outside a radius of twenty five feet around each vital equipment target unless it was determined that such piping was a major exposure contributor. Major exposure contributors were defined as the only radioactive pipe in a : target area, the radioactive pipe of closest proximity to ! the target, or a large diameter radioactive pipe in the tar-get area. R was further assumed, in order to account for pipe wall attenuation that all piping was scheditle 40. Exposures due to potentially contaminated non-piping com-ponents wera also considered. The identified radioactive components in each target area were modelled in order to perform mmial or computerized exposure rate calcula-tions. Shielding credit was taken for component structural materials as appropriate. O All piping, and some non-piping component exposure rate calculations were performed by utilizing the computer pro-gram QAD-PSA. This three dimensional code uses the
" point-Keruel" method to calculate gnmma-ray trans-mission through various shield configurations. This method ;
involves representing the source volume by a number of ' point isotropic sources and computing the line-of-sight . distance from each of these source poims to the detector (target) point. From the distance through each shielding region and the attenuating characteristics of the shielding materials, the geometric atteratation and material attenuation are determined. The transfer of energy by the uncollided flux along the line-of-sight path is then com-bined with the appropriate buildttp factor (to account for the contribution from the scattered photons) and integrated over the source volume for each source energy considered. Energy group dependent flux-to-exposure rate conversion factors are applied, exposure rates are summed over all energy groups, and a total exposure rate is calculated.
- In all target exposure calculations the exposure rate at one i p v
hour into the accident was calculated t*ing into account distance attenuation, structural walls if present, and for l water sources, self-shielding within the source volume. l
Page 12
, Report No. 02-0370-1045 Revision 0 I
, A total target one hour exposure rate was then calculated as the sum of the one hour exposure rates due to all significant radioactive sources in the target area. An integration factor-was then applied to the total one hour exposum rate resulting in a one year total target integrated exposure as discussed below. I To determina the effect of the passage of time on the dose
- rate, a series of calculations wem mada as=mimr radio-active decay times of up to one year. Using a single pipe size for liquid sources and another for gaseous sources, the dose rate was computed at the midpoint of an unshielded 10 foot pipe at a distance of 5 feet. Table No. 4 presents the 4
results as relative dose factors,1.e. the dose rates are normalfzed to unity at one hour. These tables were used to compute the doses at times other than one hour after the O aooide=t. with i=tervotatio= a=a *- elatio hei s used as necessary. To conservatively estimate the integrated doses at a target, a table of integrated dose factors was prepared from the relative dose factors. R was assumed that the relative dose factor decreased linearly, rather than exponentially as is the actual case. Thus the average relative dose for any period of time is the arithmetic average of the end points. The relative integrated dose is simply the product of the average relative dose and period of time over which the average is taken. The sum of the relative integrated dose over all time periods gives the cumninHve relative inte-grated dose. Since this is normalized to the dose rate at one hour, the integrated dose may be determined by multi-plying the integrated dose factor by the dose rate at one hour. The gaseous and liquid source term integration dose factors for periods out to one year are also shown in Table No. 4. All radiation source terms used in this analysis were
- supplied by JCP&L. The gaseous activity source term is O composed of 100 percent of the core noble gases and 25 percent of the core halogens. The liquid activity source term is composed of 100 percent of the core noble gases, 50 percent of the core halogens, and 1 % of the remaining core fission products.
, -, - - - - ---.----a r+-- , . , - , - , ,--p- v- .- , - -
f l l Page 13 Report No. 02-0370-1045
- l. Revision 0 i
t
/
- The liquid and gaseous source terms were utilized as required and homogeneously mixed in the dilution volumes presented in
{ Table No. 5. Note that the Clean Up Deminaralizer System is 4 included in this listing. Even though no cleanup system is used i following an accident, the lines of this system were assumed can+= min =+ad up to the first isolation valve. 3.1.3 Analysis The one year integrated radiation exposure for each of the Results and identified targets is listed in Table No.1. These values are Discussion the total seen by the target, i.e. , the sum of exposures re-sniting from all significant radiation sources in the target vicinity. Some of the calculated exposures may be greater than the radiation tolerance of the Class 1E component the target rep-resents. Two methods exist, other than ins +=11ina localized 2 shielding or equipment / piping rerouting, for reducing the cal-culated integrated radiation exposures. The first method con-sists of determinina the time interval over which the com-ponent must function (presumably less than the one year chosen . In this analysis), and applying the appropriate integration factor to yield an integrated dose over a shorter period of time. The second method consists of reducing the source term used, for instance, eliminn+ina the double counting of noble gases. Guidance on source term refinement may be obtained from the " Recommendations for Consideration" developed by the BWR Owners Group Subcommittee on Shielding at the April 16,1980 meeting held in San Jose, California. These recommendations are attached as Appendix B. Note - These recommendations must still be evaluated by the full owner's group. O
, Page 14 Report No. 02-0370-1045 Revision 0 3.1.4 References 1. "QAD-PSA: A Point-Kernel General Purpose i Sheilding Program", originally written by . Los Alamos Scientific Laboratory and converted to the IBM-360 computer by Oak Ridge National Laboratory, July,1968.
- 2. Enclosure 3, " Calculation of Potential Post Accident Dose Rates at the Oyster Creek Nuclear ~
Generating Station for NUREG-0578", to JCP&L 1etter to the United States Nuclear Regulatory Commission dated January 4,1980. 1
- 3. " Piping Design and Engineering", ITT Grinnel Industrial Piping, Inc. , Fifth Edition,1976.
- 4. " Reactor Shielding Design Manual", edited by O Theodore Rockwell III, D. Van Nostrand Company, Inc. , First Edition,1956.
3.2 Areas Inside of Containment 3.2.1 _ Introduction EDS has calculated post-accident radiation exposures for vital equipment located inside the contninment. In addition, the radiation exposure contribution due to the station's normal forty year operation has been considered. These calculations were performed by using the computer program QAD-PSA and simplified manual techniques. Accident case source terms provided by JCP&L were utilized. l 3.2.2 Analytical Post accident radiation exposures inside the contninment ; Methodology were determined by calculating the exposure due to each ; of the contributing sources. For the purpose of this I analysis, it was assumed that exposures within the con-tainment resulted from the following post accident sources: the reactor vessel, airborne activity within the containment atmosphere, and activity contained with-in the water chmped to the drywell sump. The exposure contribution from the reactor vessel over a forty year lifetime was also calculated. i 1
^
t I l l C) Page 15 Report No. 02-0370-1045 Revision 0 In order to calculate radiation exposures inside the contninment due to reactor vessel strenming, a one hour source term composed of 100 percent of the noble gases, the halogens, and the remaindar isotopes was calculated using the source term data supplied by JCP&L. This source term was distri-buted within the region defined by the active volume of the fuel as the source model input into the com-puter program QAD-PSA (a description of this code appears in Section 3.1.2). Shielding credit was taken for the reactor vessel wall, the biological shield, and self-shielding within the fuel region. Exposure rates were calculated at the core mid-m plane external to the reactor vessel and external to U the biological shield. Comparing the reactor vessel dose rate to the one supplied by. General Electric (1.9 x 104 Rads / hour, see Table 12.1.5 of Reference No. 3), a zero hour normalization factor was developed in order to adjust the calculated ex-posure rates to their full power value at zero hours into the postulated accident. This normalization factor (approximately 6) was applied to all exposures resulting from reactor vessel streaming. The cal-culated exposure rate outside the biological shield was integrated over one year accounting for radio-active decay for the accident exposure (see dis-cussion of integration factors in Section 3.1.2) and held constant for forty years for the normallifetime exposure. Radiation exposures resulting from the containment airborne activity were calculated by assuming a tar-get I? cation at the center of a finite spherical source of radius equal to the radius of the containment at its widest location. The contninment airborne activity source term supplied by JCP&L, composed of 100
\m/ percent of the core noble gases and 25 percent of the
\
l l l l Page 16 Report No. 02-0370-1045 Revision 0 halogens for the first day post-accident, and 100 percent of the core noble gases for the remainder of the year, was used and the appropriate integration factors were applied in order to determine the one year integrated ex-posure. The drywell sump was modelled as a cylindrical source 50 feet in diameter and 3 feet high. Evenly distributed within this volume was a liquid source term made up of 50 percent of the core halogens and 1 percent of the re-maindar isotopes. The computer code QAD-PSA was used to caleninta a one hour radiation exporare and the appropriate liquid integration factor was applied to obtain the one year integrated exposure. [] 3.2.3 Analysis Results of the integrated radiation exposure calcula-Results and tions inside the contninment are presented in Table No. Discussion 2 according to contributing source. All values shown are post accident with the normal forty year lifetime exposure included in the reactor vessel strenming com-ponent. It should be pointed out that the integrated ex-posure calcultad for this component applies for con-tninment loculons external to the biological shield. Exposures inboard of the biological shield would be considerably greater. j b.A l l
l I l
- ^s V
Page 17 Report No. 02-0370-1045 Revision 0 3.2.4 References 1. Enclosure 3, " Calculation of Potential Post Accident Dose Rates at the Oyster Creek Nuclear Generating Station for NUREG-0578", to JCP&L letter to the United States Nuclear Regulatory Commission dated January 4,1980.
- 2. "QAD-P5A: A Point-Kernel General Purpose Shielding Program", originally written by Los Alamos Scientific Laboratory and con-verted to the IBM-360 computer by Oak Ridge National La'mratory, July,1968.
- 3. " General Electric Standard Safety Analysis Report-BWR 6", General Electric Company.
- 4. " Reactor Shielding Design Manual", edited by Theodore Rockwell III, D. Von Nostrand Company, Inc. , First Edition,1956.
i
TABLE _1 01ST E R C RE E K_ N[C LE AR GE N;
') Indicates that equipment is not required to mitigate the
# E LECTRICAL EQUIPMENT ENVHION consequences of the accident outside of containment or to 9qhteve a safe shutdown for that accident. For a b( k inside containment, asterisked items are needed to mitigate the accident, however, the enviromnental conditions for these asterisked items would be normal ambient conditions.
Target Approximate , w Designation Description L_ocation Coordinates Elevation g R4 x East Drywell Wall 55' Cle
- 1. IA-83 -A ADS-Pressure Switch ADS-Pressure Switch R3 x East Drywell Wall 72' Cle IA-83-B ADS-Pressure Switch RKO-2 72' Cle IA-83-C ADS-Pressure Switch RKO-1 72' Em IA D IA-83 -E ADS-Pressure Switch RKO-3 55' Err
- 2. V-26-16* Drywell Vent & Purge R2xRp 25' Valve V-26-18* Drywell Vent & Purge R2xRy 25' Valve Containment Spray Valve 33' Cle
- 3. V-21-5* R3xRC 62' Ch V-21-11*J Containment Spray Valve R5 x Drywell Wall
- 4. V-5-167* Reactor Bldg. Closed Lp. R 4 x Drywell Wall 49' j Ch Cooling System Valve V-5-147' Reactor Bldg. Closed Lp.
Cooling System Valve R4 x Drywell Wall 49' Cl. Containment Spray Valve 27' Cl<
- 5. V-21-13 R1 -R2xRC-R9 _
l V-21-17 Containment Spray Valve RG-R7 x RB RC 27' j M: Containment Spray Valve Southeast Corner Rm. - 19 ' - V-21-3 Containment Spray Valve Northeast Corner Rm. -19' M: V-21-9 Containment Spray Valve Southeast Corner Rm. - 10 ' M: V-21-1 Containme'nt Spray Valve Northeast Corner Rm. -19' M: V-21-7 RE-04-A* Drywell Pressure 55' Er
- 6. L Reactor Vessel x R 6~R7 Scram Switch RE-04-B* Drywell Pressure L Reactor Vessel x Rg-R7 55' El Scram Switch Drywell Pressure R x North Drywell Wall 55' El RE-04-C
- Scram Switch Drywell Pressure RE x Noah Drywell Wall 55' Ei RE D
- Scram Switch 4
I
,\
..aL._.
\
i Sq.o STAnos } gNTA1. CONDffIONS i L
?
3 q I i Peak Temperature ( Peak Total Integrated Temperature Profile Pressure Radiatiation Exposure'(1 Yr.); orst Case ,_, , NUREG 0578 Assumptions (RADF (oF) (PSIA) he Break - l' 205 Fig. 10 16 PSIA ' 3. 7 x 10 R mup System ~ inup System 215 Fig 11 16 PSIA __< G.1 x 104 R , anup System 215 Fig. Il 16 PSIA <G.1 x 10 R er. Cond. 230 Fig. 4 16 PSIA 1.4 x 104R
-L 5 er. Cond. 230 Fig. 4 16 PSIA 3. 9 x 10 R i] 5
- 77
- 15 PSIA
- 2. 0 x 10 R
- 5 77 - 15 PSIA
- 2. O x 10 R anup System ' 140 Fig. 13* 15 PSIA
- 5.1 x 10 R 5
nnup System 205 * : Fig. 10* 16 PSIA
- 5.1 x 10 R 4
anup System 950* . 16 PSIA
- 2. 5 x 10 R tanup System 950* -
16 PSIA
- 2. 5 x 10 R ranup Sy "+cm 1400* Fig. 13 i 15 PSIA G. 7 x 10 R 77 - 15 PSIA 3. 7 x 10 R
' 5 Jn R . .
1650 ! Fig. 14 15 PSIA G. O x 10 R 5 in Steam 165 0 ! Fig. 18 15 PSIA G.1 x 10 R
~ ' G dn Steara 1650 Fig.14 15 PSIA 1. 0 x 10 R -
' 5 Lin Ste'im , 1650 Fig. 18 15 PSIA 6.3 x 10 R 0
Ter. Cond. 230 Fig. 4* 1G PSIA
- 1. 5 x 10 R
* ~
- G acr. Cond. 230 Fig. 4 16 PS[A' 1. 5 x 10 g acr. Cond. 230
# Fig. 4* 16 PSIA
- 2. 8 x 105R
- 0 ner. Cond. 230 Fig. 4* 16 PSIA
- 2. 8 x 10 R I
l s
- TA BLE OYS11ER CREEK NUCLEAR E LECTRICAL EQUIPMENT EN
( - .- k - Target Approximate Designation Wtt Description Location Coordinates Elevation g
. ID- 45-A Reactor Vessel Pressure Trans RKO-1 72' Eme II'-45-B Reactor Vessel Pressure Trans RKO-2 72' Cle ID-46-A Reactor Vessel Pressure Trans RKO-1 72' Eme ID-46-B Reactor Vessel Pressure Trans RKO-2 72' Clea . 1G-06-A-1 Isolation Condenser LevelTrans R4 x L of "A" Iso. Cond. 98' Emce IG-06-A-2 , Isolation Condenser _ LevelTrans. 98' R4 x L of "A" Iso. Cond. Emes IG-06-B-1 Isointion Condenser LevelTrans, R,3 x L of "B" Iso. Cond. 98' Emed IG-06-B-2 1solatio.' Condenser Level'Irans. R4 x L of "B" Iso. Cond. 98' Emce IB-(16 switches) Reactor Isolation Temp. Switch Location to be specified by (See Item No. JCP&L ,
31) 1). ID-13A Reactor Water Level Trans RKO-1 72' Emee
' 13 B Reactor Water Level Trans R KO-2 72' Clea LA-12A Reactor Water Level Trans RKO-1 72' Eme IA-12 B Reactor Water Level Trans R KO-2 72' Clea .
- 1. H V-29-A Core Spray Pressure Switch NW Corner Room -19' R V-29-B Core Spray Pressure Switch SW Corner Room -19' R V-29-C Core Spray Pressure Switch NW Corner Room -19' R V-29-D Core Spray Pressure Switch SW Corner Room -19' !
@. . R V-40 -A Core Spray Pressure Switch R x North Drywell Wall 55' Eme D RV-10-B Core Spray Pressure Switch R2 -R3xRE-Ry - 27' ;
. H V C Core Spray Pressure Switch Ro x North Drywell Wall 55' -
Emer [ R V-40 -D Core Spray Pressure Switch R2 -R3xRE-Rp 27' R V-26-A Core Spray Flow Trans. R KO-3 55' E me r R V-26-B Core Spray Flow Trans. 81' R2 -R3xRE , Eme - c). IB-06-E Isolation Cond. Area Temp Det. C. g. Condr. NE-01A 115' IB-06-F Emer Isolation Cond. Area Temp Det. C. g. Condr. NE-01 B 115' Eme IB-06-G - Isolation Cond. Area Temp Det. 10'S of R4 x Re-RB 90'
- Eme IB-06-R Isolation Cond. Area Temp Det. 10'N of R4 x Re-RB 90' ' Eme ;
). Jn-03-A- Containment Spray Flow Mtr. RC-RB x North RB Wall 25' 03-B Containment Spray Flow Mtr. RC-RB x South RB Wall 25' Clean; t
t i.
}al .n
G NERATC;G STATION - _
/ IRON.TIENTAL CONDITIONS .
P
- a. .- ,
.l Temperature Peak Peak Total Integrated -
rst Case - Tempe rature Profile Pressure . Radiation Exposure (1 Yr.) e Break (OF) (PSIA) NUREG 0578 Assumptions (RAE
, Cond. .
230 Fig. 4 16 PSIA 1. 4 x .6 R ap System 215 Fig. 11 16 PSIA < s,1 x y oiR-
, Cond. 230 0 Fig. 4 16 PSIA 1,4 x 104 R 215 0 '
Fig. 11 16 PSIA <g,1x10 R- 4 2p System
,'Cond. 270 0 Fig. 5 16 PSIA 5. 3 x 105 R 270 0 5. 3 x 105 g , Cond. Fig. 5 16 PSIA - , Cond. 270 0 ; Fig. 5 16 PSIA 5. 3 x 105 R , Cond. 2700- Fig. 5 16 PSIA 5. 3 x 105 R
- 1. 4 x 104 R
~
Cond. 230 e Fig. 4 16 PSIA 1p System 215 Fig. 11 16 PSIA .< 6.1 x 104 R-Cond. 230 0 Fig. 4 16 PSIA 1. 4 x 104 R ip System 215 0 i Fig. 11 16 PSIA < 6.1 x 104 R l 77 - - 15 PSIA 4,'4 x 105It 77 - 15 PSIA 2. 6 x 105 It
- 4. 4 x 105 R 77 -
15 PSIA 0
, 77 -
15 PSIA 2. 6 x 105 R t 0 5 Cond. 230 i . Fig. 4 16 PSIA 1. 7 x 10 R 77 {i 15 PSIA 2. 3 x 105 R Cond. . 230 0 Fig. 4 16 PSIA 1. 7 x 105 R 0 77 - 15 PSIA 2. 3 x 105R Cond. 2300 j Fig. 4 16 PSIA 3. 9 x 105R 225, Fig. 8 16 PSIA 8. 0 x 105R Cond. Cond. 270 Fig. 5 16 PSIA <c.1 x to4R Cond.- 270g Fig. 5 10 PSIA < 6,1 x lo4 R 280 Fig. 6 6 Cond. 16 PSIA 1. 0 x 10 R Cond. 28,0 i Fig. 6 16 PSIA 9.4 x 105R 77 - + 15 PSIA 3.4 x 105R p System 140 Fig. 13 15 PSIA 1,7x 106 R r i
~
l
TADI.E
- Indicates that equipment is not required to mitigate the OYSTER CREEK NUCLE AR GI consequences of the accident outside of containment or ELECTEICAL EQUIP 31ENT ENVD tojachieve a safe shutdown for that accident. For a break insido containment, asterisked items are needed
.t(Ittigate the accident, however, the environmental chuditions for these asterisked items would oc normal ambient conditions. -
~
Target ~
! Approximate , w Designation Description Location Coordinates Elevation '
y
- 15. IP-15-A* Containment Press. Switch RKO-3 55' E mer IP-15-B* Containment Press. Switch RKO-3 l 55' ' '
Emer I P-15-C
- Containment Press. Switch RKO-3 55' ; Emer I P-15-D* Containment Press. Switch RKO-3 55' i Emer
- 16. IP-07* Drywell Press. Trans. RKO-3 55' Eme I
- 18. R V-4 6-A* Drywell Press. Switch RKO-3 '
55' Eme; R V-iG-B
- Drywell Press. Switch RKO-3 55' Eme.
R V C* Drywell Pres's. Switch RKO-3 55' Eme R V-iG-D* Drywell Press. Switch RKO-3 55' ; E me . i 19.: E A
- MSL Low Press. Switch Reactor Fd. Pump Room x 5' Reac!
North Wall 1 R E-23 -B* MSL Low Press. Switch Reactor Fd. Pump Room x 5 Rene South Wall ~ R E C* MSL Low Press. Switch Reactor Fd. Pump Room x 5' ! Reac North Wall i H E D* MSL Low Press. Switch Reactor Fd. Pump Room k 5' ! Reas South Wall
- 20. RE-22-A*. Reactor Isolation Switch RE-Rp x DrywcH WaH 27' RE-22-B
- Reactor Isolation Switch /RE-RF x Drywell Wall l 27' RE-22-C
- Reactor Isolation Switch RE-Ry x Drywell Wall 27' RE D
- Reactor Isolation Switch RE-RF x Drywell Wall 27' RE E
- Reactor Isolation Switch RE-RF x Drywell Wall 27' RE F
- Reactor Isolation Switch R E-RFx Drywell Wall 27' RE G* Reactor Isolation Switch R E-Rpx Drywell Wall 27' RE-22-H
- Reactor Isolation Switch RE-Rp x Drywell Wal! 27'
!1. IB A 1. Isolation Condenser d P Switch RKO-3 55' E m@) IB-05-A2 Isolation Condenser O P Switch RKO-3 55' I Em@; IB-05-B1 . Isolation Condenser AP Switch RKO-3 .55' Em@l Ul-05-B2 Isolation Condenser AP Switch RKO-3 55' Em@)
, 1-11-Al Isolation Condenser AP Switch RKO-3 55' Em@ ' }3-11-A2 Isolation Condenser OP Switch RKO-3 ,
55' Em@ Ip-11-B1 . Isolation Condenser d@ Switch RKO-3 ! 55' Em@
- lb-11-B2 . Isolation Condenser AP Switch RKO-3 55' Em@
, i
SEILVIt.G STATION IDS.TII'i.TAL CONDITIONS p
?
i i j. I:
~
i Peak Temperature Peak i Total Integrated ! Temperature Profile Pressurc {rst Case . Radiation Exposu res (1.Yr. ' je Break , (OF)- (PSIA) NUREG 0578 Assumptions (RAF Cond. 230
- Fig. 4
- 16 PSIA
- 3. 9 x 105R
~
Cend. 2300
- Fig. 4
- 16 PSIA
- 3. 9 x 105R Cond. 2300
- Fig. 4
- 1G PSIA
- 3. 9 x 105R Cond. 230
- Fig. 4
- 16 PSIA
- 3. 9 x 105 g
, Cond. 2300* Fig. 4* 16 PSIA
- 3. 9 x 105R I i Fig. 4* 5 I i Cond. 2 3 0
- 16 PSIA
- 3. 9 x 10 R -
LCond. 2300
- Fig. 4* 1G PSIA
- 3. 9 x 105g - ?
