ML20004B614
| ML20004B614 | |
| Person / Time | |
|---|---|
| Site: | Shoreham File:Long Island Lighting Company icon.png |
| Issue date: | 05/27/1981 |
| From: | Novarro J LONG ISLAND LIGHTING CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| SNRC-577, NUDOCS 8105290173 | |
| Download: ML20004B614 (89) | |
Text
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LONG ISLAND LIGHTING COMPANY FACO www SHORF. HAM NUCLEAR POWER STATION P.O. BOX 618. NORTM COUNTRY RJ AD
- WADING RIVER. N.Y.11'92 SNRC-577 i?
4, May 27, 1981 g"fp 6 }
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Mr. Harold R. Denton, Director g Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission 'Y 20555 N I Washington, D.C.
Shoreham Nuclear Power Station - Unit 1 Docket No. 50-322
Dear Mr. Denton:
Forwarded herewith are sixty (60) copies of LILCO's responses to the Safety Evaluation Report (SER) Outstanding Issues listed in Attachment 1.
Please note that our responses to the Outstanding Issues listed in Attachment 2 will be forwarded to you under separate cover oy June 1, 1981.
Very truly yours, O
fl/TAJ&
L/J. P. NoVarro Project Manager Shoreham Nuclear Power Station g c'
CC:me 90 Enclosures cc: J. Higgins 8105290U3 i E
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Attachment 1 - SER Outstanding Issues Number Issue 1 Pool Dynamic Loads 4 Piping Vibration Test Program - Safety related snubbers 5 LOCA Loadings on Reactor Vessel Supports and Internals ,
6 Downcomer Fatigue Analysis ;
7 Piping Functional Capability Criteria -i 15 , Inservice Testing of Pumps and Valves 16 -
Leak Testing of Pressure Isolation Valves .
17 SRV Surveillance Program 24 Appendix H-II.C.3.b - Surveillance Capsules 26 . Suppression Pool Bypass 27 Steam Condensation Downcomer Lateral Loads 28 Steam Condensation Oscillation and Chugging Loads 29 Quencher Air Clearing Load 30 Drywell Pressure History 31 Impact Loads on Grating 32 Steam Condensation Submerged Drag Loads 33 Pool Temperature Limit 34 Quencher Arm and Tie-Down Loads 39 Emergency Procedures 47 Control System Failures 48 High Energy Line Breaks 55 Q-List 59 Control of Heavy Loads 60 Station Blackout l
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ATTACHMENT 2 - SER OUTSTANDING ISSUES
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SCHEDULED FOR SUBMITTAL BY JUNE 1, 1981 Number Issue 9 Environmental Qualification ;
20 Appendix G-IV.A.2.a - Nil Ductility Temperature 21 Appendix G-IV.A.2.c - Pressure Temperature Limit 22 Appendix G - Impact Testing 23 Appendix G-IV.B - Minimum Upper Shelf Energy 35 Containment Isolation 37 Secondary Containment Bypass Leakage 38 Fracture Prevention of Containment Pressure Boundary 51 Fracture Toughness of Steam Line and Feedwater Materials
! 57 TMI-2 Requirements (. Third .ind final submittal scheduled for 5/30/81) 58
- Reactor Vessel Materials Toughness l
- Based upon discussions with J. N. Wilson, NRC Project Manager, l Shoreham, it is our understanding that Outstaading Issue Number l 58, Reactor Vessel Materials Toughness, does not require a LILCO submittal.
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SNPS Item 1 - Pool Dynamic Loads Shoreham has followed all NUREG requirements with respect to pool dynamic loads with the excep'. ions noted below. These requirements are outlined in NUREG-0487 together with Supplement 1, Supplement 2 and NUREG-0484 Rev. 1. Deviations from these requirements yet to be approved are:
1.* In lieu of Supplement 2 of NUREG-0487, Shoreham has used an interim confirmatory chugging load definition whose power spectral density (PSD) essentially bounds the lead plants interim load definition. Please refer to our response to SER Open Item 28. In addition, Shoreham has committed to perform a subsequent design evaluation step using the final generic load.
- 2. To calculate quencher air clearing loads, Shoreham has used a generic load definition based on actual full scale T-quencher test results. Please refer to our response to SER Open Item 29. It is our understanding that review of this generic load definition has been completed and official NRC approval is imminent.
- 3. Shoreham has committed to use measured T-quencher arm and tie-down loads in those areas where Karlstein test data exceeds the generic T-quencher load specification proposed in the PP&L Design Assessment Report. Shoreham will commit to increase the quencher support bending moment used in the design assessment by a factor of 1.25. Please refer to our response to SER Open Item 34.
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SNPS Item # 4 - Safety Related Snubbers Prior to installation, all safety related snubbers sre opera-tionally stroked to determine that they are not frozen, seized or jammed. After installation and prior to system pre-operational testing, snubbers are visually inspected for signs of damage or impaired operability as well as for adequate swing clearances.
In addition, their location, orientation, position setting and configuration are verified to be in accordance with approved design drawings. Fluid level and leakage are not applicable to Shoreham as hydraulic snubbers are not employed in an1 safety related systems.
The snubber thermal movements for those systems that have operating temperatures in excess of 2500F will be verified during the Startup Test phase under STP-811 " System Thermal Expansion Test Procedure".
This includes verification of the expected snubber thermal move-ments and swing clearances. Discrepancies or inconsistencies will be dispositioned prior to proceeding to the next test plateau. A preliminary list of the snubbers is attached. The final list will be included in the station technical specifications.