2300
- Fig. 4* 16. PSIA
- 5 3 Cond. 3. 9 x 10 R 2300
- Fig. 4* 1G PSIA
- 5
. Cond. 3. 9 x 10 R or Feed 21@
- Fig. 2* 23 PSIA * < 6.1 x 104R '
21@
- Fig. 2* 23 PSIA
- 4 nr Feed < 6.1 x 10 R 21@
- Fig. 2* 23 PSIA
- 4 or Feed < 6.1 x 10 R 21f
- Fig. 2* 23 PSIA
- 4 or Feed < 6.1 x 10 R I
79* I 15 PSIA
- 4 m
< 6.1 x 10 R 7f* 15 PSIM 4
< 6.1 x 10 R m 7F* -
15 PSIA * < 6. I x 10 R
, 7F* -
15 PSIN < 6.1 x 10 R
, 79* -
15 PSIX < 6.1 x 10 R m - 7F * ' 15 PSIA * < 6. I x 104 R 7P* 15 PSIA 4
< 6.1 x 10 R
_ 7F* - 15 PSbf < 6. 7. x idI R
. Cond. 23 @ Fig. 4 16 PSIA 3.9 x 105R 23 @ Fig. 4 16 PSIA 3. 9 x 105 g
.-Cond.
. Cond. 2300 Fig. 4 16 PSIA 3. 9 x 105R
. Cond. 23 @ j Fig. 4 - 16 PSIA 3.9 x 105R
. Cond. 23$ Fig. 4 -16 PSIA 3. 9 x 105R F
. Cond. 23 @ Fig. 4 16 PSIA 3. 9 x 105R !
. Cond. :2300 ' Fig. 4 16 PSIA 3.9 x 10 57t 230. Fig. 4 16 PSIA 3. 9' x 105 R -
). Cond. , ~ l I
TABLED ? Indicates that equipment is not required tc mitigate the OlSTER CREEK NUCLEAR 5) lonsequences of the accident outside of co. minment or E LECT RICAL EQUIP 31E NT E NV'80 % aclteve a safe shutdown for that accident. For a 1'reakhnside containment, asterisked items are needed - --- , M 7 "zgate the accident, however, the environmental Jon-.61ons for these asterisked items would be normal kmbient conditicas, i Target Approximate Worst _Desier.ation Description Location Coordinates Elevation l Lineh @. RE A Core Spray Press. Switch RKO-1 72' ! E me r. RE-17-B Core Spray Press. Switch RKO-2 72' Cleanug RE C Core pray Press. Switch RKO-1 72' Emer. ( RE-17-D Core Spray Press. Switch RKO " ; 72' Cleanug 5 ED. RE-15-A Reactor Vessel Press. Switch RKO-1 72' Emer. d RE-15-B Reactor Vessel Press. Switch RKO-2 72' 4 Cleanup RE C Reactor Vessel Press. Switch RKO-1 72' Emer. q RE D Reactor Vessel Press. Switch RKO-2 -72' , Cleanug
- 4. RE-03-A Reactor Pressure Switch RKO-1 72' E me r. (
RE-03-B Reactor Pressure Switch RKO-2 72' Cleanu3 RE-03-C Reactor Pressure Switch RKO-1 72' Emer. ( RE D Reactor Pressure Switch RKO-2 72' Cleanug $3.'s .E-05-A Reactor Water Level Switch RKO-1 ; 72' ' E me r. ( R E B Reactor Water Level Switch RKO-2 i 72' ; Cleanug i
- 4. RE-02-A Reactor Water Level Switch RKO-1 72
l E me r. ( RE-02-B Reactor Water Level Switch RKO-2 72' ! Cleanug RE-02-C Reactor Water Level Switch RKO-1 72' i E me r. ( RE-02-D Reactor Water Level Switch RKO-2 72' Cleanup
- 7. V-27-1* Purge Valves Top of Torus x NE of 19' hiain Std Reactor Vessel V-27-2* Purge Valves Top of Torus x NE of 19' LIain St(
Reactor Vessel V-27-3* Purge Valves E Reactor Vessel West x 83' Emer. O RE-Rp l
.V-2 7- l
- Purge Valves E Reactor Vessel West x 8?' ! Emer. q RE-Ry i V 13
- Nitrogen Valves ._ E Reactor Vessel West x >
E me r. RE-Rp ' V-23-14* Nitrogen Valves L Reactor Vessel West x 83' Emer. d RE-RF V 17
- Nitrogen Valves E Reactor Vessel West x 83' E me r. d RE-Rp g e IS
- Nitrogen Valves E Reactor Vessel West x 83' Emer. }
RE-RF l '
! l
NERATING STATION NAIENTAL CONDI* IONS r
.- w - - , -
s ! Peak Tempe rature Peak Total Integrated Case Tempe rature Profile Pressu re Radiation Exposure (1 Yr.[ g (OF) _ _(PSIA) NUREG 0578 Assumptions (RA
- ond 230 Fig. 4 16 PSIA 1. 4 x 104R System 215 0 Fig. 11 16 PSIA < 6.1 x 104R
- ond. 230 Fig. 4 16 PSIA 1.4 x 104R System 2150 Fig. 11 16 PSIA < 6.1 x 104R ond. 230 0 Fig. 4 16 PSIA 1. 4 x 104R System 2150 Fig. 11 16 PSIA < 6.1 x 104R
- ond. 230 Fig. 4 16 PSIA 1.4 x 104R Systern 215 0 Fig. 11 16 PSIA < 6.1 x 104R
- ond. 230 Fig. 4 16 PSIA 1. 4 x 104R Syster, 215 Fig. 11 16 PSIA < 6.1 x 104R
- ond. 2300 Fig. 4 16 PSIA 1. 4 x 104R System 215 0 Fig. 11 16 PSIA < 6.1 x 104R
- ond. 230 Fig. 4 16 PSIA 1.4 2: 104R System 215 Fig. 11 16 PSIA < 6.1 x 104R
- ond, 230 Fig. 4 16 PSIA 1.4 x 104R Systern 215 Fig. 11 16 PSIA < 6.1 x 104R
- ond. 230 0 Fig. 4 16 PSIA 1. 4 x iO4R fystem 215 Fig. 11 16 PSIA < 6.1 x 104R
.un 1500* Fig. 15
- 16 PSIA * < G.1 x 104R
.un 1500
- Fig. 15* 16 PSIA * < 6.1 x 104R
- ond. 2500* Fig. 9* 16 PSIA
- 1. 3 x 105R
- ond. 250
- Fig. 9* 16 PSIA
- 1. 3 x 105R
'ond. 250
- Fig. 9* 16 PSIA
- 1. 3 x 105R Cond, 250
- Fig. 9* 16 PSIA
- 1. 3 x 105R
?ond. 250
- Fig. 9" 16 PSIA
- 1. 3 x 105R
'ond. 250
- Fig. 9* 16 PSIA
- 1. 3 x 105R ?
I 1 4 4
h Indicates that equipment is not required to mitigate the consequences of the accident outside of containment or to pchieve a safe shutdown for that accident. For a brdak inside containment, asterisked items are needed to hiitigate the accident, however, the environmental TAPd OYSTER CREEK NUCLE / c{"tions for these asterisked items would be normal an_. .cnt conditions. E LECTRICAL EQUIPMENT Eb Target AppreMmate Designation Des cription Location Coordinates Eleation 1
- 3. - -
V-23-15* Nitrogen System Valves R2-R3 X IIE~ F(by Wall) 33': V-23-16' Nitrogen System Valves R2-R3xRE -RF(by Wall) 33' ; V-2 3-19
- Nitrogen System Valves R2-R3xRE-Rp (by Wall) 33' !
V-23-20* Nitrogen System 7alves R2-R3xRE-Rp(by Wall) 33!-
). V-38-16* Particulate Monitoring System R Cx NW Drywell Wall 27'
~
V-38-17 Particulate Monitoring System R Cx NW Drywell Wall 27' V-38-9 Oxygen Analyzer System RC x NW Drywell Wall 27' V-38-10* Oxygen Analyzer System RC x NW Drywell Wall 27' V-38-22* Torus Sample System RC x NW Drywell Wall 27' , V-3S-23* Torus Sample System RC x NW Drywell Wall 27' - D. [ '3-21* Ventilation Valve Top of Torus x NE Reactor 19' & Vessel Wall V-23-22* Ventilation Valve ' Top of Torus x NE Reactor 19' 2 Vessel Wall V-23-17* Ventilation Valve Top of Torus x Reactor 19' 2 Vessel East Wall' V-23-18* Ventilation Valve Top of Torus x Reactor 19' 2 Vessel East Wall V-2S-47* Ventilation Valve Top of Torus x Reactor 19' 2 Vessel East Wall . -10 (4 switches) Temperature Switch t Steam Tunnel xRG-Rp 32'3" (6' above 9 for Reactor Isolation steam line) J-10 (3 switches) Temperature Switch t Steam Tunnel xRg -RJ 32'3" (6' above p for Reactor Isolation steam line) .)-10 (1 switch ) Temperature Switch 9., Steam Tunnel x RH 32'3" (6 ' above g for Reactor Isolation steam line)
)-10 (4 switches) Temperature Switch E Steam Tunnel x RpxR H 32'3" (6' above g for Rcactor Isolation steam line) 3-10 (4 switches) Temperature Switch E steam Tunnel x By 32'3" (6 ' above g for Reactor Isolation steam line)
\
,t e
I _.______m_.________. - - - - _
1 E1 R GENERATING STATION o 4VIROlGIENTAL CONDITIONS . '
, - q_
Peak Total Integrated Worst Case Peak Temperatu rc ' Temperature Pressure Radiation Exposure (1 Yr) Line Breaks ( F) Profile - PSIA_ NUREG-0578 Assumptions (RAR
~
1 I _ 770* -- 15 PSIA
- 2.9x105 R
_ 770* -- 15 PSIA
- 2.9x105 R
_ 770* -- 15 PSIA
- 2.9x105 R
_ 770* -- 15 PSIA
- 2.9x105 R
_ ' 770** -- ! 15 PSIA 7.6x104 R 77 -- 15 PSIA
- 7.6x104 R
_ 770* -- . 15 PSIA
- 7.6x104 R
_ 770* -- i 15 PSIA
- 7.6x104 R
_ 770* - - - 15 PSIA
- 7.6x104 R
_ 770* -- 15 PSIA
- 7.6x104 R ain Steam 1500* Fig. 15* , 16 PSIA * < 6. lx104 R
- ain Steam 150
- Fig. 15
- 16 PSIA * < 6. lx104 R blin Steam 150
- Fig . 15
- 16 PSIA * <6.~1x104 R lain Steam 150
- Fig. 15
- 16 PSIA * <6.1x104 R (ain Steam 150
- Fig. 15* 16 PSIA * <6.1x104 R '
t [ain Steam 280 Fig. 3 , 21 PSIA <G.1x104 R 2900 Fig. 1 28 PSIA <6.1x104 R [ain Steam 290 . Fig 1 , 28 PSIA <6.1x104 R [ain Steam l 2900 Fig. 1 28 PSIA <6.1x104R [ain Steam 2900 Fig. I 28 PSIA <6.1x104 R- [ain Steam f - I P f I i i t a
F Indicates that equipment is not required to mitigate the onsequences of the accident outside of containment or ~TABLF OWTER Citi?EE NUCLEAR > achieve a' safe shutdown for that accident. For a ELECTRICAL EQUIPMENI' ENV resh inside containment, asterisked items are needed ) mdigate the accident, however, the environmental ~
~~~
onc ms for these asterisked items would be normal mbient conditions. Target Approximate . Worst Ca Designation - Description location Coordinates Elevation Line Bre
- 2. NZ-el-A Core Spray Pump Northwest Cor. Rm. (-) 16 ' -8"
(-) 16'-8" NZ-01-B Core Spray Pump Southwest Cor. Hm. (-) 16'-S" NZ C Core Spray Pump Northwest Cor. Hm. NZ-01-D Core Spray Pump Southwest Cor. Rm. (-) 16'-8" -
- 3. PM-51-1-1 Contmt. Spray Pump Northeast Cor. Rm. (-) 18 '-0" Main SteaJ PM-51-1-2 Contmt. Spray Pump Northeast Cor. Hm. (-) 18 ' -0" Main SteaJ PM-51-1-3 Contmt. Spray Pump Southeast Cor. Rm. (-)18'-0'! Main Stea PM-51-1-4 Contmt. Spray Pump Southeast Cor. Rm. (-) 18'-0 i Main Stea
)
- 4. IP-05A
- R 6 -R 7 29'-0" ! Cleanup S Contmt. Spray Diff.
x RA -Rn Press. Trans. IP-05B
- Contmt. Spray Diff. 29'-0" Cleanup S R6-R7 x RA-RB Press. Trans.
IP-05C* Contmt. Spray Diff. Ri -R2xRB-RC 29'-0" _ Press. Trans. 2T -0" I P-05 D
- Contmt. Spray Diff. R1-R2 x RB-RC Press. Trans. _
15 . NS-04 A -L1 MSIV Solenoid Valves Steam Tunnel 3 0 ' -G" Main Stet WS-01 A-L2 MSIV Solenoid Valves Steam Tunnel 3 0 ' -6" Main Ste: NS-04 A -L3 MSIV Solenoid Valves Steam Tunnel 3 0 ' -6" Main Stel NS-04 B-L1 MSIV Solenoid Valves Steam Tunnel 3 0 ' -6" Maln Ste: NS-04B-L2 MSIV Solenoid Valves Steam Tunnel 30'-6" Main Ste: NS-04B -L3 MSIV Solenoid Valves Steam Tunnel 30'-6" Main Ste; NS-04A-1 MSIV Position Indicators Steam Tunnel 3 0'-6" : Main Ste; NS-04 A-2 MSIV Position Indicators Steam Tunnel 3 0 ' -6" ! Main Str NS-04 B-1 MSIV Position Indicators Steam Tunnel 30'-6" Main St@ NS-04B -2 MSIV Position Ind'eators Steam Tunnel 30'-6" Main Sts Clean-Up Valve 69'-8" Cleanup
- 16. V-lS-2 R1-R2 x RD-RE Clean-Up Valve 69'-8" Cleanup '
V-li-14 R1-R2 x RD-RE V-1S-61 Clean-Up Valve R1 -R2xRD-RE 68'-10" Cleanup'
- 37. V-24-30
- Reactor Water Sample Valve R2-R3 x Re-RE 55 ' -0" Cleanup
- 18. V-17-1
- Shutdown CoolingValve RS -R6xRB-RC 4 6' -2" -
V-1!-2
- Shutdown Cooling Valve R3 -R7 x RB -RC 4 6' -2" - ,
U-17-3* Shutdown Cooling Valve R6-R7 x RB-RC 46 '-2" - 55, Shutdown Cooling Valve RS-R6 x RC-RD 67'-9" -l V-17-56 ' Shutdown Cooling Valve RS-Rs x RC-RD 57'-9" _l t V-1 -57* Phutdown Cooling Valve R5-R6 x RC-RD 57'-9" _l
\
I.
i 1 IE.NEig_i.sG ST ATION I LRONM ENTA L CONDITIONS
' l 6
Peak Total Integrated Peak Temperature Temperature Pressure Radiation Exposure (1 Yr) se Profile PSIA NUREG-0578 Assumptions (RAD g (O F) 77 0 - 15 PSIA 5. 6 x 1051t 77 0 - 15 PSIA 5. 6 x 105R 77
- 15 PSIA 5. G x 105 R 770 - 15 PSLA 5. 6 x 105R Fig. 18 15 PSIA R 165e 6.1 n 15 PSIA 6. 3 xx10 10fR 1 165 Fig. 18 3
I .0 Fig. 14 15 PSIA 1. 0 x 10 6R Fig.14 15 PSIA 1.O x 106R 65 o* Fig. 13* 15 PSIA
- 5
/ste m 140 7.5 x 10 R 5
ystem 140 Fig . 13
- 15 PSIA
- 7.5 x 10 R 5
77 0* - 15 PSIA
- 7.5 x 10 R 770* - 15 PSIA
- 7.5 x 10 R Fig. 3 21 PSIA < 6.1 x 10
.m 2800 < G.1 x 10 R 2S(P Fig. 3 21 PSIA 4 m < 6.1 x 10 R im 2800 Fig. 3 21 PSIA an 2S00 Fig. 3 21 PSIA < 6.1 x 10b tm 280 Fig. 3 21 PSIA < 6.1 x 10b un 280 Fig. 3 21 PSIA < 6.1 x 10 R tm 2800 Fig. 3 21 PSIA < 6.1 x 10 R un 2800 Fig. 3 21 PSIA < 6.1 x 10 E un 280 Fig. 3 21 PSIA < 6.1 x 10 R tm 2800 Fig. 3 21 PSIA < 6.1 x 10k 15 PSIA 4 System 145 Fig. 16 1.5 x 10 R System 145 Fig. 16 15 PStA 1.5 x 10 R Fig. 16 15 PSIA 1.5 x 10 R System 145 Fig. 17
- 15 PSIA
- 4.7 x 10 R System 2150 '
.- 15 PSIA
- 1.4 x 106 g 77o*
15 PSIA
- 1.4 x 10 R 770* -
1.4 x 10 R 77o* 15 PSIA *
- 15 PSLA
- 8.5 x 10 R 77o* 1.1 x 10 R 770*
- 15 PSIA.
1.1 x 10 R 77o* - 15 PSIA
- n i
9 h
S?rdcLihat equipmbat is nm required to mitigate the consequences of the, M ,s i a Ci;: i ' t. ( : 2 /. sdent outside of containment or to achieve a safe shutdown for that 6 dent. - 1 or a break inside containment, asterisked items are needed to ELECTRICAL EQt 11".11:S'l EN) fgate4the accident, however, the environmental conditions for these __ 3riskJd items would be normal ambient conditions. . Ta-St Location Approximate Worst Case - hs1., dtion Description Coordinates Elevations Line Break V-22-1
- Drywell Sump Disch. Valve R5-R6 x RB -R C7 39'-9" -
V-22-2
- Drywell Sump Disch. Valve RS-R g x RB -RC 39'-9" _
V-22-28
- Drywell Sump Disch. Valve R5-R6 x RB -R C 39'-9" ,
t ' V-22-29
- Drywell Sump Diseb. Valve R5-R6 x RB -RC , 39'-9" >
6 l Core Spray Valve RS-RCxRD-RE 65'-6" Emer. Cond. h V-20-15 l V-20-40 Core Spray Valve R5-R6 x RD-RE 58'-10" , Emer. Cond, PS-153* Contmt. Iso. Valve Switch R6-R7 x RD-RE i 55'-0" Emer. Cond. r. [ NZ-03-A
- f. Core Spray Boost Pump R6-R7 x RD-RE 54'-1" i Emer. Cond.)
NZ-03-B Core Spray Boost Pump R2-R3 x RE 26'-4" l Emer. Cond. NZ-03-C Core Spray Boost Pump R6-R7 x RD-RE 54'-1" ;
)
NZ-03-D Core Spray Boost Pump R1-R2 x RE : 26'-4" -
. V-20-21 Core Spray Valve R2-R3 x Rp-RE 83'-4" Emer. Cond V-20-41 Core Spray Valve R2-R3 x Rp-RE 83'-4" Emer. Cond.= . V-14-30 Emerg Cond. Valve R3-R4 x RB-RC 9 0'-0" i Emer Con $
( Vc14-31 Emerg. Cond. Valve R3-R4 x RB-RC 9 0 '-0" Emer. Cond k.14-32 Emerg, Cond. Valve R4-R5 x RB-RC- 90'-0" , Emer. Conh 2 V-14-33 Emerg, Cond. Valve R4 R5 x RB-RC 90'-0" Emer. Con $ V-14-34 Emerg, Cond. Valve R4 R5 x RB-RC i 86'-10" i Emer. Condj V-14 -35 Emerg, Cond. Valve R3-R4xRB-RC l 8 7'-5" Emer. Cond l i
.$. V-11-34 Emerg. Cond. Vahe R4-RS x RA-RB 99'-9" i Emer. Conf V-11-36 Emerg. Cond. Valve R4-R5 x RB-RC 99'-9" Emer. Con $
l l B. RE-18-A Reactor Vessel Level RKO-1 72' Emer. Con % Switch RE-18-B Reactor Vessel Level RKO-2 72' j Cleanup Syd Switch ! l R E-18-C Reactor Vessel Level RKO-1 72' Emer. Conk Switch RE D Reactor Vessel Level RKO-2 72' ' Cleanup Sys
. Switch f 72' R E 19 -A Reactor Vessel Level, RKO-1 l Emer. Co Switch /Trans. .
. RE-05-19-B Reactor Vessel Level RKO-2 72' Cleanup Sy1 Switch /Trans. - - - - -
ek i i s f i
jC!. V; M ST ni i c:- (1]Q.1'1ENTAI, CO.'sblTIOM Peak Peak Total Integrated f Temperature Temperature Pressure ~ Radiation Exposure (p Yr) (OF) Profiles (PSIA) NUREG-0578 Assumptions tRAIP 770* - 15 PSIA
- 6.4 x 10 R 770* -
15 PSIA
- 6.4 x 10 R 77'*0 -
15 PSIA
- 6.4 x 10 R ^
770* - 15 PSIA * . 6.4 x 10 R 230 - Fig. 4 16 PSIA 3.9 x 10 R 5 230 Fig. 4 16 PSIA 3. 9 x 10 3 6 2300* Fig. 4
- 16 PSIA
- 1. 5 x 10 R ,
230 Fig 4 16 PSIA 1. 2 x 10 R 77 - 15 PSIA 2.9 x IO R G 230 Fig. 4 16 PSIA 1. 2 x 10 R 77 - 15 PSIA 2. 9 x 10 R 2250 Fig. 8 16 PSIA 8. 0 x 10 5R 225 Fig. 8 16 PSIA 8. O x 105R , 2800 Fig. 6 16 PSIA 1. O x 106R 280 Fig. 6 16 PSIA 1. 0 x 106R 2800 Fig. 6 16 PSIA 9.4 x 105R 280 Fig. 6 16 PSIA 9.4 x 105R 280U Fig. 6 16 PSIA 9.4 x 10 5R-2800 Fig. 6 16 PSIA 1. 0 x 106R i I 270 Fig. 5 16 PSIA 5. 3 x 105R 270 Fig. 5 16 PSIA 5.3 x 105 R
- 2300 Fig. 4 16 PSIA , 1.4 x 10 b
.e m 215 Fig.11 16 PSIA 46.1 x 10 R 4 . 230 Fig. 4 1GPSIA 1.4 x 10 R 4 Le m 2150' Fig.11 16 PSIA <6.1 x 10 R l3 2300 Fig. 4 16 PSIA 1.4 x 10 R 4 te m 2153 Fig.11 16 PSIA <6.1 x 10 R ] l I
- j i
a
.. Report No. 02-0370-1045 Revision 0
' TABLE NO. 2 OYSTER CREEK NUCLEAR STATION INTEGRATED RADIATION EXPOSURE INSIDE CONTAINMENT Total Integrated Contributing Source Radiation Exposure (RADS)
- 1. Reactor Vessel Streaming 1.57 x 10 (40 Yr. plus accident)
- 2. Containment Airborne Activity 4.14 x 10 7
- 3. Drywell Sump Activity 6.14 x 103 Total 5. 71 x 10 7 Note: All exposures shown utilize the gnmmn-ray source terms provided by Jersey Central Power & Light. Contributions due to beta radiation have not been included.