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SNPS Item 5 - LOCA Loadings on Reactor Vessel Supports and Internals This open item deals with the effects of pressurization of the annular space between the shield wall and the reactor vessel and with the integrity of the reactor vessel support system and internals when subjected to the resultant blowdown reaction forces. SER section 6.2.1.6 identifies the need for sufficient information to assess the effects of annulus pressurization on the reactor vessel support skirt. Shoreham contends that all necessary information has been submitted in a revised response to NRC question 041.1 dated September, 1978, and that this analysis is in fundamental agreement with the requirements of NUREG-0609. A nodalization sensitivity study, designed to maximize the asymmetric load on the reactor vessel "to show that conservative loads are used in assessing the design of the ... vessel support system" is described in FSAR Section 6.2.1.1.4.3 with results presented in FSAR Section 6.2.1.3.5.3 and Figure 6.2.1 - 34. Moreover, the analysis described in FSAR Section 6.2.1.3.5 has already been reviewed by NRC -CSB and the results are in substantial agreement with NRC - CSB predictionr for Shoreham. NRC - CSB. predicts a pressure in the break node approximately 6 percent greater than that predicted by Shoreham, b.t the integrated effect on the asymmetric load is less. A 6 percent increase in the break node pressure will increase the maximum asymmetric load by only 1.5 percent. Therefore, the load definition does not appear to be a significant issue.
In addition, Shoreham's response to Staff question 112.21 has been provided in SNRC 566, dated May, 1981.
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SNPS Item 6 - Downcomer Fatigue Analysis j The fatigue analysis of the ASME class 2 and 3 downcomers and !
safety relief valve discharge piping in the wetwell is being completed by Shoreham using ASME Class 1 fatigue rules.
In DAR Rev. 4, Shoreham has committed to perform the fatigue analysis and to provide documentation of the results of the analysis by November, 1981 (DAR Rev. 5). We are prepared to review the preliminary results of this effort with the Staff on June 3, 1981 and to provide a letter submittal of the final results in August, 1981 well within SER Supplement 2 time frame.
It is Shoreham's opinion that this effort should be treated as part of the long term confirmatory program since the commitment has been made to perform the analysis. This opinion is consis-tent with many other requirements called for in the Mark II containm:ent acceptance criteria and is particularly appropriate for a fatigue evaluation where potential problems, if any, would develop only after many years of plant operation.
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SNPS Item 7 - Piping Functional Capability Criteria In Section 3.9.2.1 of the Shoreham SER, the concern is raised that the Shoreham basis for assuring functional capability may not be the same as that approved by the NRC staff (NEDE-21985).
Shoreham will amend the summary statement in DAR Revision 4, Appendix E (Section El.0) to read as follows:
El.C
SUMMARY
This Appendix provides " Functional Capability Criteria" for i evaluation of essential piping in Mark II nuclear power l plants. The criteria were established so as to be conservative and to assist in assuring maximum reliability of the piping considering all aspects of design, fabrication, in-service inspection, and operation.
The criteria are contained in pages E-5 and 6. The criteria are structured to make maximum use of the equations and definitions contained in the Code 2 . However, the functional capability criteria are not intended to substitute for or supersede any requirement of the Code.
The basis for the criteria is described in pages E-7 through 16.
The criteria are based, in large part, on the conservative approach contained in NUREG/CR-0261 2 ; i.e. on the single-hinge, limit moment concept with little or no consideration of strain hardening or dynamic effects. Recommendations or concepts given in NUREG/CR-OR61 for B indices are used. For elbows with 4,<90*, excess conservatism has been avoided by using a right-hand-side limit of 1.5Sy or 2.0Sy rather than the less applicable factors on S, or S3 as used in the Code for A, B, C, or D limits.
For Do/t >50, the allowable moments are decreased by increasing the B2 indices and equivalents of (0.751). This
- is based on test data on straight pipe at room temperature
- with, for ferritic materials, a temperature factor based on l
ratios of allowable longitudinal compressive stresses from Reference 1.
Dynamic effects may make the criteria very conservative when used for conditions where the loadings are dynamic in nature.
Shoreham hereby specifies that the functional capability criteria outlined in this appendix is equivalent to that presented in NEDE-21985 (Reference 6 approved by Reference 7).
Shoreham will also amend the last sentence of DAR Revision 4, Section 9.1.1.2.4 to remove the word " representative". This correction will reflect the current status that all Shoreham essential systems meet the functional capability requirements.
/ h nutaber or valves actuated are determinea tor typical piping components.
Tnese stress values are used to convert the number or stress cycles f or various numoers or valves actuated to equivalent all valve a ctua tion stress cycles by using an equivalent ratigue '
damage tormula. l The equivalent ratigue damage f ormula is constructed such that an l Identical ratigue vrage raetor would be ootainea rrom ASME i Section III(*), Dvsign Eatigue Curve, with tne proper i consideration of alt ernating stresses at a p3 ping component.
Furthermore, a conservative number of stress cycles per SRV actuation is usec to ontain the appropriate total number of stress cycles Ior the SRV load.
For tne Snoreham plant, this has been determined to be 4,050 stress cycles at the level of an all valve actuation ann 45,4U0 stress cyci n ct tne level or four valve actuation.
9.1.1.2.4 Functional Canan111rv Tne piping systems which are required to sately shut down the reactor and maintain its shutcown conditions as well as functioning to prevent or mitigate tne consequences of LOCA are classiileo as essentral systems. The other systems which do not have to perform the essential functions are classiriec as nonessential systers.
Tne basis for functional capability evaluation c: ens essential ~
piping systems is " Functional Capability Criteria for Mark-II Piping" (Appendix E) . Tne criteria are also outlined in Section 2 or this .recort. Evaluations of functional- capability of the essential piping systems are to assure tne fluid-flow capability of the piping systems. The cap;3ility will De maintained if sionificant recuctions in cross-sectional area do not occur under any service condition. l Equation 9 of ASME Section III, NB-3652 is used, witn medi:1 cations in accordance with Appendix E, f or this evaluation.
Tne results incicate that D noreham W ve essential systems meet the tunctional capanility requirements.
9,1.1.3 Results !
9.1.1.3.1 Summarv of hesults A reevaluation of all Snoreham reactor builcing piping and supports nas been completec, accounting for the ertects or tne nydrodynamic SRV and LOCa loads. All load comninations and acceptance criteria described in Section 2 were addressed. The rollowing tnree load combinations were f ound to be potentially controlling:
9-5 Revision 4 --February 1981 p j y _. --.w -
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SNPS Item 15 - Inservice Inspection of Pumps & Valves As stated in LILCO response to Item MEB-7 dated March 26, 1981 (LILCO letter to H. Denton from J. P. Novarro, SNRC-548), the pump and valve operability program described in that response is currently being reviewed. The ISI plan, as required by 10CFR50, will be submitted by January 1, 1982. Therefore, this -
request is premature.
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SNPS Item 16 - Leak Testing of Pressure Isolation Valves The periodic leak testing of pressure isolation valves will be conducted in conjunction with the Appendix J. Type C test program, the ASME code requirements and the technical specification surveillance ;
requirements. Pressure isolation valves, as described herein, are '
defined as those redundant valves within the Class 1 piping boundary that form an interface between the reactor pressure vessel and a low pressure system (RHR, CORE SPRAY, LPCI).1 These low pressure systems are also protected by valve position indication, pressure indication and relief valves, thus long term leakage will be detected and the postulated LOCA is not likely to occur.
Specific valves have been identified as pressure isolation valves and given the appropriate ASME Section XI category of A or AC.
This information is provided in Table 1. In each case, the pressure isolation valve is also a containment isolation valve. Leakage tests for each of the pressure isolation valves are presented in Table 2.
As described in FSAR response 212.106, these valves were grouped according to their leak tightness capabilities at reactor operating and reduced pressure conditions. Specifically,' reduced pressure leak tests (Appendix J. Type C) are proposed for pressure isolation valves for which pressure tends to enhance leak tightness (e.g., globe valve with pressure overseat). For tests at system pressure, the acceptance limit on leakage will be 1 gpm.
The frequency of testing is presented below:
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'a ) At least once per refueling outage approximately every 18 months.
b) Prior to returning the valve to service following maintenance,
- repair or replacement work on the internals of the valve which l requires full valve disassembly and repairs to seat and/or disc.
1 This situation does not exist on the RCIC system.
Note: ,
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TABLE 2 PRESSURE ISOLATION VALVE LEAK TESTS SHOREHAM NUCLEAR POWER STATION Valve Proposed Test Ell - MOV037 (A&B) Leak test at reduced pressure in conjunction with Appendix J, Type Ell - MOV047 C Test Program.
Ell - MOV048 Ell - MOV054 E21 - MOV033 (A&B)
Ell - MOV053 Leak test at Reactor System pressure Ell - AOV-081 (A&B) at frequency indicated.Ac'ceptance Ell - MOV-081 (A&B) Value 1 gpm E21 - AOV081 (A&B)
E21 - MOV081 (A&B) l s'
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+ i t PRESSURE ISOLATION VALVE DESCRIPTICN AND }
'! CATEGORIZATION SIIOREllAM NUCLEAR POWER STATION SYSTEM: Core Spray (E21) 4, ASME
- Section XI Size Valve Actuator Normal
] i Valve Description Category Class (in) Type Type Position AOVO81 A Testable Check on AC 1 10 C AO C AOVO81 B Core Spray Discharge AC 1 10 C AO C l' MOVO33 A Outboard Isolation A 1 10 GT MO C l MOVO33 B Valve on Core Spray T. 1 10 GT MO C Discharge vessel MOVO81 A Bypass on A 1 2 GL MO C MOVO81 B Testable Check A 1 2 GL MO C li SYSTEM: Residual lleat Removal
- & LPCI (Ell)
MOVO37 A LPCI Injection A 1 24 GT MO C MOVO37 B A 1 24 GT MO C i
^
MOVO47 RIIR Suction from RPV A 1 20 GT MO C MOVO48 A , 1 20 GT MO C AOVO81 A LPCI Injection AC 1 24 C AO C AOVO81 B Testable Check AC 1 24 C AO C MOVO81 A Bypass on LPCI A 1 1 GL MO C i
,' MOVO81 B Injection Testable A 1 1 GL MO C Check
- MOVO54 Ilead Spray (Inboard) A 1 4 GT MO C MOVO53 IIead Spray (Outboard) A 2 4 GT MO C I
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i SNPS Item #17 - SRV Surveillance Program 1.0 Purpose The purpose of the program is to accumulate information in sufficient detail to allow identification of generic safety / relief valve problems.
2.0 Scope This docu:aent defines the data records which will be used for monitoring performance of the Main Steam Line Safety /
Relief Valves (S/RVs) throughout the useful service life of each such valve.
2.1 The program is structured to collect sufficient data to allos the identification of safety / relief valve problems to minimize the possibility of a failure of the valve.
2.1.1 Data is also to be collected on inadvertent safety / relief valve operation, and on the failure of safety / relief valves to open or close.