O
4 Report No. 02-0370-1045 Revision 0 TABLE NO. 3 POITNTIALLY CONTAMINATED SYSTEMS POST-ACCIDENT
- 1. Emergency Cce naer System
- 2. Core Spray System
- 3. Containman* Spray System
- 4. Automatic Depressurization System
- 5. Reactor Water Sample System
- 6. Clean Up Deminaralizer System
- 7. Reactor and Recirculation Loop
- 8. Reactor Shutdown Cooling
- 9. Drywell and Supression System I
i i i i l t
A Report No. 02-0370-1045 U Revision 0 TABLE NO. 4 SOURCE TERM DECAY AND INTEGRATED DOSE FACTORS Gaseous Source Term
- Liquid Source Term **
(No Shielding) (No Shielding) Exposure Integrated Exposure Integrated Time After Rate Exposure Rate Exposure Accident (R/HR) (R) (R/HR) (R) Ihour 1. 0 0 1. 0 0 2 hours .78 .89 .78 .89 3 hours .44 1. 5 .58 1. 5 8 hours .18 3. 0 .32 3.8 16 hours .096 4.1 .19 5.8 1 day .078 4. 8 .16 7.2 2 days .061 6.5 .10 10 3 days .052 7. 8 .077 12 10 days .020 14 .035 22 20 days .0056 17 .017 28 30 days .0016 18 .010 31 3 months 3.4 (-5) 19 .0039 42 1 year 2. 9 (-5) 19 .0012 58
- Gaseous source term is 100% Noble Gases, 25% Halogens
** Liquid source term is 100% Noble Gases, 50T, Halogens,1% Remainder O
Report No. 02-0370-1045 Revision 0 I l l i TABLE NO. 5 , SYSTEMS, SOURCES, AND DILUTIONS l 1 Dilution Volume System Source Dilution (ft.3)
~
Containmant Spray 105% NG Reactor Coolant 89,000 System 50% Hal. System Plus Torus 1% Rem. Liquid Core Spray System 100% NG Reactor Coolant 89,000 50% Hal. System Plus Torus 1% Rem. Liquid Emergency Condanner 100% NG Reactor Coolant 89,000 System 50% Hal. System Plus Torus 1% Rem. Liquid 1 i Clean Up 100% NG Reactor Coolant 7,000 Deminaralizer 50% Hal. System System 1% Rem. Shutdown Cooling 100% NG Reactor Coolant 89,000 System 50% Hal. System Plus Torus ; 1% Rem. Liquid i 1 4 O
j i j . i D N O E3 S$ @ ~ a h % k h! .. m % g S 4 .. g% 3 W m 5 e e $m E w h Q O W k . e ..
$ I ~
Q gnQ n S g$ < g%g - O -- aI 8 itM. m
%J e3
.. g
.. g a
I F f O 8
' 8
" ~8 $E t_.
(do)38/2LVU?dWEL 39V0?AV l
.~
o 2~ A E D O E l U O M O C I U T E O H TA F - 0 0 E S Y 6 . E R E G 'k 0 2 2 N O T D A
- 0 6
C I E E . 7 S E 2 2 H F R 4 I - E B 0 0 2 E E : U 7 5 U E U
/ E 4
)S G Y E M I A U
/
. 0 0 I D
R O G T M. l~ 3 L O
~
CE 0 0 (5 N A2 0 0 1 O TI AI
- 9 0 T I
UI 0 / l 6 A O E R
- 7) B R
E 0 T 6 F A E 0 5 MI T
- 0 4
~ : 0 3
- 0 2
0 1 0 o 0 00 O 0 4 n 0 2 0 7 1 l g eu,q%ti uW y9
O O O FIGURE 3
~
OYSTER C2EEE UUCLEAR -.... - GEUE2ATIUG STATIOAl~ T/ME HISTORYFOR20DE 55 . 400~" MAIUSTEAM c _.
- LIUE BREAE 500 -
N 4 1
- 2002 u,
G g /00~- 7 70 -
. j : : : : : : : :
70 ~ 20 30 40 50 60 70 80 90 l00 200 300 400_. 500 600.__ 4 . i TIME AFTER BREAI/IllITIATION (5ECOUDS)
- - - - - - - - - - - -_.____?'__
O O O FIGU2E 4 OY5TE2 c2EEE MUClEA2 -_- GEUE2ATIUG STATION TIMEHISTORYFORMODE 19 40~~ EME2GEUCY COUDEUSE2 ._ y LlUE BREAK' 300 - D' N M 20 0 - I'd u, D h, too Ct 70 10 20 30 40 50 60 70 60 90 100 200 300 400 000 600
. TIME AFE2 BREAK'IMITIA7702 (6KOUD5) l
6 flGU2E , OYSTE2 C2EEE HUCLEA2 ....... GEUE2ATIUG STAT /OU T/MEHISTORYFORMODE 7
- M- EWERGEUCY COUDEA15E2 ..:
UUE BREAK' l
~
300 - r . l l 200 ^ - ,. ~ G , .l00' )
- n 70 "
. j : . : : : : : :
b : :- : : 1 10 ~ 20 30 40 50 60 70 60 90 10 0 200 300 400 _ 000 600.- l l i . TIME AFTER BREAlllUITIAT/02 (SECOLIDS) _ _ _ - _ _ _ _ _ _ _ _ _ _ r- -e - _ _ _ _ _ _ _ _ _ _ - - . _ _ _ _ _
O o O flGURE 6 OY57E2CEEEE HUCLEA2 . .. GEUE247/UG STAT /0Al TIME ///6TORYFORMODE 16 -
.. 400~- -
~EMERGEUCY COUDEAI5E2 _
LIUE BREAK' k g soo . -
,l zoo x.
G
/00 '
N 70 --
'IO ~ 20 30 40 60 60 70 60 90 /00 200 300 400 000 600
. TIME AFTER BREAI/IUITIATICA) (SECOUDS)
O O O FIGURE 7 OYSTE2 CEEEE HUClEA2 ... ... GEUE2ATIUG STAT /OU i TIME HISTORYFORMODE lb _. . 400~-- .EMERGEUCY COMOEUSEE -
; LIVE B2EAk' k '
300 - n - N - 200 - - R' . m G h, l00 ~ 9 70 - 1 N : : -: :
- 'to ^ 20 30 40 50 60 70 oo .90 too av 300 400 000 600 .
- , TIME AFTER BREAK'IUITIATION (6ECOUDS)
- ---- L) _ O' .
l 8 f/GU2E OYSTE2C2EEE HUClEA2 . . . GEUE247/UG STATIOAl
\
T/ME HISTORYFORMODE /7
.- 4% ~- -
.EME2 GEL /CY COUDEUSE2 .a u
UNE BREAM k I. g s0o . .
- i a
R( i 300 - N .; R . , - M 200' - W . w . G l00 ~ N 70 - - 50 ~ 20 30 40 50 60 70 60 90 10 0 200 300 400 000 600 TIME AFTER BREAllIUITIATION (SECOLIDS)
l l O N s tu W
.a W i k 1 l
l O W Q 1 MJ \ Q \ O
O o !. O FIGURE /3 OY57E2CEEEE HUClEA2 GEUE2ATIUG STATICAI TIME HISTORYFORHODE 30 ) 400=- .Cl6AuuP . L/UE BEE 4!! t : I i 300 - no . . 200 - l00 ~ n 70 - 10 ~ 20 30 40 60 60 70 60 90 100 200 300 400. 500 600 .. TIME AFTER BREAl/IUITIATION (SECOUDS) 9-
, RGU2E 14 OYSTE2CREEE HUCLEA2 . 1 GEUE247/U6 STATIOAl l T/ME HISTORYFORMODE 33 . t 400~- - l .MA/U 6 TEAM .- OUE B2EAW R( - 500 - ' m .i . 200 - 100 ~ 7 70
. j : : : : : : : :
b : : -: : 10 ~ 20 30 40 60 60 70 60 90 NO 200 300 400 %0 600 .
, TIME AFTER BREAl/IlllTIATICAI (5ECOLIDS)
- ~ - - - ~ ~ ~ - ~ ~ ~ - - - - - ~- ~
O O. O
. RGURE I5 OYSTE2CEEEE_ MUClEA2 .
6EUE247/UG STAT /0Al T/ME HISTORYFORMODE 34
- M ~- MAIU 6 TEAM .-
UNE B2EAk' k 500 - 4u 200'- - LJJ g . 100 ~ 7 70 , j : : : : : : : : : : : : 10 ~~ 20 30 40 50 60 70 60 90 l00 200 300 4C0 $00 600 _
, TIME AFTERBREAWIMIilATION (SECOUDS)
~
6
- 2 2' E A D E U O .
O G U T/ O H R H TA F O : x e _ E S Y E G R w
'k E O s
6' __ U C TI_. T 2 P 4E S U R E 2 A 2 H I U B
- x 4
R a 7 E 4 E )5 U 5 U E E N o 9 G Y E R O G M I T L C U sUO i! c w a (K 5 NO mT I A I WIT U I o 'k . e A o ' wB E R R E o T s F A E o s M I T
- o g
- s o
- o g
. :j- o 0
o o o 0 00 0 0 7 o 4- s 2 u 1 i k yg1 4 7
l .\ l{l1 3 4 Q E A D E 2 O O L C 0
/
U T U R 0 H A O 0 T F : 6 K S Y 7 1 E E G E U R O W A 0 0 0 T p E C I T S 2 E 2 A /
// U 8 :
0 0 2 E 2 7 E U 4
)S U E A E D G 5 Y
U E M E L U : 0 0 I L I F O G T/ C U 3 C O E
'l . i v (5 a I A
- b. o TC o
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- o s
T I MI
- 0 6
W AE O : 10 R B R E X : 0 6 A T F E 0 5 MI T
- 0 4
- 0 _
3 _ 0
- 2
. :_0 *1 F - -
. i 0 0 0 00, 0
4 0 5 0 2 0 7 1 , O
-I Rt ,O' ,Su Lo te h 7
O O - O FIGU2E la OYSTE2C2EEE HUCLEA2 . __ GEUE2ATIUG STAT /0Al l TIME HISTORYFORMODE 32
-. 400~- -
.MA/Al 6 TEAM . . .
U UE BEEAM R( t 300 - s N W 20 0'- -
% ~
O 4 g /00 ~ . n 70 I 1 5 5 5 5 % y
~IO ~ 20 30 40 50 60 70 60 90 10 0 200 300 400 500 600 .
, TIME AFTER BREAl/IUITIATIOM (6ECOUDS) 4 0
c- - _ . . . .- - - .. _ . . . _ . . . _ . _ . - . . _ _ _ - . f 8 m 3
.'el ; ??
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Report No. 02-0370-1045 Revision 0 APPENDIX A EDS FLOW Computer EDS FLOW is a modified version of the RELAP4/ MOD 5 Program Abstract computer code developed at the Idaho National Engineering Laboratory. R analyzes the thermal-hydraulic behavior of light water reactor systems sub-subjected to postulated transients such as those resulting from loss of coolant, pump failure, or nuclear power excursions ~. EDS FLOW considers a thermal and hydraulic system as a series of interconnecting user-defined fluid or control volumes. The program solves the mass and energy balances for volumes assumed to contain one-dimensional homogeneous fluid (water and steam) with Q the vapor and liquid phases in thermodynnmf e equilibrium. The momentum transport equation is solved at the inter-faces or junctions between the control volumes. The code requires specific input in order to solve the conservation equations for both the modeled volume contents and the connecting junctions. Additional input is required to describe component models which affect the mass, momentum and energy bninnnes. The main assumptions in EDS FLOW are: One-dimensional fluid and heat conduction equations. Homogeneous fluid equations with the phases in thermodynamic equilibrium. Steady-state empirical correlations to estimate heat transfer coefficients, critical heat fluxes, two-phase friction factors, cnd critical mass fluxes. EDS FLOW is a proprietary program which is mainwned on the CDC 7600 series computer and is controlled by EDS Nuclear.
O APPENDDC B ' RECOMMENDATIONS FOR CONSIDERATION
- BY BWR OWNERS CROUP DEVELOPED BY SHIELDING SUBGROUP ON APRIL 16,1980 - SAN JOSE. CALIFORNIA In order to present a unified position for the development of radiological source terms to be used in response to NUREG-0578 Section 2.1.6 (b) and I&E Bulletin 79-01B, the subgroup recommends that the following ====ntions be used by all BWR plant owners:
I. Time Dependent Releases of Core Fission Pro & cts
- a. The NRC presently requires that core fission products are instanta-eausly released from the core and available for leakage from the containment. For purposes of this analysis (NUREG-0578 and I/E 79-01B), a time dependent release can be assumed. For example, O the core degradation curves shown in Figure A-7 of the Rasmussen Report Appendix A (Section VIII) could be utilized. Alternatively, it 4 is reasonable to assume that reactor fission pro & cts are in$ dan +a-eously released from the core one hour after reactor scram and that these fission pro & cts are simnitaneously available for recircu-lation through Engineered Safety Systems. This latter assumption should be primarily used for determining integrated exposures to vital equipment.
IL Fraction of Core Fission Products Released to Containment 4 For near term submittals in response to NUREG-0578 and I/E 79-013, it is appropriate to use Reg. Guide 1.3/1.7 release frac . tions (i.e.100% noble gases, 50% halogens and 1% others). However, the Owners Group should investigate the behavior of the volatile solids (expecially cesium, rubidium, barium and strontium) at TMI-2, and evaluate the overall appropriateness of the Reg. Guide 1.3/1.7 release fractions in light of these releases. l !O
- - , - . , - . -y
III. Distribution and Behavior of Fission Products Table A provides a set of consistent and reasonable assumptions for the distribution of core fission product releases and could be used in all analyses. These assumptions are supported by the actual distribution of fission products at TMI-2; elimin=+a the obvious double counting inherent.in NRC guidelines; and are supported by mechaniatic accident analysis. , IV. Equipment Leakages Outside Containmant The current NRC position is that for purposes of NUREG-0578 (Section 2.1.6 (b)) analyses no equipment leakage outside containmant need be evala=+ad; This is entirely appropriate when evnin=+ina radiation exposures to vital equipment. However, equipment leakage source terms can potentially affect personnel exposures (both onsite and offsite). The sub-
, group recommends, therefore, that for aspects of the analysis 2
involving personnel exposure, equipment leakage assumptions be made based upon either (1) the results of system leakage tests being performed in accordance with NUREG-0578 (2.1.6. (a)) or, (2) realistic leakages expected (or experienced) during normal system operation. No " passive failures" in O s7 stems are to be assumed. The colldion of equipment leakages in the plant drainage sys-tems should be controlled by administrative procedures to pre-vent overflowing of front-end radwaste system collection tnnks. V. Containment Leakages Several options should be available to the BWR owners when considering this source term.
- a. The " design basis" leakage could be assumed for the duration of the accident (generally taken as 30 days), or;
- b. A containment leakage rate based upon the contninment pressure profile (as evaluated in FSAR Chapter 15) could be utilized.
- The subgroup did not feel that the differences between (a) and (b) would be very significant, especially for short term analysis; however the two (2) opticas should be available.
~ , --y - , y n -
- c. /~ o k T .. EA DISTRIBUTION OF FISSION PRODUCTS TO BE USED IN BWR ANALYSES FOR NUREG-0578 AND I/E BULLETIN 79-01B Torus (or Torus (or Reactor Coolant Reactor Coolant Drywell Air Suppression Pool) Suppression Pool) System Steam System Water Space Air Space Watar Volume Space Volume Class of Fraction (l) Dilution Fraction Dilution Fraction (1 Dilution Fraction Dilution Fraction Dilution Fission Prod. Volumo(2) (1) Volume (2) Volume (2) (1) Volume (2 I Volume (2)
~ - ~
Noble Gases 100% Drywell' 100% Drywell 0-2%(5) Torus 100% Sormal 0-2%(5) Torus or
~
Air Air Water R. C. S. suppress Plus pool Halogens 25%(3) Torus 0% Phis 50% Plus 25% Vapor 50% Water plus
& 0%I4) Air Reactor Space Others 0% 0% Torus 1% Coolant 0% 1% R.C. S. Water~
System Air _.W ater _
- 1. Expressed in percentage (%) of total core inventory for that class of fission product (F. P.).
- 2. Represents the total volume in which the fraction of core F. P. 's is assumed to be homogeneously mixed.
- 3. 25% should be used for short term effects (approximately 1-2 days).
- 4. 0% can be used for long-term analysuc based on TAII-2 experience.
- 5. Use 0% when calculating direct shine doses from torus water source. When calculating the radiation effects of torus (or suppression pool) leakage, the exact percentage of nobic gases dissolved in the water should bo derived based on noble gas solubility and vapor pressures. This percentage should not exceed 2% in most cases.
6 Contamination extends up to AI.S. L. isolation valves and RCICS steam lines. This radiation source is rapidly depleted once RCICS is started. Thus, fractions shown are for start of event only.
- 7. This dilution volumo (Torus and R.C.S.) should be used for all water systems connected to R.C.S. and which recirculate or process contaminated water.
\
O CHAPTER 4 OYSTER CREEK NUCLEAR GENERATING STATION ENVIRONMENTAL QUALIFICATION OF ELECTRICAL EQUIPMENT l Reactor Building Flood Level Analysis 4 O W 4 O
eldis@ nuclear
- EDS Nuclear Inc.
445 Broad Hollow Road Melville, New York 11747 (516)454-0200 October 24, 1980 l 0370-024-NY-013 B/P No. 37-66 Jersey Central Power 6 Light Company 161 Madison Avenue Marristown, New Jersey 07960 ATTENTION: Mr. Y. Nagai
SUBJECT:
Oyster Creek Nucicar Generating Station Reactor Building Flood Level Analysis
REFERENCES:
- 1. EDS Letter No. 0370-024-NY-011, Oyster Creek Nuclear Generating Station Flood Level Analysis, dated 10/14/80.
- 2. EDS Memorandum File No. 0370-024-831, Preliminary Results for Oyster Creek Nuclear 77 Generating Station Reactor Building Flood Level Analysis, dated 10/16/80.
- 3. EDS Report No. 02-0370-1045, Rev. 0 " Environmental Effects in Safety Grade Electrical Equipment Due to High Energy Line Rupture", dated April 1980.
Gentlemen: At the request of Jersey Central Power and Light Company (JCPSL), EDS performed an assessment of the Reactor Building Flood Level following postulated high energy line breaks (HELB). The scope of work for this effort is described in Reference 1. EDS telecopied preliminary unverified results to JCPSL on October 16, 1980 via Reference 2. This letter presents the final checked results of the flooding analysis. The final results included with this transmittal are basically the same as the preliminary results with the following exceptions: For the cleanup system line break, the depth of the S1'-3" elevation increased from 2 " to 2-7/16". L ,' San Francisco
- New York Atlanta
- Pans
eldis$ nuclear Il
'~'
Jersey Central Power 4 Light Company October 24, 1980 0370-024-NY-013 B/P No. 37-66 Page Two l l l The depth of the torus compartment and two east corner rooms l increase from 2-3/16" to 2-k" for the cleanup system line i break. The complete results are presented below. Oyster Creek Nuclear Generating Station Reactor Building Flood Level Analysis EDS has performed a final study evalut. ting the reactor building flood levels for the Oyster Creek Nuclear Generating Station following a High Energy Line Break (HELB) outside containment. The HELB's con-sidered for this study are those defined in the reference (3) report and include line breaks in the main steam system, emergency condenser system, and the cleanup demineralizer system (letdown portion). (O j Main Steam Line Break This break is postulated to occur in the steam tunnel at the 23'-6" elevation. Due to the pressure increase in the tunnel after a steaa line break, the blowout panel opens to the Turbine Building. Any water forming due to condensation in the steam tunnel will drain to the Turbine Building and will collect near sump pump 1-2 at the -8' elevation. This sump pump is located near the Auxiliary Flash Tank 1-3. Emergency Condenser System Line Break This break occurs on the 75'-3" elevation. The steam released from this break is assumed to be contained in the reactor building and is distributed on elevations 75'-3" and 95'-3". The calculated water depths are 1-3/8" and 9/16" respectively. Cleanup System Line Break This is the most limiting break in the reactor building in terms of flooding. A break in the letdown protion of the cleanup system releases more water than a break in the emergency condenser system. 'u
i eldis() nuclear l 1 m l j Jersey Central Power 6 Light Company October 24, 1980 0370-024-NY-013 B/P No. 37-66 Page Three The water released from this break initially flashed tc steam. It is then assumed to condense and collect on the 51'-3" and 75'-3" elevations. This results in depths of 2-7/16" and 13/16" on these elevations, respectively. It is then assumed all the water from this break drains via open stairwells, penetrations, and open hatches in the floor to the 23'-6" elevation, yielding a depth of 2-7/8". All the water then drains to the torus compartment and the two east corner rooms, re- . sulting in a 2 " depth of water at the -19'-6" elevation. Assumptions This study was approached in a conservat'c;' manner, assuming no steam escapes the building, when actual: 7 hiowout panels rupture on the 119'-3" elevation relieving steam to the outside atmosphere. The normal drainage system and the sump pumps were not considered to function for this study. The door to the Shutdown Cooling Heat (a_) Exchanger Room on the 51'-3" elevation is assumed shut preventing a water buildup in this compartment. The results presented here should be considered final. A check of this study has been performed to complete EDS's Quality Assurance requirements. If you have any further questions or comments, please do not hesitate to contact the undersigned. Very truly yours, ffk/ Yb~/ Michael F. Ballard Lead Senior Engineer MFB/cjc p ( /
O CHAPTER S OYSTEE CREEK NUCLEAR GENERATING STATION ENVIRONMENTAL QUALIFICATION OF ELECTRICAL EQUIPMENT System /Equipnent List i O l l l l 0
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\.)
i
OYSTER CREEK NUCLEAR GEN. STATION CLASS IE EOUIPMENT LIST System: Isolation Condenser System Location Item Equip. Inside Outside No. ID. No. Generic Name Pri. Cont. Pri. Cont. 44 V-14-30,31,32,33,34 , Valve Motor Operator X 35 21 IB-05 A1, A2 DP Switch X 21 IB-05 B1, B2 DP Switch X 21 1B-11 Al, A2 DP Switch X 21 IB-ll B1, B2 DP Switch X 8 IG-06 A, B Level Transmitter X 45 V-11-34,36 Air Operator X () 23 RE-15 A thru D Pressure Switch X 26 RE-02 A thru D Reactor Wa ter Level X Switch 13 IB-06 E thru H Area Temp. Dectector X i o
OYSTER CREEK NUCLEAR GEN. STATION O V CLASS IE EOUIPMENT LIST System: Containment Spray System Location Item Equip. Inside Outside No. ID. No. Generic Name Pri. Cont. Pri. Cont. 1 33 PM-51 A thru D Pump Motor X 3&5 V-21-1,3,5,7,9,11, Valve Motor Operator. X 13,17 14 IP-03 A, B Flow Transmitter X 15 IP-15 A thru D Pressure Switch X 26 RE-02 A thru D Level Switch X 34 IP-05 A thru D D.P. Transmitter X O i I !. O
l ! t l l OYSTER CREEK NUCLEAR GEN. STATION l CLASS IE EOUIPMENT LIST i System: Automatic Depressurization System Location Item Equip. Inside Outside No. ID. No. Generic Name Pri. Cont. Pri. Cont. 18 RV-46 A thru D Pressure Switch X 1 IA-83 A thru E Controller X i O I ) 'h l l 4 O l
l l OYSTER CREEK NUCLEAR GEN. STATION CLASS IE EQUIPMENT LIST System: Drywell Pressure Switches & Indications Location Item . Equip. Inside Outside No. ID. No. Generic Name Pri. Cont. Pri. Ccat. 18 RV-46 A thru D Pressure Switch X 6 RE-04 A thru D Pressure Switch X 15 IP-15 A thru D Pressure Switch X 16 IP-07 Pressure Transmitter X 41 PS-153 Pressure Switch X O O
i OYSTER CREEK NUCLEAR GEN. STATION 01
\s / CLASS IE EQUIPMENT LIST System: Reactor Water Level Location Item Equip. Inside Outside No. ID. No. Generic Name Pri. Cont. Pri. Cont.
26 RE-02 A thru D Level Switches X l l 46 RE-18 A thru D Level Switches X l 25 RE-05 A, B Level Switches (Scram) X 46 RE-05/19 A, B Level Switches (Scram) X 10 IA 12 A B Level Transmitter X 10 ID 13 A, B Level Transmitter X-46 RE 05/19 A, B Level Transmitter X O t_.-
'I i
OYSTER CREEK NUCLEAR GEN. STATION (D
. (,./ CLASS IE EQUIPMENT LIST System: Reactor Vessel Pressure Location Item Equip. Inside Outside No. ID. No. Generic Name Pri. Cont. Pri. Cont.
23 RE-15 A thru D Pressure Switch X 22 RE-17 A thru D Pressure Switch X 24 RE-03 A thru D Pressure Switch X (Scram) 7 ID46 A, 3 Pressure Transmitter X 7 ID45 A, B Pressure Transmitter X d (J i
A
-g OYSTER CREEK NUCLEAR GEN. STATION \/ CLASS IE EQUIPMENT LIST Sys ter.: Reactor Isolation Location Item Equip. Inside Outside No. 1D. No. Generic Name Pri. Cont. Pri. Cont.
26 RE-02 A thru D Reactor Water Level X 19 RE-23 A thru D MSL Low Pressure X Sensors 20 RE-22 A thru H MSL High Elev. X 31 IB10 A thru P MSL High Temperature X 35 NSO4 A-Li thru L3 MSlV Solenoids X 35 NSO4 B-L1 thru L3 MSlV Solenoids X 35 NSO4 A 1 & 2 MSiv Valve Position X Switches (~ } 35 NSO4 B 1 & 2 MS1V Valve Position X Switches i O
'%)
i 1 l i j l OYSTER CREEK NUCLEAR CEN. STATION
- CLASS IE EQUIPMENT LIST System: Drywell Isolation Valves Location Item Equip. Inside Outside No. ID. No. Generic Name Pri. Cont. Pri. Cont.
4 V-5-147, 167 Closed Cooling X System 36 V-16-2,14,61 Clean Up System X 38 V-17-1,2,3,55,56, Shutdown Cooling X 57 System 39 V-22-1.2,28,29 Drywell Sump Discharg i X System (A0 & Solenoid ) l 27,28& V-23-13,14,15,16,11 , Drywell & Torus X 30 18,19,20,21,22 Purge & Vent System (A0 & Solenoid)
/"' (15 is A0)(17,18 are MOV) 37 V-24-30 Reactor Water Sample X System 2 V-26-16, 18 Reactor Building to X Torus Vacuum Breaker System (A0& Solenoid) 27 V-27-3, 4 Drywell Purge & Vent
, (A0 & Solenoid) X 30 V-28-17, 18, 47 Torus Purge & Vent X (A0 & Solenoid) (47 is MOV) 29 V-38-9, 10, 22, 23 Drywell & Torus oxygen X Sample System (Solen-oid) 29 V-38-16, 17 Containment Particulate X l Monitor ' 49 V-31-2 Cooling System X , Isolation Valve
--, , f v .- % - - - e
4 [ OYSTER CREEK .I) CLEAR GEN. STATION 2 CLASS IE EOUIPMENT LIST - O Location I Item Equip. Inside Outside No. ID. No. Generic Name Pri. Cont. Pri. Cont. NS03A MSlV Solenoid Valve' X .
! NS03B MS1V Solenoid. Valve X y, ,9 Reactor Water Sample X
Valve { V-14-36 Emergency Condenser X, Isolation Valve - 1 V-14-37 Emergency Condenser X i Isolation Valve V-17-19 Shutdown Cooling X
- Isolation Valve ,
j V-17-54 Shutdown Cooling X Isolation Valve [) V-16-1 Cleanup System X Isolation Valve - l V-1-106 Main Steam Line. Bypass X valve 4-V-1-107 Main Steam Line X Bypass Valve K-10, X-13, ~ X-18 Cannister Type X Electrical Penetration Connector Electrical Connectors X i Terminal Block Terminal Block X Cables Electrical Cables X NR108A Electromatic Relief Valve Operator X NR108B Electromatic Relief X Valve Operator (~' NR108C Electromatic Relief X Valve Operator , i
.-l. -
- . - - _ . . . _ . , . J,..- - , _ . _ _ . , . _ . . , , , , , . . _ . . . . . . - . . . - - - - . , . - - , _ , . , . . . _ , . _ . . , . ,
. . ., . . . ~ . - . _ , . -
4 OYSTER CREEK NUCLEAR GEN. STATION CLASS IE EOUIPMENT LIST Location Item Equip. Inside Outside No. ID. No. Generic Name Pri. Cont. Pri. Cont. NR108D Electromatic Relief X Valve Operator NR108E Electromatic Relief X Valve Operator Splice Splice I X i O 4 1 4 1 t h 1 i i 1 i l 4 O
.. . _ --_ _ - _ . =. ._ _
I j
+ CHAPTER 6 OYSTER CREEK NUCLEAR GENERATING STATION ENVIRONMENTAL QUALIFICATION OF ELECTRICAL EQUIPMENT System Component Evaluation Work Sheet O
1 4 A 1 O
cs F:cility: Oyster Creek Sheet 1 o f . _ __._ unit: System Component Evaluation Work Sheet Environment Documentation Ref
- Equipment Descript,on i Qualification Outstanding Parameter Specification Method items Qualification Specification Quatihcation System: Automatic Operating See Chpt. 7 See Note A See Note C See Note B Depressurization Sys+em Time 3 hours Plant ID No: IA-83-A Temperature 205 See Note A g hhB h @. 7 Component: ADS Pressure I fl
%g Switch Manufacture: Dresser Pnassure 16 See Note A 1 See Note B See Chpt. 7 (I SIA)
Item 1 Model Number: 1593 VX llelative See Chpt. 7 Huinidity 100 See Note A 1 See Note B Function: (%) Item 1 Chemical Not Not - - - - - - Not --- Spray Applicable Required Required Spec: Radiation 3.7x10 0R See Note A 1 See Note B See Chpt. 7 Item 1 Service: 114 x East Drywell Wal l Aging 40 years See Note A 1 See Note B l.ocation: El. 55 Item 1 Flood Level: Not Not Above Flood Level: Yes 3ubmergence Negligible Required 2 Required Analysis Documentatiun Heferences: Notes: A. Not available at this time.
- 1. EDS Nuclear Report #02-0370-1045 B. This equipment will either be replaced
- 2. EDS Report Ref. File 0370-024-831 Dated 10/16/80 or qualifled by July 1, 1982.
C. Tech. Spec. 3.3 Annnn441
O O O Facility: Oyster Creek Sheet 2 o g _,__ unii: System Component Evaluation Work Sheet Environment Documentation Ref. Qualification Equipment Descript. ion outstandind Parameter Specification Memod umns Qualification Specification Qualification System: Automatic Operatin9 3 hour See Note A See Note C See Note B See Chpt. 7 Depressurization Systeni Time Item 1 Plant ID tio: IA-83-B Temperature See Note A See Note B See Chpt. 7 15 1 Component: ADS Pressure ("FI Item 1 Switch Manufacturo: Pressure 16 See Note A 1 See Note B See Chpt. 7 Dresser (PSlA) Item 1 Model tJumber: 1593 VX Relative See Chpt. 7 ilumidity 100 See No*e A 1 See Note B Item i Function: Wal Chemical Not Not - --- Not --- Accuracy Spray Applicable Required Required Spec: Dmno: See Note A See Note B See Chpt. 7 Radiation 1 (6.1x10 4R Item 1 Service: See Chpt. 7 R x East Drywell Aging 40 years See Note A 1 See Note B Location: 3 Item 1 Well E. 72' ,_ , Flood Level: Not Not Above Flood I.evel: Yes .iubmergence Negligible Required 2 Required Analysis Documentation fielerences: tJotes:
- 1. EDS Nuclear Report #02-0370-1045 B. This equipment will either be replaced
- 2. EDS Report Ref. File 0370-024-831 Dated 10/16/80 or qualified by July 1, 1982.
C. Tech. Spec. 3.3 A nt W)O4 41
O O O Fccility: Oyster Creek Sheet 3 og __ unii: System Component Evaluation Work Sheet Environment Documentation Ref. Qualification Equipment Descr. t.ip ion outstandind Parameter Specification Method items Qualification Specification Qualification System: Automatic Operatin0 3 hours See Note A See Note C See Note B See Chpt. 7 Depressurization Time PlanbOkTS: IA-83-C - Temperature See Note A See Note B See Chpt. 7 y Component: ADS Pressure (F) 215 Item 1 Switch Manufacture: Dresser Pressure 16 See Note A 1 See Ncite B See Chpt. 7 (PSIA) Item 1 Model Number: 1593 VX Helative See Chpt. 7 Humidity 100 See Note A 1 See Note B (%) Item 1 Function: Chemical Not Not - - - - - - Not --- Accuracy Spray Applicable Required Required Spec:
""* "" ^ ""
- Hadiation *
(6.1x10 4R Item 1 Service: Sce Chpt. 7 RK-02 A0in9 40 years See Note A 1 See Note B Location: Item 1 Flood Level: Not Not Above Flood Level: Yes .iubmergence Negligible Required 2 Required Analysis Documentation Helerences: Notes:
- 1. EDS Nuclear Report #02-0370-1045 B. This equipment will either be replaced
- 2. EDS Report Ref. File 0370-024-831 Dated 10/16/80 or qualified by July 1, 1982.
C. Tech. Spec. 3.3 Annon4 4 *4
O (3 U (v~'t v Fs:cility: Oyster Creek Sheet 4 o f .___ __ unit: System Component Evaluation Work Sheet Environment Documentation Ref. Qualification outstanding Equipment Descript,on i Parameter Specification Mmod hems Quahfication Specification Qualification System: Automatic Operating See Note A See Note C See Note B See Chpt. 7 Depressurization Syste' ' Tinie 3 hour Item 1 Plant ID No; IA-83-D Temperature 230 See Note A See Note B See Chpt. 7 I Component: ADS Pressure ( "F) E** I Switch Manufacture: Dresser Pressure 16 See Note A 1 See Note B See Chpt. 7 (PSIA) Item 1 Model Number: 1593 VX Relative See Chpt. 7 Hmnidity 100 See Note A 1 See Note B Item 1 Function: (%)
" l Not Not - - - - - -
Not Accuracy Spray Applicable Required Required Spec:
*; See Chpt. 7 Radiation 1.4x104R See Note A 1 See Note B Item 1 Service:
See Chpt. 7 A9in0 40 years See Note A 1 See Note B Location: RK-01 Item 1 Flood Level: Not Not Above Flood Level: Yes submergence Negligible Required 2 Required Analysis Documentation
References:
Notes: A. Not available at this time.
- 1. EDS Nuclear Report #02-0370-1045 B. This equipment will either be replaced
- 2. EDS Report Ref. File 0370-024-831 Dated 10/16/80 or qualified by July 1, 1982.
C. Tech. Spec. 3.3 AOAOO441
-s
% U %)
F:cility: Oyster Creek Sheet 5 ,, _ unit: System Component Evaluation Woric Sheet Environment Documentation Ref. Qualification Equipment Description outstandintf th Iteens Parameter Specification Qualification Specification Qualification Systein: Automatic OperatinD 3 hours See Note A See Note C See Note B See Chpt. 7 Depressurization System Tirne Item 1 Plant ID No; IA-83-E Ternperature , 8"" " ^
'30 l Cornponeot: ADS Pressure ( F) Item 1 Switch Psessure 16 See Chpt. 7 Manufacture: Dresser See Note A 1 See Note B (PSIA)
Itm 1 Model Nurnber: 1593 VX Relative See Chpt. 7 Huinidity 100 See Note A 1 See Note B Item 1 Function: (%) Cheniical Not Not - - - - - - Not ------ Accuracy Spray Applicable Required Required Spec:
" 5 See Chpt. 7 Radiation 3.9x10 R See Note A 1 See Note B Item 1 Service:
See Chpt. 7 Aging 40 years See Note A 1 See Note B Location: RK-03 Item 1 Flood Level: . Not Not Above Flood Level: Yes 3ubmergence Negligible Required 2 Required Analysis Documentation Heferences: Notes: A. Not available at this time.
- 1. EDS Nuclear Report #02-0370-1045 B. This equipment will either be replaced
- 2. EDS Report Ref. File 0370-024-831 Dated 10/16/80 or qualified by July 1, 1982.
C. Tech. Spec. 3.3 AOOOO441
O O O Facility: Oyster Creek Sheet 6 o g __ _, , unit: System Component Evaluation Work Sheet Environment Documentation Ref. Equipment Description cualificat. ion outstanding Parameter Specification cualification Specification Qualification System: y 1 Isolatioi crating 30 days 30 days See Note A 3 analysis Plant ID No: V-26-16 Temperature 77 346 1 3 i test Component: Drywell Vent ( F) 6 Purge Valve Manufacture: ASCO Pressure 110 3 t'est 15 1 (PSIA) Model Number: NP8344A70E Helative liumidity 35 100 1 3 test Function: (%) Chernical not not not Accuracy -- P' pplicable required required Spec: Demo:
" "" 2.0x10 R 8
2x10 R 1 3 test Service:
, Aging 40 years 40 years 1 3 test Location: R xR , r.l. 25 Flood Level:
not not Above Flood Level: Yes Submergence Negligible required 2 required analysis Documentation Relesences: Notes: *
- 1. EDS Nuclear Report #02-0370-1045 A. Oyster Creek FDSAR Amend. 68
- 2. EDS Report Ref. File 0370-024-831
- 3. ASCO Test Report AQS21678/TR Rev. A ASCO Catalog No. NP-1 ASCO letter dated September 26, 1980 AOOOn441
O O O Fecility: Oyster Creek unit: Sheet 7 o f __ __ _ System Component Evaluation Work Sheet Environment Documentation ReI. Equipment Description Qualification outstanding Parameter Specification Qualification Specification Qualification et henis System: y 1 Isolatio n crating 30 days 30 days See Note A 3 analysis Plant ID No; V-26-18
, Temperature 77 346 1 3 test Component: Crywell Vent ( F)
& Purge Valve Manufacture: ASCO Pressure 15 110 1 3 test (PSIA)
Model Number: NP8344A70E flelative Humidity 35 100 1 3 test Function: (%) . Accuracy Chemical not not not - - - - - Spec: pplicable required requ W Demo:
"" "" 5 2.0x10 R 2x10 R 1 3 test Service:
Location: R xR , El. 25, Aging years years test Flood Level: not not Above Flood Level: Yes Submergence Negligible required 2 required analysis Documentation fleferences: Notes:
- 1. EDS Nuclear Report #02-0370-1045 A. Oyster Creek FDSAR Amend. 68
- 2. EDS Report Ref. File 0370-024-831
- 3. ASCO Test Report AQS21678/TR Rev. A ASCO Catalog No. NP-1 ASCO letter dated September 26, 1980
.-~. ..,
O O O facility: Oyster Creek Sheet 8 of unit: System Component Evaluation Work Sheet Environment Documentation Ref. Equipment Description Qualification outstanding Parameter Specification Qualification Specification Method items Qualihcatiori System: Containment Operating Spray System Tirne 30 days 30 days See Note A 4 Test Plant ID No: V-21-5 Teenperature Component: Containment ( F) 77 212 1 4 Test Spray Valve Manufacture: Limitorque Pressure 15 15 1 4 Test see note B (PSIA) Model Number: SMB-00 Helative ilumidity 35 100 1 4 Test Function: (%) Chernical not not not A C*"CY Spray applicable required ----- required Spec: ,
"* 5 Hadiation 5.1x10 R 1x107R 1 3 Test Service:
"9 Location: R*C 3
El. 33' 40 years 40 years 1 5 Test Flood Level: Above Flood Level: Yes Submergence negligible not 2 not analysis required required Documentation
References:
Notes:
- 1. EDS Nuclear Report #02-0370-1045 EDS Report Ref. File 0370-024-831 dated 10/16/80 A*
- 2. g, Reliance motor, EDS R port Temp. Profile:
- 3. Limitorque Test Report #B0003 Limitorque asteric item, insulation class: B letter to CPU dated 3/21/80
- 4. Limitorque Test Report #F-C3271 Ditto.
- 5. Limitorque letter dated:
- Annnn441
A p) e N k
' Fgcility: Oyster Creek Sheet 9 of unit: System Component Evaluation Work Sheet Environment Documentation Ref.
Equipment Description Qualification outstanding Parameter Specification Qualification Specification I "'" Qualification Syst sm: Containment operating Spray System Timo 30 days 30 days See Note A 4 Test Plant ID No:. V-21-11 Temperature Component: Containment ( F) 77 212 1 4 Test Spray Valve Manufacture: Limitorque Pressure 15 15 1 4 Test see note B IPSIAI Model Number: SMB-00 fletative limnidity 35 100 1 4 Test Function: (%) Chemical not not not ACC"'*CY Spray applicable required ----- required - - - - Spec: fladiation 5.1x10 R lx10 7R 1 3 Test Service:
^ "
Location: R x Dryw 11 Wall 40 years 40 years 1 5 Test 5 El. 62' Flood Level: Above Flood Level: Yes Submergence negligible not 2 not analysis required required Documentation
References:
Notes: *
- 1. EDS Nuclear Report #02-0370-1045
- 2. EDS Report Ref. File 0370-024-831 dated 10/16/80 A. Oyster Creek. FDSAR Amend. 68
- 3. Limitorque Test Report #B0003 Limitorque
. anc m r, EDS Report Temp. Profile:
letter to CPU dated 8/21/80 steric item, insulation class: B
- 4. Limitorque Test Report #F-C3271 Ditto.
- 5. Limitorque letter dated:
AOnOO441
( ) [ i L/ ) J O' Fccility: Oyster Creek She:t 10 cf . Unit: System Component Evaluation Work Sheet Environment Documentation Ref. oualification Equipment Description outstanding Parameter Specification 'h"I '** Qualification Specification Qualification Systi.m: Drywell Operating Isciation Valves Time 30 days 30 days See Note A 4 Test Plant ID No: V-5-167 Temperature Component: Reactor Bldg. 1 "F) 77 212 1 4 Test Closed Loop Cooling Syst m Valve Manu acture: Limitorque Pressure 15 15 1 4 Test see note B IPb3^I Model Number: SMB-000 Helative Humidity 35 100 1 4 Test Function: (%) Cliemical not not not ACCURACY Spray required required applicable --
==
Spec: _ _ Demo: Hadiatimt 2.5x10 4R 1x107R 1 3 Test Service: _ R x Drywell Wal 4 Aging Location: El. 49' 40 years 40 years 1 5 Test Flood Level: Above Flood Level: Yes Submergence negligible not 2 not analysis required required Documentation
References:
Notes:
- 1. EDS Nuclear Report #02-0370-1045 A. Oyster Creek FDSAR Amend. 68
- 2. EDS Report Ref. File 0370-024-831 dated 10/16/80 eliance m t r, EDS Report Temp. Profile:
- 3. Limitorque Test Report #B0003 Limitorque asteric item, insulation class: B letter to CPU dated 8/21/80
- 4. Limitorque Test Report #F-C3271 Ditto.
- 5. Limitorque letter dated:
AOOOn443
O O O Facility: Oyster Creek Sheet II of _ unit: System Component Evaluation Work Sheet Environment Documentation Ref. Equipment Description Qualification outstanding Panameter Specification dualification Specification "' ""** Qualification System: Drywell Operating Isolation Vavles Timo 30 days 30 days See. Note A 4 Test Plant ID No:V-5-147 Temperature Compcaent: Reactor Bldg. (m 77 212 1 4 Test Closed Loop Cooling lb5Ndfa*ct reNimitorque Pressure 15 15 1 4 Test see note B (PSIA) Model Number: SMB-000 flelative liumiditY 35 100 1 4 Test function: (%) Chemical not not not ACCURACY Spray applicable required required - - - Spec: fladiation 2.5x10 4R lx10 7R 1 3 Test Service: R4 xDrywell Wall Aging Location: El. 49' 40 years 40 years 1 5 Test Flood Level: Above Flood Level: Yes Submergence negligible not 2 not analysis required required Documentation ficierences: Notes: '
- l. EDS Nuclear Report #02-0370-1045 EDS Report Ref. File 0370-024-831 dated 10/16/80 A. Oyster Creek FDSAR Amend. 68 2.