2.2 For each safety / relief valve problem that is found to exist, the following information would be reported:
i) A description of the problem; ii) The operating conditions; -
iii) The failure mode (s) and the reason (s);
. iv) The remedial action l
l 3.0 Description of Program 3.1 Introduction 3.1.1 A MSL safety / relief valve data record will be maintained for the Shoreham SRV's.
3.1.2 The information identified in Section 3.2 through 3.5 will be recorded and maintained for each safety / relief valve.
3.1.3 The maintenance records will be updated with entries each time any work is donc on the valve (s) . This will include information regarding scheduled maintenance, and unscheduled maintenance, as described in Section 3.4.
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3.1.4 If subassemblies or components are interchanged for maintenance, the records will provide traceability for the components and the valve assembly.
3.1.5 The SRV maintenance history will be analyzed annually
, for the purpose of identifying potential trends which could lead to generic safety / relief problems, as des-cribed in Section 3.5.
3.2 Records Maintenance records will be maintained for each S/RV.
The record will, as a minimum, identify the following:
i) Manufacturer and model number; ii) Type and cize (including throat bore);
iii) ManufactIrer's serial number; iv) Set presuure stamped on nameplate; v) Date first installed on main steam line vi) Vessel hydrotest pressure after the valves have been installed; vii) Copy of original production test records; viii) Results of startup tests and any special tests; ix) Manufacturer's drawing and instruction manual references (GE-VPF) .
3.3 Scheduled Maintp. lance Rec crds 3.3.1 Identification of whether the complete valve o'r subassembly has been serviced and/or removed from the steam line.
3.3.2 Dates when equipment is:
i) Removed from service; ii) Reworked; iii) Retest ed; iv) Reinstalled in Service.
3.3.3 Operating history will be recorded for prior service cycle.
This will include the following information:
i) The number of power actuations, and date; ii) The number of pressure actuations, date and cause of actuations; iii) Leak detection device indications / signatures and history of these; iv) Other events; such as whether the steam lines were flooded for reactor shutdown; v) Ambient temperature, air and electrical supply condition.
l
3-3.3.4 Results of tests conducted prior to refurbishing, if applicable, will be recorded.
i) Performed where; ii) Performed by; iii) References used for cleaning, testing and refurbishing procedures; iv) Extent of disassembly performed; v) Details of any machining, rework or nondestructive examinations performed on components vi) Post reassembly bench test details wil) refer to test report number and attach summary of results.
3.4 Unscheduled Maintenance Records 3.4.1 The following information will be maintained for unscheduled maintenance.
3.4.1.1 The reason why a valve is being removed from service.
3.4.1.2 The following information will be recorded about the -
operating history of the valve:
i) Has the valve malfunctioned in the past? If yes, a brief summary of past history will be provided.
ii) Has this valve caused trouble in the past? If yes, ,
a'brief summary of past history will be provided.
3.4.1.3 Record the results of diagnostic tests if and when performed prior to disassembly.
3.5 Analysis of Data 3.5.1 The data accumulated on this program may be stored manually
, or by electronic data processing in a mant.er which gives' l flexibility in data retrieval.
t 3.5.2 A summary of cumulative failure rates will be maintained for predominant failure modes and for overall failures.
This summary will be updated annually.
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SNPS Item 24 - Appendix H - II.C.3.B - Surveillance Capsules This question 121.37 Paragraph 11.C.3B of Appendix H requires that four (4) surveillance capsules be included in the Surveillance Program. Provide technical justification for the fact that Shoreham Unit 1 Surveillance Program has three (3) rather than four (4) surveillance capsules.
Response
The Shoreham Surveillance Program was designed prior to the require-ments under 10CFR50 Appendix H, and three (3) Surveillance capsules were provif-d. Under 10CFR$0 January 1, 1980 Part 50, Appendix H,Section II .aragraph C3 Surveillance Program criteria Revised With-drawl Schedule four capsules are required with the fourth capsule indicated as standby.
For Shoreham, three (3) capsule supports are available on the reactor vessel and it is no longer advisable to perform additional welding on the reactor vessel. Test coupons are available and it is proposed that a fourth capsule will be installed when the first capsule is removed. This will serve to continue to provide a standby capsule should one be needed. The withdrawl schedule will be in compliance with 10CFR50, Appendix H Section II.
Additional justification for this surveillance program in Shoreham is based on the fact that the weld material, which is limiting, is also used in the LaSalle #2 Surveillance Program. (See Response to Shoreham Question 121.36). Thus, between Shoreham and LaSalle 1, a total of seven (7) capsules (four (4) on Shoreham, three (3) on LaSalle) will be irridated to study the effects on the properties of limiting material.
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1 Open Item 26 - Suppression Pool Bypass
- 1. With respect to the high pressure bypass leakage test, the way in which the vacuum breakers are to be included is presently under review by Shoreham and will be addressed separately. The acceptance criterion for the high pressure test given in SER Section 6.2.1.7 (10 percent of Afs/7 = 0.05 ft2) is unacceptabic. Shoreham has demonstrated that the allowable A/s/E for breaks capabic of producing large differential pressures on the drywell floor is at least an order of c gnitude larger thar. 0.05 f t: (refer to FSAR Figure 6.2.1-23B). In SSRC-318 dated September 18, 1978, it was further demonstrated that even with an acceptance criterien of 10 percent of the large break capability, it would be impossible to establish the 35 psi structural acceptance test differential pressure across the dryweIl floor with the size compressors to be used unless the leakage across the floor were within acceptance values.