- 3. Limitorque Test Report #B0003 Limitorque B. Reliance motor, EDS Report Temp. Profile:
asteric item, insulation class: B letter to CPU dated 8/21/80
- 4. Limitorque Test Report #F-C3271 Ditto.
- 5. Limitorque letter dated:
AnOnO441
O O O Facility: Oyster Creek She2t 12 cf _ unit: System Component Evaluation Work Sheet Environment Documentation Ref. Equipment Desen,ption Qualification outstandmo Parameter Specification Method items Qualification Specification Qualification System: Containment Spray operating 4 hours 5 hours See Note B 4 test System Time Plant ID No: V-21-13 Temperature 212 1 4 test Component: Containment ( F) 140 Spray Valve Manufacture: Limitorque Pnassure (PSIA) 15 ' 15 1 4 test See Note A Model Numtier: SMB-000 Helative Huniidity 100 100 1 4 test Function: (%) Cheniical Accuracy Spray not not - - - - not --- - applicable required required Spec:
""' 7 Radiation 6.7x10 5 g lx10 R 1 3 test Service:
Aging 40 years 40 years 1 test Rg-R2 xR C ~D 5 Location: El. ??' Flood Level: not not Above Flood Level: Yes Submergence Hegligible required 2 required Analysis Documentation References; 1. EDS Nuclear Report # 02-0370-1045 Notes: A. Peerless motor, EDS Report Temp. Profile:
- 2. EDS Report Ref. File 0370-024-831 dated 10/16/80 13, insulation class: B
- 3. Limitorque Test Report # B0003 Limitorque B. Tech. Spec. 3.3 letter to CPU dated 8/21/80
- 4. Limitorque Test Report # F-C3271. Ditto.
- 5. Limotorque letter dated:
Annoo441
OC/ C\ J \,_,/ Facility: r, ster Creek Sheet of __ unit: System Component Evaluation Work Sheet Environment Documentation Ref. Equipment Desen.p tion . Qualification outstanding Parameter Specification Qualification Specification Method items Qualification System: Containment Spray Operating System 4 hours 5 hours See Note B 4 test Time Plant ID No: V-21-17 Temperature 212 1 4 test Component: Containment ( F) 77 Spray Valve Manufacture: Limitorque Psessure 15 15 1 4 test (PSIA) See Note A Model Number: SMB-000 Helative ilumidity 100 100 1 4 test Function: (%) Clunnical Accusacy not not -- not -- Spray arplicable required required Spec: Hadiation 3.7x10 R Service: R6-R7 xR 3 -R C 40 years 40 years 1 5 test Aging Location: El. 27' Flood Level: not nt Above Flood Level: Yes Submergence Negligible required 2 required Analysis Documentation
References:
- 1. EDS Nuclear Report. # 02-0370-1045 Notes: A. Peerles motor, EDS Report Temp. Profile:
- 2. EDS Report Ref. File 0370-024-831 dated 10/16/80 none, insulation class: B
- 3. Limitorque Test Report # B0003 Limitorque B. Tech. Spec. 3.3 letter to GPU dated 8/21/80
- 4. Limitorque Test Report # F-C3271 Ditto.
- 5. Linotorque letter dated:
An(W Dn4 41
O O O Facilit y: Oyster Ciaek Sheet I4 of unii: System Component Evaluation Work Sheet Environment Documentation Ref. Equipment Description ou lification outstanding Parameter Specification Qualification Specification Method items Qualification System: Containment Spra> Operating System 4 hours 5 hours See Note B 4 test Tiene Plant ID No; V-21-3 Temperature 212 1 4 test Component: Containment (*F) 165 Spray Valve Manufacture: Limitorque Pressure - 33 15 1 4 test (PSIA) See Note A Model Number: SMB-00 Iklative ilumidity 100 100 1 4 test function: I%)
" I Accuracy not not not Spray Spec: applicable required ,
required 5 Demo '- 6.0x10 R 7 Radiation lx10 R 1 3 test Sarvice: AUinU 40 years 40 years 1 5 test Location: SE Corner Rm. _ E1.- 19 ' Flood Level: not not Above Flood Level: Yes Submergence Negligible required 2 required Analysis Documentation fleferences: 1. EDS Nuclear Report # 02-0370-1045 Notes: A. Reliance motor, EDS Report Temp. Prtfile:
- 2. EDS Report Ref. File 0370-024-831 dated 10/16/80 14, insulation class B
- 3. Limitorque Test Report # B0003 Limitorque B. Tech. Spec. 3.3 letter to CPU dated 8/21/80
- 4. Limitorque Test Report # F-C3271 Ditto.
- 5. Limotorque letter dated:
AfWV14 41
I n /'"%
]
~
racility: Oyster Creek 15 Srini et
- t. .. System Component Evaluation Work Sheet Environment Documentation Ref.
Equipment Descu,pt,oni Qualification outstandine Parameter Specification Method items Qualification Specification Qualification Systern: Containment Spras System P ' li"9 4 hours 5 hours See Note B 4 test Time Plant 10 No: V-21-9 Teniperature 212 1 4 test 165 Component: Containment Iofl Spray Valve -- Manufacture: Limitorque P' essure 15 15 1 4 test (PSIA) See Note A Model Numlier: SMB-00 Helative Humidity 100 100 1 4 test Function: (%I Cheinical Accuracy Spray not not not ----- Spec: applicable required required Demo: Radiation 6.1x10 5R lx10 R 1 3 rest Service: Aging 40 years 40 years 1 5 test NE Corner Ibn. Location: E1.- 19 ' Flood Level: ' not not Above Flood Level: Yes Submergence Negligible required 2 required Analysis Documentation Helerences: 1. EDS N.* clear Report # 02-0370-1045 Notes: A. Reliance motor, EDS Report Temp. Profile:
- 2. EDS Report Ref. File 0370-024-831 dated 10/16/80 18, insulation class: B
- 3. Limitorque Test Report # B0003 Limitorque B. Tech. Spec. 3.3 letter to CPU dated 8/21/80
- 4. Limitorque Test Report # F-C3271. Ditto.
- 5. Limotorque letter dated:
Anf MM4 4 "I
i O W""~' " k O% u/% Frcility: Sheet 16 og unit: System Component Evaluation Work Sheet Environment vu .u...m.m . S P-' ! Equipment Descr,pt,on i i s,.,,, inn
- cuistanding Parameter Specification Qualification Specification Method items oualification i
System: Containment Spra' Operating 4 hou's System 5 hours See Note B 4 test Time Plant ID No: V-21-1 Temperature 165 212 1 4 test Component: Containment I II Spray Valve Manufacture: Limitorque P' essure (PSIAI 15 ' 15 1 4 test See Note A Model Number: SMB-00 Helative ilumidity 100 100 1 4 test Function: I"4I I Accuracy nt not not --- - Spray Spec: applicable required required 6 Demo: 1.0x10 R lx10 R 1 3 test Service: Aging 40 years 40 years 1 5 test L ocation: SE Corner Rm. E1.- 19 ' Flood Level: not 2 not Above Flood Level: Yes Submergence Negligible required required Analysis Docuenentation
References:
- 1. EDS Nuclear Report # 02-0370-1045 Notes: A. Reliance motor, EDS Report Temp. Profile:
- 2. EDS Report Ref. File 0370-024-831 dated 10/16/80 14, insulation class: B
- 3. Limitorque Test Report # B0003 Limitorque B. Tech. Spec. 3.3 letter to CPU dated 8/21/80
- 4. Limitorque Test Report # F-C3271. Ditto.
- 5. Limotorque letter dated:
AnOttn441
r~% C Y u) Facility: Oyster Creek gg, 17 g unii: System Component Evaluation Work cheet Environment Documentation Ref. Equipment Description Qualification outstanding Paranneter Specification * "" Quahlication Spacification Qualification Systern: Containment Operating 4 hours 5 hours See Note B 4 test Spray System lime Plant ID No: V-21-7 T empesature 165 212 1 4 test Component: Containment i F) Spray Valve Manufacture: Limitorque P' essure 15 15 1 4 test (PSIAI See Note A Model Number: SMB-00 lielative ilumidity 100 100 1 4 test Function: (*M Acctuacy not not not --- - Spray Spec: applicable required required Desno: 6.3x10 R Radiation lx10 R 1 3 test 3 Service: 40 years 40 years 1 5 test Location: NE Corner km. AUi"9 E1.-19' Flood Level: not not Above Flood Level: Yes Submergence Negligible required 2 required Analysis Documentation lleferences: 1. EDS Nuclear Report # 02-0370-1045 Notes: A. Reliance motor, EDS Report Temp. Profile:
- 2. EDS Report Ref. File 0370-024-831 dated 10/16/80 18, insulation class: B
- 3. Limitorque Test Report # B0003. Limitorque B. Tech. Spec. 3.3 letter to CPU dated 8/21/80
- 4. Limitorque Test Report # F-C3271 Ditto.
- 5. Liiaotorque letter dated:
AOrW)O 4 il
fm O Q kJ O Frcility: Oyster Cree' Sheet 18 of unit: System Component Evaluation Work Sheet I Environment Documentation Ref. Qualification outstanding Equipment Descn,ption Method items Parameter Specification Qualification Specification Qualification System: Drywell Pressure operating Switches 6 Indicators Time 30 days 30 days See Note A 4 analysis Plant ID No: RE-04-A Temperature 77 77 1 4 analysis Component: ijrywell ( "F) Press. Scram Switch Manufacture: Static-0-Rini P' essure (PSIAl 15 15 1 4 analysis Model Number: 12NKA Helative ilumidity 35 35 1 4 analysis Function: 1%) Chemical not not not Accuracy Spray applicable required required Spec: Hadiation 1.5x10 6R 6 5x10 R 1 3 analysis Service: Heactor Vessel x R6~ 7 Aging *- See Chpt.7 Location: Y*"#8 Y""#8 ** E1. 55' Flood Level: not not Above Flood Level: Yes Submergence negligible required 2 required analysis Documentation
References:
Notes:
- 1. EDS Uuclear Report #02-0370-1045 A. Oyster Creek FDSAR Amerd. 68
- 2. EDS Report Ref. File 0370-024-831 dated 10/16/80
- 3. Gilbert Ascociates, Inc., letter dated 6/4/80
- 4. Wyle Lab. Study -
A0000443
rh f% U O d Frcility: Oyster Creek Sheet _19 of unit: System Component Evaluation Wo.-k Sheet Environment Documentation Ref. Qualification outstanding Equipment Descr,ptioni Method items Parameter Specification Qualification Specification Qualification System: Drywell Pressure Operating Switches & Indicators Time 30 days 30 days See Note A 4 analysis Plant ID No: RE-04-B Temperature 77 77 1 4 analysis i Component: Drywell I FI Press. Scram Switch Manufacture: Static-0-Rin) P' essure (PSIAl 15 15 1 4 analysis Model Number: 12NKA Relative ilumidity 35 35 1 4 analysis Function: (%) Chemical not not not Spray applicab1( required required Spec:
""* 6 Radiation 1.5x10 R 5x10 R 1 3 analysis Service:
Reactor Vessel x Aging . See Chpt.7 l_ocation: R6-R7 El. 55' 40 years 40 years 1 4 Item 6 Flood Level: not not Above Flood Level: Yes Submergence negligible required 2 required analysis Documentation Heferences: Notes:
- 1. EDS Nuclear Report #02-0370-1045 A. Oyster Creek FDSAR Amend. 68
- 2. EDS Report Ref. File 0370-024-831 dated 10/16/80
- 3. Gilbert Associates, Inc., letter dated 6/4/80
- 4. Wyle Lab. Study AOOOo443
O O O Facility: oyster Creelt Gh:et M of unii: System Component Evaluat. ion Work Sheet Environment Documentatiots Rel. Equipment Description Qualification outstanding Parameter Specification Quahlication '"*' 'I""'S Specification Qualification System: Drywell l'ressure operating Switches & Indicators Time 30 days 30 days See Note A 4 analysis Plant ID No; RE-04-C Temperature 77 77 1 4 analysis Component: Drywell ( F) Press. Scram Switch Manufacture: Static-0-Rin; P' essure (PSIA) 15 15 1 4 analysis Model Number: 12t!KA gegag;ve , ilumidity Function: (%) 35 35 1 4 analysis Clieniscal not not not Accuracy Spray applicab1( required - required ---- Spec:
"* 6 Radiation 2.8x10 5R 5x10 R 1 3 analysis Service:
R x North Drywell Wall Aging See Chpt.7 L8tation: El. 55' 40 years 40 years 1 4 L tem 6 Flood Level: not not Above Flood Level: Yes Submergence negligible required 2 required analysis Documentation Heferences: Notes: *
- 1. EDS Nuclear Report #02-0370-1045 A. Oyster Creek FDSAR Amend. 68
- 2. EDS Report Ref. File 0370-024-831 dated 10/16/80
- 3. Gilbert Associates, Inc., letter dated 6/4/80
- 4. Wyle Lab. Study ArWlOO441
O O O Sheet 21 og Facility: Oyster Creek unit: System Component Evaluation Work Sheet Environment Documentation Ref. Qualdicaiion outstanding Equipment Descr,ipt,on i Mediod Itms Parameter Specilication Qualification Specification Qualification F System: Drywell Pressure OperatinD 30 days 30 days See Note A 4 analysis Switches & Indicators Time Plant ID tio: RE-04-D Temperature 77 77 1 4 analysis Component: Drywell I "II Press. 'dcram Switch Manufacture: Static-0-Rin Pressure (PSIA) 15 15 1 4 analysis Model flumber: 12i Relative llumidity gg 35 35 1 4 analysis Function: __ Chemical not not not Accur cy Spray applicabit required - required Spec: Demo: 6 Radiat.iort 2.8 x 10 R 5 5x10 R 1 3 analysis Service: R x North Drywell Aging a See Chpt.7 E I_acation: Wall El. 55' 40 years 40 years 1 4 item 6 Flood Level: not not Above Flood Level: Yes Submergence negligible required 2 required analysis Documentation
References:
Notes:
- 1. EDS Nuclear Report #02-0370 -1045 A. Oyster Creek FDSAR Amend. 68
- 2. EDS Report Ref. File 0370-024-831 dated 10/16/80
- 3. Gilbert Associates, Inc., letter dated 6/4/80
- 4. Vyle I.ab. Stady AOOoo443
Facility: Oyster Creek - Sheet 77, of ; unit: System Component Evaluation Work Sheet , i. Environment Documentation Ref. Equipment Desen,ption Qualification outstanding Parameter. Specification Qualification Specification Method items Qualification System: Operatinq' Reactor Vessel Pressure Time See Note A See Note C See Note B See Chpt.7 30 days Item 7 ' Plant ID No: ! ID 45 A Temperature
- See Chpt.7 See Note A See Note B Component: ( *F) 230 1 Item 7 Reactor Vessel -
Pressure Transmitter Manufacture: G.E./MAC Pressure I See Chpt.7 See Note A See Note B (PSIA) 16 Item 7 Modes Number: Relative 51 Humidity 100 I See Chpt.7 See Note A See Note B Function: (%) Item 7 Chemical Not Not Not Accuracy Applicable Required Required S ray Spec: Radiation 1.4x10 R See Note A 1 See Note B 3ee Chpt. 7 Item 7 Service: Location: . A0i"0 40 years See Note A I See Note B See Chpt.7 It*" El. 72' RK 01 Flood 1.evel: Above Flood Level: Yes Submergence Negligible not required 2 act required analysis Documentation fleferences: 1. EDS Nuclear Report #02-0370-1045 Notes: A. Not available at this time.
- 2. EDS Report Ref. File 0370-02*e-831 B. This equipment will either be replaced or Dated 10/16/80 qualified by 7/1/82.
- 3. G.E. Letter and Test Report #C-EN-0-163 C. Oyster Creek FDSAR Amend. 68 Dated 10/16/80.
Annnn441
3 O f3 (dOyster Creek rb V { Frcility: Sheet 23 of i unii: System Component Evaluation Work Sheet ' i Equipment Description "" " " Oualification outstanding Parameter ^ Specification Qualification Specification Qualification System: Operating Reactor Vessel Pressure Time 30 days See Note A See Note C See Note B See Chpt.7 Item 7 Plant ID No: ID 45 B 215 Temperature See Note A See Note B See Chpt.7 Component: ( "F) 1 Item 7 Paactor Vessel Pressure Transmitter 16 Manufactme: (; . E . /MAC W' e I See Chpt.7 See Note A See Note B Item 7 Model Number: fletative 551 ilurnidity 100 I See Chpt.7 See Note A See Note B Function: (%) Item 7 Chemical Not Not Not Accuracy A P plicable Required Required Spray Spec: fladiation 1.4x10 0R See Note A 1 See Note B 3ee Chpt. 7 Item 7 Servico: AUinD 40 years See Note A 1 See Note B See Chpt.7 Location: EL.72 ' RK-02 Item 7 Flood Level: Above Flood Level: Yes Submergence Negligible nbt required 2 .m t required analysis Documentation fleferences: 1. EDS Nuclear Report #02-0370-1045 fJotes: A. Not avatlable at this time.
- 2. EDS Report Ref. File 0370-024-831 B. This equipment will either be replaced or Dated 10/16/80 qualified by 7/1/82.
- 3. G.E. Letter and Test Report #G-EN-0-163 C. Oyster Creek FDSAR Amend. 68 Dated 10/16/80.
( Annnn44*4 5
O O O Fccility: Oyster Creek Sheet 24 of, unit: System Component Evaluation Work Sheet Environment Documentation Ref. Qualification Outstanding Equipment Descr,pt, i ion Method items Pararneter Specification Qualification Specification Qualification System: Reactor Vessel Opesatin0 See Chpt. 7 See Note A See Note C See Note B Pressure Time 30 days Item 7 Plant ID No: ID46A Temperature See Note A See Note B See Chpt. 7 g Component: Reactor ( *F) 230 Item 7 Vessel, Pressure Transmitter ressm e 16 See Note A See Note B See Chpt. 7 Manufacture: GE/MAC 1 (PSIA) Item 7 Model Number: Catalog # Relative VI,F g438 See Chpt. 7 ilumidity 100 See Note A 1 See Note B (%) Ite" 7 Function: Chemical Not Not - - - - - - Not - - - Accuracy Spray Applicable Required Required Spec:
'" See Chpt. 7 Radiation 1.4x10 R See Note A 1 See Note B Item 7 Service:
See Chpt. ? Aging 40 years See Note A 1 See Note B Location: El. 72' RK-01 Item 7 Flood Level: Not Not Above Flood Level: Yes 3ubmergence Negligible Required , 2 Required Analysis Documentation
References:
Notes: A. Not available at this time.
- 1. EDS Nuclear Report #02-0370-1045 B. This equipment will either be replaced
- 2. EDS Report Ref. File 0370-024-831 Dated 10/16/80 or qualified by July 1, 1982.
C. Oyster Creek FDSAR Amend. 68 A rWin4 4 *4
l's - V' Facility: Oyster Creek Sheet _ 2 5, _ o f . _,_ __ unit: System Component Evaluation Work Sheet Environment Documentation Ref. Qualification outstanding Equipment Description Specification * ** Pasameter Qualification Specification Qualification System: Reactor Vessel Operating See Note A See Note C See Note B See Chpt. 7 Pressure Time 30 days Item 7 Plant ID No: ID46B Temperature See Note A See Note B See Chpt. 7 215 I Item 7 Component: Reactor g p3 Vessel Pressure Manufacture: Transmitter Pressure 16 See Note A 1 See Note B See Chpt. 7 (PSIA) Item 7 GE/MAC Model Number: Catalog flelative See Chpt. 7 Number VPF 1438 Humidity 100 See Note A 1 See Note B Function: N "" Chemical Not Not ------- Not --- Accuracy Spray Applicable Required Required Spec:
"'" See Chpt. 7 Radiation 1.4x10 R See Note A 1 See Note B IE**
Service: See Chpt. 7 AginD 40 years See Note A 1 See Note B Location: El. 72' RK-02 Item 7 Flood Level: Not Not Above Flood Level: Yes Submergence Negligible Required 2 Required Analysis Documentation
References:
Notes:
- 1. EDS Nuclear Report #02-0370-1045 B. This equipment will either be replaced ,
EDS Report Ref. File 0370-024-831 Dated 10/16/80
- 2. or qualified by July 1, 1982.
C. Oyster Creek FDSAR Amend. 68 aannn m
/ f'\ -
(3) V,O U I Frcility: Oyster Creek Sheet 26 of unit: System Component Evaluation Work Sheet i i l Environment Documentation Ref. i Equipment Description Qualification outstanding Parameter' Specification Qualification Specification Qualification System: Operating Isolation Condenser 4 hrs. See Note A See Note C See Note B See Chpt.7 Time INNt 15No: 10 06 A-1 Temperature 270 See Note A See Note B See Chpt.7 l Component: ("F) 1 Item 8 Level Transmitter Manufacture: G.E./MAC Pressure I See Chpt.7 16 See Note A See Note B (PSIA) Item 8 Model Number: 553 Relative Humidity 100 I See Chpt.7 See Note A See Note 5 Function: (%; Item 8 Chemical Not Not Not Accuracy 3,y Applicable Required Required Spec:
"* See Note A See Note B 'see Chpt. 7 Radiation 5.3x10 R 1 Item 8 Servico:
Reactor Bldg. Agin0 40 years See Note A 1 See Note B See Chpt.7 Location: El. 98' Item 8 Flood Level: Above Flood Level: Yes Submergence Negligible not required 2 oot required analysis Documentation
References:
- 1. EDS Nuclear Report #02-0370-1045 Notes: A. Not available at this time.
- 2. EDS Report Ref. File 0370-024-831 B. This equipment will either be replaced or Dated 10/16/80 qualified by 7/1/82.
- 3. G.E. Letter and Test Report #G-EN-0-163 C. Tech. Spec. 3.3 Dated 10/16/80.
AGOOO441
Facility: Oyster Creek - Sheet 27 of unit: System Component Evaluation Work Sheet ! I
" " " " "N Equipment Description -
Qualification Outstanding Parameter' Specification Qualification Specification Qualification System: Isolation 4 hrs. Operating Condenser System Time See Note A See Note C See Note B See Chpt.7 Plant ID No: Ite.a 8 IC-06-A-2 Temperature 270 See Note A See Note B See Chpt.7 Component: ( F) 1 Item 8 Level Transmitter Manufacture: G.E./MAC Pressure I See Chpt.7 16 See Note A See Note B (PSIA) Item 8 Model Number: 553 Helative ilumidity 100 See Note A' See Note B See Chpt.7 Function: (%) Item 8 Chemical Not Not -- Not Accuracy Applicable Required Required Spray Spec: Radiation 5.3x10 R See Note A 1 See Note B ice Chpt. 7 Service: Item 8 Location: Reactor Bldg. A9inD 40 years See Note A 1 See Note B See Chpt.7 E1. 98 Item 8 Flood Level: Above Flood Level: Yes Submergence Negligible not required 2 act required analysis Documentation lleferences: 1. EDS Nuclear Report #02-0370-1045 N. tes: A. Not available at this time.
- 2. EDS Report Ref. File 0370-024-831 B. This equipment will either be replaced or Dated 10/16/80 qualified by 7/1/82.
- 3. G.E. Letter and Test Report #G-EN-0-163 C. Tech. Spec. 3.3 Dated 10/16/80.
Aoooo443
O O - O Fecility: Oyster Creek
- Sheet 20 of unii: System Component Evaluation Work Sheet :
Environment Documentation Ref. Equipment Desen.ption Qualification outstanding Parameter' Specification Method hems cualification Specification cualification System: Isolation operating Condenser System 4 hrs. See Note A See Note C See Note B See Chpt.7 Time , Item 8 Plant ID No: IG-06-B-1 Temperature See Chpt.7 270 See Note A See Note B Component: I FI 1 Item 8 Level Transmitter Manufacture: C.E./MAC Pressure I See Chpt.7 16 See Note A See Note B (PSIA) Item 8 Model Number: 553 fletative Humidity 100 See Chpt.7 See Note A See Note B Function: (%) Item 8 Chemical Not Not Not Accuracy Spray Applicable Required Required Spec: _ Denm: 5.3x10 R5 11adiation See Note A 1 See Note B See Chpt. 7 Item 8 Service: Location: Reactor Bldg. AUi"9 40 years See Note A 1 See Note B See Chpt.7 E1. 98' Item 8 Flood Level: Above Flood Level: Yes Submergence Negligible not required 2 act required analysis Documentation fleferences: 1. EDS Nuclear Report #02-0370-1045 Notes: A. Not available at this time.
- 2. EDS Report Ref. File 0370-024-831 B. This equipment will a. her be replaced or Dated 10/16/80 qualified by 7/1/82.
- 3. G.E. Letter and Test Report #G-EN-0-163 C. Tech. Spec. 3.3 Dated 10/16/80.
ADOOO441
O Oe- u pJ . Facility: Oyster Creek ~ unit: Sheet 29 of System Component Evaluation Work Sheet Environment Documentation Ref. Equipment Description Qualification outstanding Parameter. Specification Qualification Specification Method items Qualification System: Isolation Operating Condenser System Time 4 hrs. See Note A See Note C See Note B See Chpt.7 Item 8 Plant ID No: IC-06-B-2 Temperature See Chpt.7 See Note B 270 See Note A Component: ( F) 1 Item 8 Level Transmitter Manufacture: G.E./MAC Pressur I See Chpt.7 (PSIA) 16 See Note A See Note B Item 8 Model Nuniher: 553 Helative flumidity 100 See Note A See Note B See Chpt.'/ Function: (%) Item 8 Chemical Not Not Not Accuracy Applicable Required Required S ray Spec: Demo'- Radiation See Note A 1 See Note B ice Chpt. 7 5.3x10 5R Item 8 Service: Reactor Bldg. Agi"U 40 years See Note A 1 See Note B See Chpt.7 Location: El. 98' Item 8 Flood Level: Above Flood Level: Yes Submergence Negligible not required 2 act required analysis Documentation
References:
- 1. EDS Nuclear Report #02-0370-1045 Notes: A. Not available at this time
- 2. EDS Report Ref. File 0370-024-831 H. This equirment will either be replaced or Dated 10/16/80 qualified by 7/1/82.
- 3. G.E. Letter and Test Report #C-EN-0-163 C. Tech. Spec. 3.3 Dated 10/16/80.
. 900n4 4'l
O O O Facility: Oyster Creek og __ Sheet 30 unit: System Component Evaluation Work Sheet Environment Documentation Ref. Equipment Description Qualdication outstanding Parameter Specification Qualification Method items Specification Qualification System: Reactor Water Operating See Note A See Note C See Note B See Chpt. 7 Level Time 30 days Item 10 Plant ID No; ID-13A Temperature 230 See Note A g he Note B See M . 7 Component: Transmitter ( F) Item 10 Manufacture: P SS 16 See Note A 1 See Note B See Chpt. 7 - CE (PSIA) Item 10 Model Number: 553 Relative See Chpt. 7 Humidity 100 See Note A 1 See Note B Item 10 Function: (%) Chemical Not Not - - - - - - Not Accuracy -- Spray Applicable Required Required Spec:
"" See Chpt. 7 Radiation 1.4x10 R See Note A 1 See Note B Item 10 Service:
See Chpt. 7 Containment Aging 40 years See Note A 1 See Note B l_ocation:Bld g. El. 72' Item 10 Flood Level: Not Not Above Flood 1evel: Yes submergence Negligible Required- 2 Required Analysis Documentation
References:
Notes: A. Not availabic at this time.
- 1. EDS Nuclear Report #02-0370-1045 B. This equipment will either be replaced
- 2. EDS Report Ref. File 0370-024-831 Dated 10/16/80 or qualified by July 1, 1982.
C. Oyster Creek FDSAR Amend. 68 A (nOfW)4 4 "I
r rg (]J V(3, V Facility: Oyster Creek Sheet 31 of___ unii: System Component Evaluation Work Sheet Environment Documentation Ref. Qualification outstanding Equipment Descript. ion Method items Parameter Specification Qualification Specification Qualification System: Reactor Water crating 30 days See Note A See Note C See Note B See Chpt. 7 Item 10 Plant ID No: ID-13B - Temperature See Note A See Note B See Chpt. 7 I Com- onent: Transmitter (T) 215 Item 10 Manufacture; cg Pressure 16 See Note A 1 See Note B See Chpt. 7 (PSIA) Item 10 Model Number: 553 flelative See Chpt. 7 ilumidity 100 See Note A 1 See Note B Function: (%) Itera 10 Chemical Not Not ------- Not Accuracy Spray Applicable Required Required Spec: Derno: See Note A See Note B See Chpt. 7 Radiation 46.1x10 R 1 Item 10 Service: See Chpt. 7 Containment Bldg. Aging 40 years See Note A 1 See Note B t.ocation: E1. 72' Item 10 Flood I.evel: Not Not Above Flood Level: Yes Submergence Negligible Required 2 Required Analysis Documentation fleferences: Notes:
- 1. EDS Nuclear Report #02-0370-1045 B. This equipment will either be replaced-
- 2. EDS Report Ref. File 0370-024-831 Dated 10/16/80 or qualified by July 1, 1982.
C. Oyster Creek FDSAR Amen. 68 ADANO441
Fecility: Oyster Creek Sheet 32 o g , _ __ ,,, unii: System Component Evaluation Work Sheet. Environment Documentation Ref. Qualification Equipment Descr,pt,oni i Outstandin6 Method items Parameter Specification Qualification Specification Qualification System: Reactor Water Operating 30 days See Note A See Note C See Note B See Chpt. 7 1.evel Tirne Item 10 Plant ID No; IA-12A Temperature See Note A See Note B See Chpt. 7 g Component: Transmit ter ( F) 230 Item 10 Manufacture: Pressure 16 See Note A 1 See Note B See Chpt. 7 GE (PStA) Item 10 Model Number: 553 Relative See Chpt. 7 ilumidits 100 See Note A 1 See Note B WI Item 10 Function: _ Accuracy H Not Not - - - - - - Not ------ Spray Applicable Required Required Spec: Deino: See Note A 1 See Note B Sec Chpt. 7 Radiation 1.4x104R Item 10 Service: See Chpt. 7 C( ntainment Bldg. Aging 40 years See Note A 1 See Note B Location: El. 72' Item 10 Flood Level: Not Not Abcve Flood Level: Yes submergence Negligible Required 2 Required Analysis Documentation
References:
Notes: A. Not available at this time.
- 1. EDS Nuclear Report #02-0370-1045 B. This equipment will either be replaced
- 2. EDS Report Ref. File 0370-024-831 Dated 10/16/80 or qualified by July 1, 1982.
C. Oyster Creek FDSAR Amend. 68 AOOOO441
O () /~'i G L/ L ,1 Facility: Oyster Creek 33 _,,__. Sheet unit: System Component Evaluation Work Sheet Environment Documentation Ref. Qualification Equipment Descript. ion Outstanding Parameter Specification Method iteins Qualification Specification Qualification System: Reactor Water Operatin9 30 days See Note A See Note C See Note B See Chpt. 7 Level Time Item 10 Plant ID No: IA 12B Temperature See Note A See Note B See Chpt. 7 215 1 Component: Transmitter ("F) Item 10 Manufacture: Psessure 16 See Note A See Note B See Chpt. 7 CE 1 (PSIAI Item 10 Model Numbu: 553 fletativo See Chpt. 7 IlumiditY 100 'See Note A See Note B N 1 Item 10 Function:
" Not Not - - - - -
Not Accuracy Spray Applicable Required Required Spec: Demo: I adiation 6.1x10 4R See Note A 1 See Note B See Ch p t . -7 Item 10 Service: Containment Bldg. Aging 40 years See Note A 1 See Note B Location: El. 72 Item 10 Flood Level: Not Not Above Flood I.evel: Yes Jubmergence Negligible Required 2 Required Analysis Documentation fleferences: Nutes:
- 1. EDS Nuclear Report #02-0370-1045 B. This equipment will either be replaced
- 2. EDS Report Ref. File 0370-024-831 Dated 10/16/80 or qualified by July 1, 1982.
C. Oyster Creek FDSAR Amend. 68 annnn44,
O O O Facility: Oyster Creek unit: Sheet 34 _ og System Component Evaluation Work Sheet Environment Documentation Rei, Equipment Description coatification outstanding Parameter Specification Qualification Specification . Memod hems l Qualification System: Core Spray Operating System See Chpt.7 Time 30 days See Note A See Note C See Note B Item 11 Plant ID No: RV-29-A Toniperate See Chpt.7 77 See Note A Component: Presuure ( "F) 1 See Note B Item 11 Switch Manufactuse: Mercoid Pressure See Chpt.7 15 See Note A (PSIA) 1 See Note B ltem 11 Model Number: 9-51/ DAW Helative 43-156-R2IE ilumidity 100 See Note A 1 See Note B Function: (%) m Accuracy Not Applicabl( Not Required ------ Not Required --- Spec: Demo: Radiation 4.4x10 5'R 5 lx10 R 1 3 analysis Service: Reactor Buildin) Aging See Chpt.7 Location: El. -19, 40 years See Note A 1 See Note B Item 11 Flood Level: Above Flood Level: Yes Submergence Negligible Not Required 2 Not Required analysis Documentation Heferences: 1. EDS Nuclear Report #02-0370-1045 Notes: A. Not available at this time.
- 2. EDS Report Ref. File 0370-024-831 B.
Dated 10/16/80 This equipment will either be replaced or qualified by 7/1/82.
- 3. IE Bulletin 79-OlB C. Oyster Creek FDSAR Amend. 68
.noen..,
i n
%J O O i Fecility: Oyster Creek Sheet _35__ of i
unit: System Component Evaluation Work Sheet l l Environment Documentation Ref. Equipment Descript,on i Qualification ouistanding Parameter Specification Qualification Meinod liems Specification Qualdication System: Core Spray Operating 30 days E System Time See Note A See Note C See Note B Item 11 Plant ID No: RV-29-B - Temperatuse See Chpt.7 77 See Note A 1 See Note B Component: Pressure ( *F) Item 11 Switch - Manufacture: Mercoid 'e 15 See Chpt.7 See Note A 1 See Note B Item 11 Model Number: 9-51/ DAW Relative 43-156-R2IE ilumidity See Chpt.7 100 See Note A See Note B Function: (%) 1 Item Il Chemical Accuracy Not Applicablo Not Required ------ Not Required Spray Spec: Demo: 5 5 Radiation 2.6x10 R lx10 R 3 1 analysis Service: - Reactor Buildin ; Aging
- See Chpt.7
, 40 years See Note A 1 See Note B Item 11
-19 El.
Flood Level: Above Flood Level: Yes Submergence Negligible Not Required Not Required analysis Documentation
References:
- 1. EDS Nuclear Report #02-0370-1045 Notes: A. Not available a this time.
- 2. EDS Report Ref. File 0370-024-831 B. This equipment 7111 either be replaced or Dated 10/16/80 qualified by 7/1/82.
- 3. IE Bulletin 79-01B C. Oyster Creek FDSAR Amend. 68 AnnOO441
f" ( N.)s L 0) Fccility: Oyster Creek Shest 36 og unit: System Component Evaluation Work Sheet Environment Documentation Ref. Equipment Description Qualification outstandin0 Parameter Specification Mediod iteins Qualification Specification Qualification System: Core Spray ope,atino 30 days S e Chpt.7 System Time See Note A See Note L, See Note B Item 11 Plant ID No: RV-29-C See Chpt.7 Tennperature 77 See Note A 1 See Note B Component: Pressure t*F) Item 11 Switch -
""' 8 See Chpt.7 Manufacture: Mercold 15 See Note A 1 See Note B 7te,11 Model Number: 9-51/ DAW Helative 43-156-R21E Humidity See Chpt.7 100 See Note A 1 See Note 11 Function: IN Item 11 Cliendcal Accuracy Not Applicablo Not Required ------
Not Required Spray Spec: Hadiation 4.4x10 5R lx10 R 1 3 analysis Service: ,
^9 in0 See Chpt.7 beation, Reactor Buildini 40 years See Note A 1 See Note B Item 11
-19 E1.
Flood Level: Above Flood Level: Yes Submergence Negligible Not Required Not Required analysis Documentation
References:
- 1. EDS Nuclear Report #02-0370-1045 Notes: A. Not available at this time.
- 2. EDS Report Ref. File 0370-024-831 B. This equipment will either be replaced or Dated 10/16/80 qualified by 7/1/82.
- 3. IE Bulletin 79-01B C. Oyster Creek FDSAR Amend 68 AOOOO441
O O O Facility: Oyster Creek Sheet 3 ' _ . . o f __ __ unit: System Component Evaluation Work Sheet Environment Documentation Ref. Equipment Description Qualification Outstanding Paraineter Specification Method heens Qualification Specification Qualification System: Core Spray
- Operating E*
System Time " N A "N * " " 30 days Itemil Plant ID No: RV-29-D Temperatuto See Chpt.7 77 See Note A 1 See Note B Cornponent: Pressure ( F) Item 11 Switch Manufacture: Mercold Pe re 15 See Chpt.7 Model Number: 9-51/ DAW Relative 43-156-R2IE ilumidity See Chpt.7 100 See Note A 1 See Note B Function: (%) Item 11 b" I Accuracy Not Applicablo Not Required Not Required Spray Spec: Radiation 2.6x10 5R 5 lx10 R 1 3 analysis Service: A 9inD See Chpt.7 Location: Reaytor BuMm . 40 years See Note A 1 See Note B Item 11
-19 El.
Flood Level: Above Flood Level: Yes Submergence Negligible Not Required Not Required analysis Documentation
References:
- 1. EDS Nuclear Report #02-0370-1045 Notes: A. Not available at this time.
- 2. EDS Report Ref. File 0370-024-831 B. This equ ment will either be replaced or Dated 10/16/80 qualifleu / 7/1/82.
- 3. IE Bulletin 79-OlB C. Oyster Creek FDSAR Amend 68 A0000443
O (h () V V V Fccility: Oyster Creek Sheet 38 og unit: System Component Evaluation Work Sheet Environment Documentation Rel. Equipment Descript. : Qualification outsianding Parameter Specification Qualification Method items Specification Qualification System: Core Spray Operating 30 days See Chpt.7 System See Note A See Note C See Note B Time Item 12 Plant ID No: RV-s0-A - Temperature See Chpt.7 230 See Note A 1 See Note B Component: Pressure g p3 Item 12 - Switch Manufacture: Mercold Pres mr See Chpt.7 16 See Note A 1 See Note B Item 12 Model Number: 9-51/ DAW Ilciative 43-156-R2II. Humidity See Chpt.7 100 See Note A 1 See Note B Function: I"M Item 12 Chemical Accuracy Not Applicabl( Not Required - - = Not Required -- Spray Spec: Demo: lladiation
- 1.7x10 5R ""#'Y" "
Service: ..
"'d" ^ '"
- See Chpt.7 Location: E . 55' 40 years See Note A 1 See Note .r5 Item 12 Flood I.evel:
?
Above Flood Level: Yes Submergence Negligible Not Required - Not Required analysis Documentation
References:
- 1. EDS Nuclear Report //02-0370-1045 Notes: A. Not available at this time.
- 2. EDS Report Ref. File 0370-024-831 B. This equipment will either be replaced or Dated 10/16/80 qualified by 7/1/82.
- 3. IE Ilulletin 79-01B C. Oyster Creek FDSAR Amend. 68 ADOOO441
t i Fccility: Oyster Creek Sheet 39 og-L unii: System Component Evaluation Work Sheet l { Environment Documentation Ref. Equipment Descript,on i c .alification Outstanding Parameter Specification Qualification Meitiod lie m s Specification Qualification System: Core Spray Operatino 30 days 6"" O'P'*2 System See N te A See Note C See Note B Time Item 12 Plant ID No: RV-40-B See Chpt.7 Temperatuse 77 See Note A 1 See Note B - Component: Pressure i F) Item 12' Switch Manufacture:
- See Chpt.7 Mercold 15 S See Note A 1 See Note B Item 12 Model Number: 9-51/ DAW Ilelative 43-156-R2IE ilumidity See Chpt.7 100 See Note A See Note B Function: (%)
1 Item 12 Accuracy , Not Applicabl( Not Required - Not Required ------- Spec: 5 Demo: 2.3x10 R Iladiation lx10 R 1 3 analysis Service: -
"I d I" ^ " '
See Chpt.7 Location: 40 years See Note A 1 See Note B Item 12 Flood Level: Ahove Flood Level: Yes Submergence Negligible Not Required Not Required analysis Documentation fleferences: 1. EDS Nuclear Report #02-0370-1045 Notes: A. Not available at this time.
- 2. 1:DS Report Ref. File 0370-024-831 B. This equipment will either be replaced or Dated 10/16/80 qualified by 7/1/82.
- 3. IE Bulletin 79-OlB C. Oyster Creek FDSAR Amend. 68 AGOOn441
O O O Facility: Oyster Creek Sheet 40 og unit: System Component Evaluation Work Sheet Environment Documentation Ref. Equipment Description Qualification Ouisianding Paraincter Specification Method items Qualification Specification Qualification System: Core Spray Operating " "E ' System 30 days See Note A See Note C See Note B Time Item 12 Plant ID No: RV-40-C See Chpt.7 Teniperature See Note A 1 See Note B Component: pressure ( FI 230 Item 12.- Switch Manufacture: Mercold * '" See Chpt.7 See Note A 1 See Note B 3 Item 12 Model Number: 9-51/ DAW Relative 43-156-R2IE See Chpt.7 Humidity 100 See Note A 1 See Note B Item 12 Function: (%) Clienncal Accuracy Spray Not Applicable Not Required ------ Not Required -- Spec:
*" 5 Radiation 1,7xlo SR 1x10 R 1 3 analysis Service: .,
Reactor Build. See Chpt.7 A9 no 40 years Location: El. 55' See Note A 1 See Note B Item 12 Flood Level: Above Flood Level: Yes Submergenct Negligible Not Required Not Required analysis Documentation
References:
- 1. EDS Nucicor Report #02-0370-1045 Notes: A. Not available at this time.
- 2. EDS Report Ref. File 0370-024-831 B. This equipment will either be replaced or Dated 10/16/80 qualified by 7/1/82.
- 3. IE Bulletin 79-01B C. Oyster Creek FDSAR Amend. 68 Annnn441
O O O Facility: Oyster Creek Sheet __4.I_ o f - unit: System Component Evaluation Work Sheet Environment Documentation Ref. Equipment Descr,pt,on i i Qualification Ouistandina Parameter Specification Method items Qualification Specification Qualification System: Core Spray Operatino S Chpt.7 System 30 days See Note A See Note C See Note B Time Its 12 Plant ID No: RV-40-D See Chpt.7 Temperature 77 See Note A 1 See Note B Component: Pressure ( F) Item 12-Switch Manufacture: Mercold Pressure See Chpt.7 (PSIA) See Note A 1 See Note B 16 Item 12 Model Number: 9-51/ DAW Delative See Chpt.7 43-156-R21E ilumiditY 100 See Note A 1 See Note B Function: I"A'I Item 12 C m cal Accuracy Not Applicablo Not Required -- Not Required 3 Spec: Demo: 5 5 Radiau.on 2.3x10 R lx10 R 1 3 analysis Service: Containment A0inD , See Chpt.7 Location: El. 27, 40 years See Note A' 1 See Note B Item 1 Flood Level: Above Flood Level: Yes Submergence Negligible Not Required Not Required analysis I)ocumentation
References:
- 1. EDS Nuclear Report #02-0370-1045 Notes: A. Not available at this time.
- 2. EDS Report Ref. File 0370-024-831 B. This equipment will either be replaced or Dated 10/16/80 qualified by 7/1/82.
- 3. IE Bulletin 79-01B C. Oyster Creek FDSAR Amend. 68 ADOOO 4 4 'l
,) n /~N '
i, m- ( U i Facility: Oyster Creek - Shrst 42 cf . unit: System Component Evaluation Work Sheet ! t 1 Environment Documentation Ref. Equipment Descrs,pt,on i Qualification Outstanding Par ameter' Specification Qualification Specification Method items Qualification System: Operating Core Spray System Time 30 days See Note A See Note C See Note B See Chpt.7 , Item 12 l' Plant ID fJo: RV A Temperatus e 230 See Note A See Note B See Chpt.7 Component: ( F) 1 Item 12 Pressure Switch Manufacture: G.E./MAC Pressur I See Chpt.7 16 See Note A See Note B (PSIA) Item 12 Model Number: 552 Relative ilumidity 100 See Note A See Note B See Chpt.7 Function: (%) Item 12 Chemical Not Not Not Accuracy Applicable Required Required Spray Spec: Radiation 3.9x10 5R See Note A 1 See Note B ' ice Chpt. 7 Item 12 Service: bcatiom Reactor Bldg. Agin0 40 years See Note A 1 See' Note B See Chpt.7 E1. 55' Item 12 Flood I.evel: Above Flood Level: Yes Submergence Negligible not required 2 act required analysis Documentation
References:
- 1. EDS Nuclear Report #02-0370-1045 Notes: A. Not available at this time.
- 2. EDS Report Ref. File 0370-024-831 B. This equipment will either be replaced or Dated 10/16/80 .
qualified by 7/1/82.
- 3. C.E. Letter and Test Report #C-EN-0-163 C. Oyster Creek FDSAR Amend. 68 Dated 10/16/80.