Shoreham, therefore, considers the high pressure bypass leakage test to be a closed issue except for treatment of the vacuum breakers as noted above.
- 2. With respect to the actual small brea'i. capability, Shoreham has reviewed the possible reasons for the discrepancy between the NRC staff calculation (operator .
response time <15 minutes) and that done by Shoreham (operator response time = 26 minutes). It is Felieved that the model used by NRC-CSS is excessively conservative in the following ways:
- a. The NRC model uses a single control volume with an initial pressure of 30 psig. The operator response time is calculated from the time steam addition begins to the time the drywell design pressure is exceeded. The 30 psig used by NRC-CSB is approximately 6 psi higher than the wetwell pressure corresponding to complete air carryover for Shoreham (which, when exceeded, requires spray actuation). The effect of using the higher pressure is to shorten the availcSle operator response time by approximately 5 minutes.
- b. The NRC model uses an 8 percent revaporization j fraction which is based on large braak, single-I cha=ber containment data where the steam mole fraction exceeded that of the wetuall airspace during steam bypass. Since heat transfer is reduced by a relatively high air mole fraction, the quantity of condensate removed by the wetwell sinks during bypass is small compared to the large break, single-chamber containment data identified above.
One would expect, however, that the revaporization fraction of this smaller quantity of condensate would bc greater. Preliminary studies have shown virtuc11y no effect on crerator resr nse time for revaporization fractions between 23 and 100
- . . . - y
percent. Below 20 percent there. is a moderate decrease in operator response time with decreasing revaporization fraction. Shoreham's bypass version of LOCTVS uses an equilibrum treatment of condensate which corresponds to 100 percent revaporization. Use of 8 percent revaporization shortens the available operator response time by an estimated 5-6 minutes.
- 3. With respect to the low pressure test acceptance criterion, Shoreham proposed a criterion of 20 percent of the small break bypass capability, but the NRC staff has remained adamant on a 10 percent criterion. The NRC position has the effect of introducing a 1,000 percent margin between the maximum expected response and the design capability of the plant. In view of this 10 percent acceptance criterion and the NRC position on revaporization fraction, Shoreham has performed a reanalysis of pool bypass which includes:
- drywell and wetwell heat sinks (8 percent revaporization) downcomer heat addition
- heat and mass transfer between the airspace and the pool 7he results of this analysis are presented in the following section.
Shoreham Bynass Reanalysis This analysis supercedes all previous bypass analyses for Shoreham. A revision to FSAR Section 6.2.1.3.6 and the response to NRC question 041.32 will be made consistent with the following information.
The reanalysis of bypass for Shoreham has been performed with the Stone & Webster computer code CONSBA (Containment Small Break Analysis). The results of this reanalysis must be considered preliminary at this time since documention and qualification of CONSEA will not be complete until July, 1981. However, the reactor model is that of CONTORT, a fully qualified computer code developed by Stone & Webster for analysis of pool temperature transients due to safety / relief valve (SRV) discharge. It includes models for SRV operation, high and low pressure ECCS operation, and feedwater. It does not include a level swell or' bubble rise model and is, therefore, not suitable for prediccion of maximum containment pressure response to large breaks.
Modeling of high pressure ECCS and SRV operation permits calculation of long term depressurization/ pool hentup effects which could previously be approximated in LOCTVS only thrugh the use of a suppression pool heat addition curve.
A relatively simpic containment model has been combined with the CONTORT code to create CONSBA. The vent clearing and vent flow models are appropriate for the relatively small drywell floor differentici pressures which charactrize small (and even moderately large steam) breaks. Drywell and wetwall heat sinks are available with revaporization fraction an input variable.
_ . _ . _ ~ _ _ _ __ ___ - __
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Drywell and wetwell heat sink data are the same as that given in i Tabic 6.2.1-1 of the Shorcham TSAR. Heat and mass transfer between the airspace and the pool are modeled in the same aanner as that described for horizontal vent containments (Mark III) in Section 7.3.2 of SWECO C101 submittal by R.B. Bradbury (SSV) to J.R. Miller (NRC) on Mcrch 6, 1981 with the exception that the emissivity for radiant heat transfer from the airspace to the pool is set at 0:8 rather than 1.0. No explicit model has been included for downcomer heat addition, but the effect has been calculated iteratively and included by means of a net wetwell airspace heat cddition curve.
For consistency with SBA pool temperature transients presented in Section 10 of the Shoreham Plant Design Assessment for Hydrodynamic Loads (DAR), feedwater is used as the makeup source rather than ECCS. This represents a significant conservatism for three reasons:
- 1. More energy in the reactor / containment system.
- 2. Higher pool icvel increasing vent sub=ergence and drywell floor differential pressure.
- 3. Higher pool icvel compressing the we well airspace.
The objective.of the reanalysis is to determine the caximum value of A/.,fE that will permit a spray delay time of 30 minutes (1,800 sec) when considering the worst case break size. In determining the critical break size, it is necessary to assume a value of A/s/E, If the value assumed is reasonably close to the final result, it is not necessary to repeat the iteration. A value of A/vlic 0.16 f t was 2
chosen to study the effects of break -
size since it approximates the final expected result of including heat and mass transfer from the airspace to -he pool. In performing this critical break size study, heat transfer from the downcomers to the airspace and the effect of miscellaneous steel heat sinks in the wetwell were ignored. Neither is a large effect and both must be included in the analysis by manual l calcultion and application of heat addition (or subtraction) l curves to the wetwell aispace state calculation.