A OOOO4 4 'l
A C \
! ) U e Facility: Oyster Creek ' ,
Sh:et _.43___ of !' unit: System Component Evaluation Work Sheet I 1 Environment Documentation Ref. Equipment Description oualification outstanding Parameter. Specification Qualification Specification Method items Qualification System: operating Core Spray System Time 3n days See Note A See Note C See Note B See Chpt.7 Item 12 Plant ID No-RV-26-B Temperature 225 See Chpt.7 See Note A See Note B Component: ( *F) 1 Item 12 Pressure Switch Manufacture: G.E./MAC Pressure I See Chpt.7 16 See Note A See Note B (PSIA) Item 12 Model Number: 552 flelative
'llumidity 100 See Note A See Note B See Chpt.7 Function: 1%) Item 12 Chemical Not Not Not Accuracy Applicable Required Required Spray Spec:
Demo- ' Iladiation See Note A 1 See Note B fice Chpt. 7 8.0x10 5R Item 12 Service: g Reactor Bldg. Aging 40 years See Note A 1 See Note B See Chpt.7
;9 El. 81 Item 12 Flood Level:
Above Flood Level: Yes Submergence Negligible not required 2 not required analysis Documentation lleferences: 1. EDS Nuclear Report #02-0370-1045 Notes: A. Not available at this time.
- 2. EDS Report Ref. File 0370-024-831 B. This equipment will either be replaced or Dated 10/16/80 qualified by 7/1/82.
- 3. C.E. Letter and Test Report #C-EN-0-163 C. Oyster Creek FDSAR Amend. 68 Dated 10/16/80.
AOoOo443
i w w a
)
Oyster Creek 44 Facility: Sheet o f _ __. __ unii: System Component Evaluation Work Sheet Environment Documentation Ref. Qualification Equipment Desen,pt,on i outstandind Method items
,%rameter S;)ecification Qualification Specification Qualification Con U S Note A See Note C See Note B See Chpt. 7 er t ,f,',"'* 10 minutes Plant ID No; IB06E U**
Temperature 270 See Note A g hhB h @. 7 Component: Temperature Detector Manufacture: Pressure 16 See Note A 1 See Note B See Chpt. 7 Rochester (PSIA) Instrument System Item 13 Model Number: Correct # Relative See Chpt. 7 not. available at Humidity 100 See Note A See Note B 1 Item 13 Funt.r.un: pr sent. (%) Chentical Nor Not ------- Not -- Accurucy Spray Applicable Required Required Spec: -~ __ Demo: See NMe A See Note B hM7 Radiation 1
~
(6.1x10 4R Item 13 Service: See Chpt. 7 Reactor Buildin) Aging 40 years See Note A 1 See Note B Location: El. I15 Flood Level: Not Not Above Flood Level: Yes submergence Negligible Required 2 Required Analysis Documentation
References:
Notes. A. Not available at this time.
- 1. EDS Nuclear Report #02-0370-1045 B. This equipment will either be replaced
- 2. EDS Report Ref. File 0370-024-831 Dated 10/16/80 or qualified by July 1, 1982.
C. Oyster Creek FDSAR Amend. 68 umnow.
1 (3> f3 C/ Q V Facility: Oyster Creek Sheet 45 og _ _ unit: System Component Evaluation Work Sheet Environment Documentation Ref. Qualification Equipment Description . outstandin6 Specification Method umns Parameter Qualification Specification Qualification Systern: Isolation Operating See Note A See Note C See Note B See Chpt. 7 Condenser System Time 10 minutes Item 13 Plant ID No IB06F Ternperature See Note A See Note B See Chpt. 7 270 g Component: Temperature I "fl
%g Detector Manufacture: Rochester P' essure 16 See Note A 1 See Note B See Chpt. 7 Instrument System (PSIA) Item 13 Model Nurnber: Correct # Relative See Chpt. 7 not available at present. Humidity 100 See Note A 1 See Note B Function: (%) Item 13 Chemical Not Not -------
Not = - - Accuracy Spray Applicable Required Required Spec: I
*" See N te A See Note B hM7 Hadiation 1 (6.1x10 R Item 13 Service:
See Chpt. 7 Containment Bldg. Aging 40 years See Note A 1 See Note B Location: El. 115' Item 13 Flood Level: Not Not Above Flood Level: Yes 3ubmergence Negligible Required 2 Required Analysis Documentation
References:
Notes:
- 1. EDS Nuclear Report #02-0370-1045 B. This equipment will either be replaced
- 2. EDS Report Ref. File 0370-024-831 Dated 10/16/80 or qualified by July 1, 1982.
C. Oyster Creek FDSAR Amend. 68 AnOOn441
n U f3 ,Q L/ () Facility: Oyster Creek Sheet 46 og __ u nit: System Component Evaluation Work Sheet Environment Documentation Ref. Qualification Equipment Descr,pt,on i i Outstandind Method items Parameter Specification Qualification Specification Qualification System: Isolation Operating See Chpt. 7 See Note A See Note C See Note B Condenser System Time 10 minutes g3 Plant ID No: IB06G Temperature 280 See Note A See Note B See Chpt. 7 g Component: Temperature ( F) Item 13 Detector Manufacture: Rochester Pressure 16 See Note A 1 See Note B See Chpt. 7 Instrument System (PSIA) Item 13 Model Number: Correct # Helative See Chpt. 7 not available at present. IlumiditV 100 See Note A See Note B (%) 1 Item 13 Function: Chemical Not Not ------- Not --- Accuracy Spray Applicable Required Required Spec: Demo: 1.0x10 R See Note A 1 See Note B See Chpt. 7 Item 13 Service: - See Chpt. 7 Reactor Buildini AginD 40 years See Note A 1 See Note B Location: El. 90 Item 13 Flood Level: Not Not Above Flood 1.evel: Yes submergence Negligible Required 2 Required Analysis Documentation Heterences: Notes: A. Not available at this time.
- 1. EDS Nuclear Report #02-0370-1045 B. This equipment will either be replaced
- 2. EDS Report Ref. File 0370-024-831 Dated 10/16/83 or qualified by July 1, 1982.
C. Oyster Creek FDSAR Amend. 68 A000044'4
O O O Facility: Oyster Creek Sheet 47 of'-~ unit: System Component Evaluation Work Sheet Environment Documentation Ref. Qualification Outstanding Equipment Description th h nis Parameter Specification Qualification Specification Ot.alification f'I" 10 minutes See Note A See Note C See Note B bee Chpt. 7 nder S tem Plant ID No; I B0611 Temperature 280 See Note A See Note B See Chpt. 7 g Component: Temperature ( F) Item 13 Detector Manufacture: Rochester Pressure 16 See Note A 1 See Note B See Chpt. 7 (F,SI A) Instrument System Item 13 Model Number: Correct # Relative not available at present. Ilumidity 100 See Note A 1 See Note B Item 13 Function: (%) Chemical Not Not - - - - - - Not ------ Acceracy Spray Applicable Required Required Spec: Demo: 5 See Note A See Note B See Chpt. 7 Radiation 1 9.4x10 R Item 13 Service-Containment See Chpt. 7 E1. 90, Aging 40 years See Note A 1 See Note B Location: Item 13 Flood Level: Not het Above Flood Level: Yes Submergence Negligible Required 2 Regtired Analysis Documentation
References:
Notes: A. Not available at this time.
- 1. EDS Nuclear Report #02-0370-1045 B. This equipment will either be replaced
- 2. EDS Report Ref. File 0370-024-831 Dated 10/16/80 or qualified by July 1, 1982.
C. Oyster Creek FDSAR Amend. 68 AnOOO441
O O O 48 gg Facility: Oyster Creek Shut unii: System Component Evaluation Work Sheet Environment Documentation Ref. Qualification Equipment Description outstandin9 Method items Parameter Specification Qualification Specification Qualification System: Containment Operating See note Spray 4 hours 5 hours C 3 test Time Plant ID No: IP-03-A - Temperature 77 184 1 3 test Component: Flow ( F) Transmitter - Manuf acture: G.E. Pressure 3 test 15 15 1 (PSIA) Model Number: 553 Relative Humidity 100 95 1 3 test Function: (%) _ Chemical not not = not = - - - Accuracy S ray applicable required required Spec: See note See note See Chpt. 7 Demo: 5 Radiation 3.4x10 A 1 B E;rvice: See note See note See Chpt. 7 North Aging 40 years A 1 B Location: C B RB Wal1 Flood Level: not Above Flood Level: Yes Submergence negligible requ ired 2 not required analysis Documenution Relerences: 1. EDS Nuclear Report #02-0370-1045 Notes: A. Not available at this time.
- 2. EDS Report Ref. File 0370-024-831 dated 10/16/80 B. This equipment will either be replaced
- 3. G.E. Letter and Test Report #G-EN-0-163 or qualified by July 1, 1982.
dated 10/16/80 C. Tech. Spec. 3.3 AoooO443
O O . O Fecility: Oyster Creek Sheet 49 of unit: System Component Evaluation Work Sheet Environment Documentation Ref. Equipment Description Qualification Outstanding Parameter Specification ** 'I"** Qualification Specification Qualification System: Containment Operating See note Spray 4 hours 5 hours C 3 test Time Plant ID No; IP-03-B
. Temperature 184 1 3 test 140 Component: Flow ( "F)
Transmitter Manuf acture: G .E . Pressure 15 15 1 3 test (PSIAI Model Number: 553 flelativo Humidity 100 95 1 3 test Function: (%) Chemical not not -- not Accinacy applicable required required Spray Spec: Demo: See note See note See Ch pt. 7 fladiation 1.7x106R A 1 B Service: S n te See note See Chpt. 7 R -R South 40 years A Location: C B AUi"9 1 B Item 14 Rn Wall Flood Level: not Above Flood Level: Yes Submergence negligible required 2 not required analysis Documentation fleferences: 1. EDS Nuclear Report #02-0370-1045 Notes: A. Not available at this time.
- 2. EDS Report Ref. File 0370-024-831 dated 10/16/80 B. This equipment will either be replaced
- 3. G.E. Letter and Test Report #G-EN-0-163 or qualified by July 1, 1982.
dated 10/16/80 C. Tech. Spec. 3.3 A nt M A 41
O O O Frcility: Oyster Creek Sheet 50 of. unit: System Component Evaluation Work Sheet Environment Documentation Ref. Equipment Descript,on i Quahlication Ouistandin0 Parameter Specification Method items Qualification Specification Qualification System: Containment 'O I 9 30 days 30 days See Note A 4 analysis Spray System T Plant ID No: IP-15-A Temperature 77 212 1 4 test Component: Contat. ( *F) Press. Switch Manufacture: Barton Pressure 15 15 1 4 test (PSIA) Model Number: 288A Relative ilumidity 35 100 1 4 test Function: (%) Chemical Not Not __ Not ______ Spray Applicable Required Required Spec: Radiation 3.9x10 5R lx MR 1 3 test Service: 40 years 40 years 5 See Chpt.7 Aging 1 . Location: RKO-3 El. 55' Item 15 Flood I.evel: Not Not Above Flood Level: Yes Submergence Negligible Required 2 Required analysis Documentation
References:
- 1. EDS Nuclear Report # 02-0370-1045 Notes: A. Oyster Creek FDSAR Amend. 68
- 2. EDS Report Ref. File 0370-024-831 Dated 10/16/80
- 3. Radiation Effects on Organic Materials by Robert Bolt & James Carroll, Academic Press, 1963
- 4. ITT Barton Document #9999.1217.2
- 5. Wyle Lab Study a nnnn wi
N O % O (v 3 J Facility: Oyster Creek Sheet 51 og unii: System Component Evaluation Work Shoot Environment Documentation Rel. Qualification outstanding Equipment Description Specification Method items Parameter Quaiilication Specification Qualification System: Containment oD I9 30 days 30 days See Note A 4 analysis Spray System 7; Plant ID No: IP-15-B Temperature 77 212 1 4 test Component: Centmt. Press. ( F) Switch Manufacture: Barton Pressure 15 15 1 4 test (PSIA) Model Number: 288A Ilciative llumidity 35 100 1 4 test Function: 1%) Chemical Not Not Not ___,__ Y Spray Applicable Required Required Spec:
"# 5 Radiation 3.9x10 R lx10 6R 1 3 test Service: --
40 years See Chpt.7 Aging 40 years 1 .. 5 Location: RKO-3 El. 55' Item 15 Flood Level: Not Not Above Flood Level: Yes Submergence Negligible Required 2 Required analysis Documentation lleferences: 1. EDS Nuclear Report # 02-0370-1045 Notes: A. Oyster Creek FDSAR Amend. 68
- 2. EDS Report Ref. File 0370-024-831 Dated 10/16/80
- 3. Radiation Effects on Organic Materials by Robert Bolt & James Carroll, Academic Press, 1963
- 4. ITT Barton Document #9999.1217.2
- 5. Wyle Lab Study ADOOO441
[] O [') G U %) Facility: Oyster Creek Sheet _52 og unii: System Component Evaluation Work Sheet Environment Documentation Ref. oualdication ouistamhng Equipment Description _ Parameter Specification Qualification Specification Qualification System: Containment Operating Spray System 30 days 30 days See Note A 4 analysis Time Plant ID No: IP-15-C Temperature 77 212 1 4 test Component: Contmt. Press. ( F) . Switch -- Manufacture: Barton Pressure 15 1 4 test (PSIA) Model Number: 288A Ilelative ilumidity 35 100 1 4 test Function: (%) Chemical Not Not Not Accuracy S ray Applicable Required Required Spec: __ Radiation 3.9x10 5g lx10 6R 1 3 test Service: See Chpt.7 Aging 40 years 40 years 1 . 5 Location: RKO-3 El. 55' Item 15 Flood Level: Not Not Above Flood Level: Yes Submergence Negligible Required 2 Required analysis Documentation fleferences: 1. EDS Nuclear Report # 02-0370-1045 Notes: A. Oyster Creek FDSAR Amend. 68
- 2. EDS Report Ref. File 0370-024-Dated 10/16/80
- 3. Radiation Effects on Organic h erials by Rabert Bolt. & James Carroll, Academic Press, 1963
- 4. ITT Barton Document #9999.1217.2
- 5. Wyle Lab Study AOOOO441
(~ 0 O
, a -
O Facility: Oyster Creek Sheet 53 og _ unit: System Component Evaluation Work Sheet Environment I Documentation Ref. Qualification outstanding Equipment Deser,pt,on t i Pararneter Specification Method iteens Qualification Specification Qualification System: Containment Spray System
""i"9 30 days 30 days See Note A 4 analysis Plant ID No: IP-15-D Temperature 77 212 1 4 test Cornponent: Contmt. Pres: . ( *F)
Switch Manufacture: Barton Psessure 15 15 1 4 test (PSIA) Model Nuinber: 288A Relative llurnidity 35 100 1 4 test Function: (%) Chesnical Not Not Not Accuracy S ray Applic.51e Required Required Spec: Hadiation 3.9x10 5g 6 lx10 R 1 3 test Service. 40 years 40 years 5 See Chpt.7 Aging 1 .. l_ocation: RKO-3 El. 55' Item 15 Flood Level: Not Not Above Flood Level: Yes Submergence Negligible Required 2 Required analysis Docuruentation Heterences: 1. EDS Nuclear Report # 02-0370-1045 Notes: A. Oyster Creek FDSAR Amend. 68
- 2. EDS Report Ref. File 0370-024-831 Dated 10/16/80
- 3. Radiation Effects on Organic Materials by Robert Bolt 6 James Carroll, Academic Press, 1963
- 4. I'IT Barton Document #9999.1217.2
- 5. Wyle Lab Study anann..,
O O 54 O Fgcility: Oyster Creek Shnt gg _ , unii: System Component Evaluation Work Sheet Environment Documentation Ref. Qualification outstanding Equipment Descr,ipt,oni
,g , ,,
Parameter Specification Qualification Specification Qualification System: Drywell Pressure Operating See Note C 3-30 days '30 days analysis Switches & Indicators Time Plant ID No: IP-07 Temperature 77 184 1 3 Test Component: Drywell Prs. ( F) Pressure Trans. - Pressure 15 15 1 3 Test Manufacture: CE/MAC (PSIA) Model Number: 553 IIelative catalog # VPF 1680 liumidity 35 95 1 3 Test Function: (%) Chemical Accuracy Spray not not not == Spec: applicable required required Demo fladiation 3.9x10 5R Fee Note A 1 See Note B See Chpt.7 Item 13 Service:
^9 5"O 40 yrs. See Note A 1 See Note B Location:RKO-3 El. 55, SeeCgt.7 Item Flood Level: not not Above Flood Level: Yes Submergence negligible required 2 required analysis Documentation fieferences: Notes:
- 1. EDS Nuclear Report #02-0370-1045 A. Not available at this time.
- 2. EDS Report Ref. File 0370-024-831 dated 10/16/80 B. This equipment will either be replaced or
- 3. CE letter and Test Report #G-EN-0-163 dated 10/16/80 qualified by 7/1/82.
C. Oyster Creek FDSAR Amend 68. AOOOO441
Frcility: Oyster Creek sheet 55 of unit: System Component Evaluation Work Sheet 4 Environment Documentation Ref. Equipment Description Qualification outstanding Parameter Specification Qualification Specification Qualification Systern: Core Spray P I"U 30 days 30 days See Note A 4 analysis System Time Plant ID No: RV-46-A Temperature 77 212 1 4 test Component: Drywell Press (*F) . Switch Manufacture: Barton Pressure 15 15 1 4 test (P SIA) Model Number: 288A Relative Humidity 35 100 1 4 rest Function: (%) Chemical Not Not Not Accuracy - Spray Applicable Required Required Spec: Radiation 3.9x10 5g 6 lx10 R 1 3 tesc Service:
@ mm @ mm 1 5 M El. 55, no .
Location: RKO-3 Item 18 Flood Level: Not Not Above Flood Level: Yes Submergence Negligible g Required 2 Required analysis Documentation
References:
- 1. EDS Nuclear Report # 02-0370-1045 Notes: A. Oyster Creek FDSAR Amend. 68
- 2. EDS Report Ref. File 0370-024-831 Dated 10/16/80
- 3. Radiation Effects on Organic Materials by Robert Bolt & James Carroll, Academic Press, 1963
- 4. I'IT Barton Document #9999.1217.2
- 5. Wyle Lab Study annno m
O O O Fecility: Oyster Creek Sheet _ _56 og unit: System Component Evaluation Work Sheet Environment Documentation Ref. Equipment Description G" 8'Heation ouistanding Specification Method lle m s Parameter oualification Specification oualification Systern: Core Spray P "O 30 days 30 days See Note A 4 analysis System Time Plant ID No: RV-46-B Temperature 77 212 1 4 test Component: Drywell Press . I II - Switch Manufacture: Barton Pressure 15 15 1 4 test (PSIA) Model Number: 288A Relative Humidity 35 100 1 4 test function: ("M Chemical Not Not Not Accuracy Spray Required Required Applicable Spec-Demo: 5 Radiation 3.9x10 R 1xto 6R 1 3 test Service: 40 years 40 years 5 See Chpt.7 Aging 1 .. Location: RKO-3 El. 55' Item 18 Flood Level: Not Not Above Flood I.evel: Yes Submergence Negligible Required 2 Required analysis Documentation Refceences: 1. EDS Nuclear Report # 02-0370-1045 Notes: A. Oyster Creek FDSAR Amend. 68
- 2. EDS Report Ref. File 0370-024-831 Dated 10/16/8G
- 3. Radiation Effects on Organic Materials by Robert Bolt & James Carroll, Academic Press, 1963
- 4. ITT Barton Document #9999.1217.2
- 5. Wyle I.ab Study A(M.A441
O O O Facility: Oyster Creek Sheet 57 og_ unit: System Component Evaluation 1Nork Sheet Environment Documentation Ref. Qualification outstanding Equipnient Description Parameter Specification Qualification Specification Qualification Systern: Core Spray Operating 30 days 30 days See Note A 4 analysis System Time Plant ID No: RV-46-C -- Temperature 77 212 1 4 test Cornponent: Drywell ( *F) . Press. Switch Pressure Manufacture: Barton 15 15 1 4 test (PSIAI Model Nurnber: 288A Relative llunudity 35 100 1 4 test Function: (%) Chemical Not Not Not Y Spray Required Required Applicable Spec: Radiation 3.9x10 R 6 lx10 R 1 3 test Servico: See Chpt.7 A 0ing 40 years 40 years 1 . 5 Location: RKO-3 El. 55' Item 18 Flood I.evel: Not Not Above Flood Level: Yes Submergence Negligible Required 2 Required analysis Documentation lleferences: 1. EDS Nuclear Report # 02-0370-1045 Notes: A. Oyster Creek FDSAR Amend. 68
- 2. EDS Report Ret. File 0370-024-831 Dated 10/16/80
- 3. Radiation Effects on Organic Materials by Robert Bolt & James Carroll, Academic Press, 1963
- 4. ITT Barton Document #9999.1217.2
- 5. Wyle Lab Study AOoOo44*l
/*)
'w (G) ('))
x. Facility: nyster Creek Sheet 58 og umi: System Component Evaluation Work Sheet Environment Documentation Ref. Equipment Description Qualification outstanding Parameter Specification Qualification Specification Qualification Sys'em: Core Spray System P ' 'i"9 30 days 30 days See Note A 4 analysis liene Plant ID No: RV-46-D Temoerature 77 212 1 4 test Component: Drywell ( "F) . Press. FJitch n,+ afacture: Barton Pressure 15 15 1 4 test tiSIA) Model Number: 288A Relative flunddity 35 100 1 4 test Function: (%) Chemical Not Not Not g Spray Applicable Required Required Spec:
"# 5 Radiation 3.9x10 R 6 lx10 R 1 3 test Service:
40 years 40 years 5 See Chpt.7 Aging 1 .. Location: RKO-3 El. 55' Item 18 Flood Level: Not Not Above Flood Level: Yes Submergence Negligible Required 2 Required analysis Documentation
References:
- 1. EDS Nuclear Report # 02-0370-1045 Notes: A. Oyster Creek FDSAR Amend. 68
- 2. EDS Report Ref. File 0370-024-831 Dated 10/16/80
- 3. Radiation Effects on Organic Materials by Robert Bolt & James Carroll, Academic Press, 1963
- 4. ITT Barton Document #9999.1217.2
- 5. Wyle Lab Study AAAAA4A't
O O O Facility: Oyster Creek Sheet 59 og unii: System Component Evaluation Work Sheet Environment Documentation Ret. Equipment Descript. ion Qualification outstanding Paraincter Specification Qualification Meniod items Specification Qualification Systein: Reactor Isolatio3 operating Time 30 days 30 days See Note A 3 analysis Plant ID No: RE-23-A Temperature Component: MSL 1.ow Press I FI 77 77 1 3 analysis Switch Manufacture: Meletron Prom: 15 15 1 3 analysis Model Nuniber: 372 Helative
" " " ' 'Y 35 35 1 3 analysis Function: (%)
Chemical not not not ACcmaCY Spray applicable required required Spec:
- 5 Hadiation O .1x10 R lx10 R 1 4 analysig Service:
Reactor Fd. Pump
^9 i "9 40 years 40 years 1 analysis l_ocation: Room 3 wall E1. x 5N9rth Flood Level:
Above Flood level: Yes Submergence negligible not 2 not required required analysis Documentation Heferences: Notes: *
- 1. EDS Nuclear Report #02-0370-1045 A. Oyster Creek FDSAR Amend 68
- 2. EDS Report Ref. File 0370-024-831 dated 10/16/80
- 3. Wyle Lab. Study
- 4. IE Bulletin 79-01B AOOOO441
Facility:0yster Creek She::: 60 og unii: System Component Eva'uation Work Sheet Environment Documentation Ref. Equipment Description Qualification outstanding Parameter Specification Qualification Specification Qualification System: Reactor Isolati t operating Time 30 days 30 days See Note A 3 analysis Plant ID No: RE-23-B Temperature Component: MSL Low I fI 77 77 1 3 analysis Press. Switch Manuf acture: Meletron Pressur 15 15 1 3 analysis (PSIA) Model Number: 372 Helative
""*IdI'Y 35 35 1 3 analysis Function: 1%)
Chemical not not not Accuracy S ray applicable required required Spec: Radiation (6.1x10 R lx10 R 1 4 analysis Service: Reactor Fd. Pump A 40 years 40 years 1 3 analysis Location: gomgSgyth Flood Level: Above Flood Level: Yes Submergence negligible not 2 not required required analysis Documentation Heferences: Notes: '
- l. EDS Nuclear Report #02-0370-1045 A. Oyster Creek FDSAR Amend. 68
- 2. EDS Report Ref. File 0370-024-831 dated 10/16/80
- 3. Wyle Lab. Study
- 4. IE Bulletin 79-01B Anoo044'l
O O O Facility Oyster Creek Sheet 61 og unit: System Component Evaluation Work Sheet Environment Documentation Ref. Equipment Description Qualification outstanding Parameter Specification Qualification Specification "' " ' ** Qualification System: Reactor Isolatica Operating Time 30 days 30 days See Note A 3 analysis Plant ID No: RE-23-C y Temperature Component: MSL Low ( F) 77 77 1 3 analysis Press. Switch Manuf acture: Meletron Ps re 15 15 1 3 analysis Model Nurnber: 372 Relative
""*Idi'Y 35 35 1 3 analysis Function: Wal Chemical not not not Accuracy S ray applicable required required Spec:
Radiation (6.1x10 R lx10 5 R 4 1 analysis Service: Reactor Fd. Pump
^Ui "9 40 years 40 years 1 analysis Location: Room x North Wall 3 E1. S' Flood Level:
Above Flood Level: Yes Submergence negligible not 2 not required required analysis Documentation
References:
Notes: '
- l. EDS Nuclear Report 802-0370-1045 A. Oyster Creek FDSAR Amend. 68
- 2. EDS Report Ref. File 0370-024-831 dated 10/16/80
- 3. Wyle Lab. Study
- 4. IE Bulletin 79-OlB AnnOO443
O O O Facility:0yster Creek Shes: 62 ,,. ; unit: System Component Evaluation Work Sheet Environment Documentation Ref. Equipment Desen,pt,on i ouaiilication outstanding Parameter Specification oualification Specification Method heins oualification System: Reactor operating Isolation Time 30 days 30 days See Note A 3 analysis Plant ID No: RE-23-D s Temperature Component: MSL Low Pres. ( F ) 77 77 1 analysis 3 Switch Manuf acture: Me letron Pressur 15 15 1 3 analyals (PSIA) Model Numlier: 372 Ilelative
""*Idi'Y 35 35 1 3 analysis Function: (%)
Chemical not not not Accuracy S ray applicable required -- required - -- Spec: Demo: ( 6.1x10 4R lx10 5g fladiation Service: El. S'
, Reactor Fd. Pump Aging 40 years 40 years 1 analysis l_ocation: Room x South Wall 3 Flood Level:
Above Flood I.evel: Yes Subme rgence negligible not 2 not required required analysis Documentation lleferences: Notes: *
- 1. EDS Nuclear Report #02-0370-1045 A. Oyster Creek FDSAR Amend. 68
- 2. EDS Report Ref. File 0370-024-831 dated 10/16/80
- 3. Wyle Lab. Study
- 4. IE Bulletin 79-OlB ArKM)n443
O f% V V Ficility: Oyster Creek Sheet 63 og unii: - System Component Evaluation Work Sheet Environment Documentation Ref. Qualification outstanding Equipment Desen. t.p ion
- 11. d tieriis Pararneter Specification Qualification Specification oualification Systern: Reactor O 9 30 days 30 days See Note A 4 Isolation T analysis Plant ID No: RE-22-A Temperature 77 212 1 4 test Component: Reactor ( *F) .
isolation Switch Manufacture: Barton Pressure 15 15 1 4 test (PSIA) Model Number: 288A Helative ilumidity 35 100 1 4 test Function: (%) Chemical Not Not Not Accuracy S ray Applicable Required Required
~
Spec:
"* 4 Hadiation (6.1x10 R 6 lx10 R I 3 test Service:
40 years 40 years 1 5 See Chpt.7 R -R x Drywell AginD . Location: E p _ Item 20 IJn 11 El. 27' Flood Level: Not Not Above Flood Level: Yes Submergence Negligible Required 2 Required analysis Documentation
References:
- 1. EDS Nuclear Report # 02-0370-1045 Notes: A. Oyster Creek FDSAR Amend. 68 EDS Report Ref. File 0370-024-831 2.
Dated 10/16/80
- 3. Radiation Effects on Organic Materials by Robert Bolt & James Carroll, Academic Press, 1963
- 4. ITT Barton Document #9999.1217.2
- 5. Wyle Lab Study AOOOn441
Fccility: Oyster Creek sheet 64 og unit: System Component Evaluation Work Sheet Environment Documentation Ref. Equipment Description Qualification outstanding Parameter Specification Qualification Specification Qualification System: Reactor Isolatior' Operating 30 days 30 days See Note A 4 analysic Tune Plant ID No: RE-22-B Temperature 77 212 1 4 test Component: Reactor ( F) Isolation Switch Manufacture: Barton Pressure 15 15 1 4 test (PSIA) Model Number: 288A Relative flutnidity 35 100 1 4 test Function: (%) Chemical Not Not Not Accmacy Spray Applicable Required Required Spec: Hadiation <6.1x10 R 0 lx10 R 1 3 test Service: See Cbpt.7 RE -Rp x Drywell Aging 40 years 40 years 1 .. 5 Location: Item 20 Un11 F1- 978 Flood Level: Not Not Above Flood Level: Yes Submergence Negligible Required 2 Required analysis Documentation lieferences: 1. EDS Nuclear Report # 02-0370-1045 Notes: A. Oyster Creek FDSAR Amend. 68
- 2. EDS Report Ref. File 0370-024-831 Dated 10/16/80
- 3. Radiation Effects on Organic Materials by Robert Bolt & James Carroll, Academic Press, 1963
- 4. ITT Barton Document #9999.1217.2
- 5. Wyle Lab Study ANeAA 4 41
-v-.- - , - _ - - - - -
. 4 O O O Fachity: Oyster Creek Sheet 65 og unit. System Component Evaluation Work Sheet lI Environment Documentation flef. Qualification outstanding Equipment Description Parameter Specification Specification ! Qualification '"*' "*"*
Qualification System: Reactor Isolatioi t opmating 30 days 30 days See Note A 4 analysis Time Plant ID No: RE-22-C Temperature 77 212 1 4 test Component: Reactor I fl - Isolation Switch Manufacture: Barton Pressure 15 15 1 4 test (PS!A) Model Number: 288A Relative ilumidity 35 100 1 4 test Function: (%) Chemical Not Not Not Accuracy Spray Required
~
Applicable Required Spec: Hadiation (6.1x10 R lx10 R 6 1 3 test Service: See Chpt.7 R E '"F x Drywell Aging 40 years 40 years 1 . 5 Location: Wall El. 27' Flood Level: Not Not Above Flood Level: Yes Submergence Negligible Required 2 Required analysis Documentation fleferences: 1. EDS Nuclear Report # O2-0370-1045 fJotes: A. Oyster Creek FDSAR Amend. 68
- 2. EDS Report Ref. File 0370-024-831 Dated 10/16/80
- 3. Radiation Effects on Organic Materials by Robert Bolt & James Carroll, Academic Press, 1963
- 4. ITT Barton Document #9999.1217.2
- 5. Wyle I,ab Study ADOAO441
O O O Fccility: Oyster Creek Sheet 66 og unit: System Component Evaluation Work Sheet
" " " "" "I I "N Equipment Description Quatirication outstanding
* "*' """'S Parameter Specification Qualification Specification Quahfication System: Reactor P r linD 30 days 30 days See Note A 4 analysis Isolation Time Plant ID No: RE-22-D Temperature 77 212 1 4 test Compar.cnt: Reactor ( F)
Isolation Switch Manufacture: Barton Pressure 15 15 1 4 test (PS!A) Model Number: 288A Rdative Humidity 35 100 1 4 test Function: (%) Chemical Not Not Not _ Y Required Spray Applicable Required Spec: Hadiation (6.1x10 R 6 lx10 R 1 3 test Service: 40 years 40 years 5 See Chpt.7 R R x Drywell Aging 1 . Location: wh,ilp El. 27' Item 20 Flood I.evel: Not Not Above Flood Level: Yes Submergence Negligible Required 2 Required analysis Documentation
References:
- 1. EDS Nuclear Report # 02-0370-1045 Notes: A. Oyster Creek FDSAR Amend. 68
- 2. EDS Report Ref. File 0370-024-831 Dated 10/16/80
- 3. Radiation Effects on Organic Materials by Robert Bolt & James Carroll, Academic Press, 1963
- 4. ITT 11arton Document #9999.1217.2
- 5. Wyle Lab Study anannee,
O
~
O V (d' V Frcility: Oyster Creek , Sheet 67 og unit: System Component Evaluation Work Sheet Environment Documentation Rel. outstanding Equipment Descr,iption Qualification g g ,,,, Parameter Specification Qualification Specification Qualification r . System: Reactor Isolation Operating 30 days 30 days See Note A 4 analysis Tune Plant ID No: RE-22-E _ Temperature 77 212 1 4 test Component: Reactor ( *F) . Isolation Switch Manufacture: Barton Pressure 15 15 1 4 test (PSIA) Model Number: 288A Helative ilumidity 35 100 1 4 test Function: (%) Chemical Not Not Not Amg _ _ _ Spray Applicable Required Required Spec: Demo: 4 Hadiation 6 (6.1x10 R lx10 R 1 3 test Service: 40 years 40 years 5 See Chpt.7 R -R x Drywell Aging 1 . p Location: Item 20 333 Flood Level: Not Not Above Flood Level: Yes Submergence Negligible Required 2 Required analysis Documentation
References:
- 1. EDS Nuclear Report # 02-0370-1045 Notes: A. Oyster Creek FDSAR Amend. 68
- 2. EDS Report Ref. File 0370-024-831 Dated 10/16/80
- 3. Radiation Effects on Organic Materials by Robert Bolt & James Carroll, Academic Press, 1963
- 4. ITT Barton Document #9999.1217.2
- 5. Wyle Lab Study AOOnO441
Q V(3 O v (_./ Facility: Oyster Creel" Sheet . 68 og unii: System Component Evaluation Work Sheet Environment Documentation Ref. Qualification outstanding Equipment Description Parameter Specification Qualification Specification Qualification System: Reactor
'"U 30 day:; 30 days See Note A 4 Isolation g, Plant ID No: RE-22-F Ten.perature 77 212 1 4 test Component: Reactor ( *F) .
Isolation Switch Manufacture: Barton Pressure 15 15 1 4 test (PSIA) Model tjumber: 288A fielative ifumidity 35 100 1 4 test Function: 1%) Chemical Not Not ___ Not ______ q Y Spray Applicable Required Required Spec: Como: 4 Radiation (6.1x10 R lx10 6R 1 3 test Service: 40 years 40 years 5 See Chpt.7 RE - Ry x Drywell Aging 1 . Location: Wall Item 20 El. 27' Flood Level: Not Not Above Flood Level: Yes Submergence Negligible Required 2 Required analysis Documentation lleferences: 1. EDS Nuclear Report # 02-0370-1045 Notes: A. Oyster Creek FDSAR Amend. 68
- 2. EDS Report Ref. File 0370-024-831 Dated 10/16/80
- 3. Radiation Effects on Organic Materials by Robert Bolt & James Carroll, Academic Press, 1963
- 4. ITT Barton Document #9999.1217.2
- 5. Wyle Lab Study AOOOO441
O O O Ftcility: Oyster Creek Sheet _69 og unit: System Component Evaluation Work Sheet Environment Documentation Ref. Qualification outstanding Equipment Descr,pt,on i i Method items Parameter Specification Qualification Specification Qualification System: Reactor O II"U 30 days 30 days See Note A 4 analysis Isolation T Plant ID No: RE-22-G Temperature 77 212 1 4 test Component: Reactor { F) . Isolation Switch Manufacture: Barton Pressure 15 15 1 4 test (PSIA) Model Number: 288A Relative ilumidity 35 100 1 4 test Function: (M Chemical Not Nor Not Accuracy - S ray Applicable Required
~~
Required Spec: Demo: 4 Radiation (6.1x10R 6 lx10 R 1 3 test Service: - 40 years 40 years 5 See Chpt.7 R -R E p x Drywell Aging 1 . l_ocation: 1b11 El_ 77e Item 20 Flood I.evel: Not Not Alave Flood 1.evel: Yes Submergence Negligible Required 2 Required analysis Docuanentation llelerences: 1. EDS Nuclear Report # 02-0370-1045 Notes: A. Oyster Creek FDSAR Amend. 68
- 2. EDS Report Ref. File 0370-024-831 Dated 10/16/80
- 3. Radiation Effects on Organic Materials by Robert Bolt & James Carroll, Academic Press, 1963
- 4. ITT Barton Document #9999.1217.2
- 5. Wyle I.ab Study ADOnn441
O O O Fccility: Oyster Creek shee ,70 og unii: System Component Evaluation Work Sheet Environment Documentation Ref. Qualdication Etluipment Descript. ion outstandinD Parameter Specification Iterm Quahtication Specification Qualification System: Reactor Isolatio i operating s 30 days 30 days See Mote A 4 analysis Time Plant ID No; RE-22-li Temperature 77 212 1 4 test Component: Reactor ( "F) . Isolation Switch Manufacture: Barton Pressure 15 . 15 1 4 test (PSIA) Model Number: 288A fletative ilumidity 35 100 1 4 test Function: (%) Chemical Not Not __ Not _____ Y Spray Applicable Required Required Spec:
""* 4 Radiation (6.1x10R lx10 6R 3 1 test Service:
See Chpt.7 RE -R x Drywell 40 years 40 years 1 5 p A0inD . Location: Wall El. 27' Flood Level: Not Not Above Flood Level: Yes Submergence Negligible Required 2 Required analysis Documeetation fleferes.ces: 1. EDS Nuclear Report # 02-0370-1045 Notes: A. Oyster Creek FDSAR Amend. 68
- 2. EDS Report Ref. File 0370-024-831 Dated 10/16/80
- 3. Radiation Effects on Organic Materials by Robert Bolt & James Carroll, Academic Press, 1963
- 4. ITT Barton Document.#9999.1217.2
- 5. Wyle I.ab Study annnn..,
~
O O 71 O Fccifity: Oyster Creek Sheet g, unii: System Component Evaluation Work Sheet
! Environment Documentation Ref.
Equipment Descn,pt, ion Qualification Outstanding Method items Parameter 5pecification Qualification Specification Qualification System: Isolation Operating See Note A See Note C See Note B See Chpt. 7 Condenser Systen "'" Time Item 21 Plant ID No: IB-05-Al Temperature 230 See Note A See Note B See g t.7 g Component: Isolation ( F) Item Condenser AP Switch '- ~ P'essur See Chpt.7 Manufacture: Barton 8 " U"E" A See Note B (PSIA) 16 1 Item 21 Model Number: 288A Helative Ilumidity
. See Note A See Note B See Chpt.7 I%I 100 1 Item 21 Function:
Chernical Accuracy Not Not - Not - - - Spray Spec: Applicable Required Required Denm: 6 Radiation 3.9x10 5R lx10 R I 3 analysis Service: RK-03 Aging 40 years See Note A 1 See Note B 3ee Chpt. 7 Location: El. 55' Item 21 Flood I.evel: Above Flood I.evel: Yes Submergence Negligible Not Required 2 not required analysis Documentation Beierences: 1. EDS Nuclear Report #02-0370-1045 Notes: A. Not available at this time.
- 2. EDS Report Ref. File 0370-024-831 B. This equipment will either be replaced Dated 10/16/80 or qualified by 7/1/82.
- 3. Radiation Effects on Organic Materials C. Tech. Spec. 3.3 by Robert Bolt & James Carroll, Academic Press, 1963 AOOOO441
O O O Facility: Oyster Creek Sheet U of unit: System Component Evaluat. ion Work Sheet Environment Documentation fief. Equipment Description Qualification outstanding Parameter Specification Quahfication Specification Mafiod items Qualification System: Isola t.lon Operating Condenser System 4 hours See Note A See Note C See Note B See Chpt.7 Time hem M Plant ID No: IB-05-A2 Temperatura 230 See Note A See Note B See Chpt.7 I .,H I Item 21 Component: Isolation Condenser aP Switch - Psessur See Chpt.7 Manufacture: Barton See N te A See Note B (PSIA) 16 Item 21 Model Number: 288A fletative Ilunudity See Note A o See Note B See Chpt.7 100 Item 21 Function: IM 1 Chendcol Accuracy Not Not - Not --- Spray Spec: Applicable Required Required UU* 6 11adiation 3.9x10 5 R lx10 R I analysis 3 Service: -- RK-03 Aging 40 years See Note A 1 See Note B sae Chpt. 7 location: El. 55' Item 21 Flood Level: Above Flood Level: Yes Submergence Negligible Not Required 2 not reqstred analysis Documentation
References:
- 1. EDS Nuclear Report #02-0370-1045 Notes: A. Not available at this time.
- 2. EDS Report Ref. File 0370-024-831 B. This equipment will either he replaced Dated 10/16/80 or qualified by 7/1/82.
- 3. Radiation Effects on Organic Materials C. Tech. Spec. 3.3 by Robert Bolt 6 James Carroll, Academic Press, 1963 A fVVW)4 41
O O O Facility: Oyster Creek S;.eet 73 et unit: System Component Evaluation Work Sheet Environment Documentation Ref. Qualification outstanding Equipment Descn,pt, ion Mettiod hems Parameter Specification Quahlication Specification Qualification System: Isolation Operating See Note A See Note C See Note B See Chpt.7 Condenser System 4 hours Time Plant ID No: IB-05-B1 Temperature 230 See Hote A g See Note B See Chpt.7 Component: Isolation ( *F) Item 2 Condenser AP Switch - See Chpt.7 Manufacture: Barton P .sur See Note A See Note B (PSIA) 16 g Item 21 1 Model Nurnber: 288A fletative See Note A See Note B See Chpt.7 llumidity 100 1 Item 21 Function: 1%) Chernical Accuracy Not Not - Not S P '"Y Applicable Required Required Spec:
"# I Radiation 3.9x10 R lx10 R 3 analysis Service:
RK-03 Aging 40 years See Note A 1 See Note B See Chpt. 7 Location: El. 55' Item 21 Flood Lt. vel: Above Flood Level: Yes Submergence Negligible Not Required 2 not required analysis
)
Documentation
References:
- 1. EDS Nuclear Report #02-0370-1045 Notes: A. Not available at this time.
- 2. EDS Report Ref. File 0370-024-831 10 This equipment will either be replaced Dated 10/16/80 or qualified by 7/1/82.
- 3. Radiation Effects on Organic Materials C. Tech. Spec. 3.3 by Robert Bolt & James Carroll, Academic Press, 1963 AoOnO441
() - O V O V fccility: Oyster Creek Sheet 7.4 of unit: System Component Evaluat. ion Work Sheet Environment Documentation Ref. Equipment Desen,pt, ion Quanfication outstanding Parameter Specification Qualification Method items Specification Qualification Systern: Isolation Operating Condenser System See Note A See Note C See Note B See Chpt.7 Time 4 hours Item 21 Plant 10 No: IB-05-B2 ica nperature 230 See Note A See Note B "
- Component: Isolation l fl Condenser AP Switch Manufacture: Barton P'essor See Ch t.7 See Note A See Note B (PSIAI 16 g Item 2 .
Model Number: 288A fletative See Note A See dote B See Ch t.7 t h.msdaty 100 Item 2$ Function: I /"I 1 Cimnical Accuracy !!ot Not ------ Not --- Spray Spec: Applicabie Required Required Dmno: *9* Radiation lx10 R I analysis 3 Service: RK-03 Aging 40 years See Note A 1 See Note B See Chpt. 7 1.ocation: El. 55' Item 21 Flood Level: Above Flood Level: Yes Submergence Negligible Not Required 2 not required analysis Documentation llefmences: 1. EDS Nuclear Report #02-0370-1045 Notes: A. Not available at this time.
- 2. EDS Report Ref. File 0370-024-831 B. This equipment will either be replaced Dated 10/16/80 or qualified by 7/1/82.
- 3. Radiation Effects on Organic Materials C. Tech. Spec. 3.3 by Robert Bolt & James Carroll, Academic Press, 1963 OI M OOS
(" [)\
% %/ J Facility: Oyster Creek Sheet D cf unit: System Component Evaluat. ion Work Sheet Envituamont Documentation Ref. Qualification outstanding Equipment Desen,pt,orii Method fierns Pararneter Specification Qualification Specification Qualification Systein: Isolation Operating 4 hours See Note A See Note C Sec Note B See Chpt.7 Condenser Svstem Tinie Item 21 Plant ID No: IB-ll-Al Temperature See Note A See Note E See Chpt.7
,3
- I Item 21 Component: Isolation ( *F)
Condenser AP Switch P'essue See Chpt.7 Manufacture: Barton 16 See N te A See Note B Item 21 (PSIAI Model Number: 288A fletative Ilunu.dity See Note A See Note B See Chpt.7 (%) 100 1 Item 21 Function: Chernical Accuracy Not Not Not - --- Spray Spec: Applicable Required Required Itadiation 3.9x10 R lx10 R 1 analysis 3 q Service: RK-03 Aging 40 years See Note A See Note B See Chpt. 7 1}}