Figure 26-1 provides the results of the break size study. Note that ther- is very littic variation in drywell pressure at 1,800 i
seconds with break size for a wide range of breaks. This is f primarily a consequence of het feedwater addition which tends to limit depressurization of the reactor at the end of the 2
transient. (For example, for the 1.0 ft break, the feedwater temperature at the end of the run is approximtely 317'F corresponding to a saturation pressure of approximtaly 86 psia).
The 1.0 ft2 break is considered the limiting case.
Figure 26-2 provides the reactor, drywell, and wetwell pressure for the 1.0 ft steam 2
break described above. In this figure, the effects of downcomer heat addition and wetwell =iscellaneous steel hect sinks have been included. As noted prcriously, the Shorcham centainment heat sinks are described in dr. ail in FSAR Table 6.2.1-1. It is assumed that the heat transfer rates to the
. -. - ~.. .--. .. -.
wetwell airspace from the downcomer and from the wetwell airspace to the miscellaneous steel heat sinks both decrease from t=0 to 1,600 sec. Also it is assumed that the miscellanecun steel is in equilibrium with the wetwell atmosphere at t=1,800 sec. The net result is uniform heat rate addition to the wetwell airspace.
In calculting heat t*ansfer rates from the downcomer to the wetwell airspace a convective heat transfer coefficient of 0.8 Btu /hr-ft2 *F was used and an emissivity of 0.9 was employed for radiation. An average wetwell temperature of 225'T was initially assumed and was verified in the final analysis.
Note that no thermal stratfication in the pool is considered.
Intermittant condensation at the downcomer vent during the relatiely low mass flow characteristic of small breaks has been shown in full scale tests to be an excellent mixing mechanism.
Temperatures at the pool surface are generally less than the mass average temperature of the pool.
Tables 26-1 through 26-6 provide mass and energy balance information for the Reactor Coolant, Suppression Pool, Drywell Atmosphere, Wetwell Atmosphere, Liquid on the Drywell Floor and the overall containment, respectively.
Shoreham considers the information presented above to be adequate -
for a complete review of steam bypass for Shoreham. Although the large conservatism of ignoring heat and mass transfer from the aispace to the pool has been extracted from previous analyses with 'the effect now considered explicity, there remains the following sources of conservatism:.
- 1. Feedwater addition
- 2. All-steam bypass e
- 3. Revaporization of wetwell heat sink condensate limited to 8 percent.
These conservatisms seem more than adequate in view of the 1,000 percent margin applied to the results.
In su==ary, Shoreha='s anslysis of pool bypass has shown the following:
- 1. Perfer=ing the d: rwell flocr structural acceptance test constitutes an acceptable high pressure stea= bypass test as long as the total compressor flow to be used is approxi-
=ately 5000 SCEM or less. Treat =ent of the vacuum breakers
(. exposure to test pressure) is to be addressed separately.
- 2. The ERC acceptance criterion for the low pressure test (10 percent of the largest ANE that vill per=it at least 30 minutes operator delay for =anual spray actuation at the verst case break size) is acceptable to Shoreha= as long as all relevant effects, including heat and = ass transfer fro the air space to the pool, are included in the analysis. For Shoreha=, the low percentofA//I=pressuretestacceptancecriteriavillbe10 0.16 ft2, l
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SNPS Item 27 - Steam Condensation Downcomer Lateral Loads As identified in DAR Section 4.2.5.1, shoreham has performed a static single vent, a dynamic single vent and a dynamic multivent analysis of downcomer lateral loads due to steam condensation with the bracing located at Elevation 27'9". In all cases the results are acceptable. NRC staff has not yet completed the review of the dynamic methodology proposed 5y the Mk II owners Group and used in the Shoreham dynamic analyses. Until a satisfactory review is complete, the only acceptable basis for downcomer lateral load assessment is that presented in NUREG-0487, the static load definition. Shoreham hereby commits to perform a static multivent analysis in accordance with NUREG-0487 with the results to be submitted prior to fuel load.in the event that a NRC approved dynamic multivent method is not available by January 1, 1982.
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SNPS Item 28 - Steam Condensation oscillation and Chugging Loads The condensation oscillation (CO) load definition used by Shoreham in itr confirmatory program is the Mk II program generic load. Based on PSD comparisons, this load bounds the interim CO load accepted by NRC in NUREG-0487, Supplement #2.
The chugging load definition used by Shoreham in its confirmatory program is the Mk II program generic load without averaging.
This non-averaged generic load provides additional conservatism as discussed below. A comparison between this load definition and the interim chugging load accepted by NRC in NUREG-0487, Supplement #2 is not as straightforward as that made for Co.
The primary reason for this is that generic chugging is a source load definition and wall pressures may vary somewhat from facility to facility or even from location to location within the same facility. Wall pressure time histories for Shoreham are just now being completed and PSDs are not available. Some ARS data is available for both the interim and the Shoreham confirma-tory load definitions, but in both cases, the data was generated for internal evaluations only and is incomplete. Comparisons of the limited ARS data available show the two loads to be comparable at high frequency, but the Shoreham confirmatory load bounding at low frequency. In the reactor building (secondary containment) where the majority of the piping and equipment is located, the Shoreham confirmatory load is generally bounding across the frequency range, and where it is exceeded, both loads are bounded by the design basis (DFFR 20 to 30 hz).
Because of the somewhat inconclusive nature of the limited ARS comparisons, Shoreham is submitting the following additional e information. Migure 28-1 shows a comparison of the PSD of the Shoreham confirmatory load definition in the JAERI-CRT facility with that of the accepted interim load. Two elevations are presented for the Shoreham confirmatory load: 3600 mm (vent exit plane) and 1800 mm. The Shoreham confirmatory PSD was generated with the same dephasing window as that used for plant application (50 msec), but instead of choosing the worst variance in 1,000 trials for individual chug start times, an " averaging" procedure was used (described in detail in Section 6.2 of the generic load definition report (NEDE-24302-P) to deliberately decrease the predicted wall pressures. The effect of this deliberate decrease in the predicted wall pressure loads is most pronounced at high frequencies.
Because of the decreased degree of conservatism in the method of application of the Shoreham confirmatory chugging load to the JAERI-CRT facility as compared to that used in the Shoreham evaluation, Shoreham considers the PSD comparison of JAERI-CRT shown in Figure 28-1 to be more than adequate for assessing the interim load effect on the Shoreham plant. An inspection of Figure 28-1 shows an exceedence of the Shoreham confirmatory by the interim only at approximately 2 hz. The reason for this exceedence is that the condensation event at t = 25.3 see in Run 26 was considered in the interim load definition to be a chug
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and was included in the chugging data base. The Mark II owners Group considers this event to be part of condensation oscillation because water did not enter the vent. The Mark II OG did include this event in the CO data base, where it is completely bounded by other CO events as shown in Figure 28-1.
A second concern expressed by the NRC staff is that when the final generic chugging load definition with averaging is approved, the Shoreham confirmatory load (without averaging) may not be completely bounding. This is because the averaging procedure brings seven additional chugs into the load defini-tion. Figure 28-2 shows that the PSD envelope of the 7 key chugs bounds the PSD of the 7 adjacent chugs except for slight exceedences at approximately 14 hz and 28 hz. It is evident that a load definition based solely on the key chugs is clearly conservative, even though averaging will result in some minor shifts in frequency content. The above demonstrates the con-servatism of the Shoreham "onfirmatory load definition without averaging.
In summary, in the interest of expediting closure of this open item, Shoreham will commit to a two phase approach. In Phase I, Shoreham will use the confirmatory basis described above (generic chugging without averaging) as an interim load in .
place of the interim load accepted by NRC-GIB in NUREG-0487, Supplement $2 to demonstrate design basis adequacy. In Phase II, Shoreham will evaluate the generic load definition (once accepted by NRC-GIB) against the load used in the interim evaluation. This is consistent with the position stated in NUREG-0487 that final __
loads will be used to confirm those used in the interim.
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SNPS Item 29 - Quencher Air Clearing Loads The Shoreham quencher air clearing load is that described in KWU Report R141/141/79/E " Application of the SSES - Test Measurement Results to the Overall Loading of the Suppression Chamber of the Shoreham Plant by Depressurization Processes -
Revision 1" dated October 23, 1979 with two exceptions. The
' first exception is that the ADS trace (Test 11.1) will employ frequency multiplier of 2.0 to 2.86 instead of 2.0 to 2.2.
The second exception is that Test 4.1.6 will employ an ampli-tude multiplier of 1.12 instead of 1.07.
The opening characteristics of the Shoreham safety / relief valves (SRV's) have been verified to be essentially the same as those tested at Karlstein (opening time 50: 20-30 msec). The configur-ation of the Shoreham SRV discharge line vacuum breakers is the same as that of Karlstein: two 6 inch valves in parallel.
Therefore, the data taken at Karlstein is applicable to Shoreham when corrected for pool geometry effects.
Actuation of ADS at Shoreham will not occur at a reactor pressure greater than 78 bar. In order to actuate ADS, low reactor cool-ant level condition must exist. Low reactor coolcnt level also causes a reactor SCRAM which immediately reduces reactor power to decay heat levels (e.g., 5 percent power approximately one minute after SCRAM from full power). Therefore, by the time sufficient coolant inventory could be lost from a small break to actuate ADS, decay heat would have decreased to a point where the first set of SRV's would be maintaining pressure at approxi-mately 1115 psig. This is less than the accumulator pressure for .
ADS Test 11.1. Therefore, the wall pressure loads for Test 11.1 need not be extrapolated for reactor pressure.
It is Shoreham's understanding that, subject to the above excep-tions and qualifications, the NRC staff has found this load definition acceptable and that official approvai is imminent.
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SNPS Item 30 - Drywell Pressure History (for pool swell)
Shoreham plant-unique drywell pressure history vs. generic (NEDM-10320 with Moody Slip - flow treatment of subcooled inventory) comparisons were provided to NRC-CSB at the ACRS meeting on April 28, 1981. These comparisons demonstrate that the Shoreham
! drywell pressure history is bounding and, therefore, acceptable.
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S!GS Item 31 - Impact Loads on Grating (supports)
The NRC impact load criterion provided in NUREG-0467 covers only flat and cylindrical targets. Shoreham employs wedges on certain platform supports in the pool swell zone to divert flow and reduce impact loads on the supports. The method used by Shoreham to calculate impact loads on wedges is identified in Appendix D of the Shoreham DAR Revision 4 (the response to NRC question 020.72). Shoreham is awaiting NRC review of this material.
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i SNPS Item 32 - Steam Condensation Submerged Drag Loads
, NRC-CSB has not yet reviewed the steam condensation submerged structure load methodology described in LILCO letter SNRC-445 dated November 7, 1979 and reiterated in Appendix K to Shoreham DAR Revision 4. Shoreham is awaiting NRC review of this i material.
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SNPS Item 33 Pool Temperature Limit (Review of suppression pool temperatures transients involving S3V discharge)
In Section 6.2.1.8(f) of the Shoreham SER, the NRC staff states
! that the review of the Shoreham suppression pool temperature transients involving SRV discharge is- still in progress. A meeting between representatives of Containment Systems Branch, 1
Generic Issues Branch and the Mark II OG Mass / Energy Subcommittee
) on March 17, 1981 clarified the generic versus plant-unique aspects of the NRC staff review of the pool temperature transients. Based on the results of that meeting, Shoreham is providing the following additional information to CSB to facilitate review of the plant-unique aspects of the analysis.
- 1. the transients are analyzed in accordance with " Assumptions for Use in Analyzing Mark II BWR Suppression Pool Temperature H Response to Plant Transients involving Safety / Relief Valve Discharge" dated March 24, 1980 with one exception: loss of offsite power is not assumed for SBA and isolation / SCRAM
, cases as described in IV.A.l. This assumption is conservative as discussed below. !
. 2. Feedwater is added non-mechanistically for all transients.
In all cases except SORV at power with main condenser available, steam line isolation i s- assumed to occur 3.5 seconds after the start of the transient to maximize heat addition to the pool. In reality, MSIV closure would rapidly terminate steam flow to the turbine driven faed pumps and feedwater flow to the vessel. In such a sequen e of events, feedwater could again begin to enter the vessel o0se reactor pressure fell below the shutoff head of the condensate pumps.
. This assumpt$ on is the basis for continuing to add feedwater.
However, the condensate pumps require availability of offsite power to continue operation, and therefore, availability of 7 offsite power in the conservative assumption.
- 4. Availability of drywell coolers is not assumed for SBA cases.
For SEA cases, pool cooling is suspended for 10 minutes when l
reactor pressure decreases to the permissive value for LPCI operation. The 10 minutes suspension allows for realignment of the RHR system to pool cooling mode.
! temperature cutoff for HPCI operation incorporated in the Shoreham design. Adequate HPCI pump NPSH has been verified for operation at the maximum pool temperature observed in the analyses for reacter pressure greater than 150 psia.
- 6. For Shoreham, no single active failure can result in the permanent loss of one loop of pool cooling and the ,
, simultaneous loss of shutdown cooling mode. A powcr supply I
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failure disabling the inboard letdown valve for pool cooling and one RER loop can be overcome by manual realignment of four valves in the " faulted loop" (see Figure 33-1). These valves are accessible for manual operation, and realignment can be accomplished within one hour. A one hour delay in bringing the second loop into operation would have a negligible effect on the final pool temperature. The seal cooler on the " swing" bus pump would not be 7perable in this configuration, but seal cooling is not required for pump flow temperatures less than 212*F which is satisfactory.for pool cooling mode. Passive failures (e.g. of the heat exchanger) are not postulated in the short term, and two heat exchanger operation is not required in tnt long term.
Shoreham has also implemented pressure switches in the SRV tailpipes to provide positive indication of an SRV lift.
- 8. The Shoreham main . condenser is described in detail in FSAR Section 10.4.1. Opernting procedures will be reviewed to ensure that the main condenser is identified as the preferred
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heat sink for any transient not leading to MSIV closure.
- 9. The Shoreham Main Condenser 'r Removal System, Steam Seal System, Turbine Bype.ss System . Circulating Water System are described in detail in FSAu Section 10.4.2 through 10.4.5 respectively.
The above information covers the topics identified in the March 17, 1981 meeting described above and together with DAR Section 10, Appendix I and Appendix J provides the necessary information for a review of the Shoreham suppression pool temperature transients due to SRV discharge.
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SNPS-1 FSAR Item 34 - Quencher Arm and Tie-down Leads This open item deals with the option exercised by Shoreham to use the T-quencher arm and tie-down load specification described in the PP&L Design Assessment report instead of the X-quencher method outlined in the DFFR. In actuality, Shoreham has used Karlstein test data in lieu of the generic PP&L specification where the test data was greater than the specified load. It is Shoreham's understanding that the NRC staff and its consultants have found the generic PP&L load specification acceptable with the exception of the quencher support and arm bending moments where the test data exceeded the load specified. Shoreham has used the measured quencher support and arm bending moments in its design, but to account for possible system variations, Shoreham will commit to increasing the measured quencher support bending moment by a factor of 1.25 for design assessment.
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- SNPS Item # 39 - Emergency Procedures Emergency procedures for ATh'S events will be developed and submitted to the NRC for review. Likewise, operating procedures for the primary to secondary containment leakage detection and return system will also be submitted to the NRC for review after their development.
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SNPS Item # 47 - Control Aystem Failure A review has been conducted in which it was determined that only Class lE systems are necessary to achieve cold shutdown. Failure of the power source to the reactor manual control system and other nonessential systems and components will not affect any essential equipment nor the ability to safely shutdown the plant.
It has been further determined that the loss of any one power source serving essential instrument and control systems will not affect the ability to achieve a cold shutdown since diverse equip-ment and systems are supplied from independent busses. A more detailed discussion of the above is included in our response to SER Open Item No. 46.
While the ability to achieve cold shutdown is not affected by the events described above, a failure of certain control systems may either directly or indirectly impact the characteristics and severity of anticipated transients. These control systems are categorized as follows:
Category A Directly involved in initial identification or .
detection of event.
. Reactor water level 8 trip e
Category B Directly and actively involved in event or its effects.
. Relief valve operation
. Bypass system operation
. Rod block monitor f Category C l
Indirectly cr passively involved in event or its effects.
. Reactor feedwater system
. Reactor turbine pressure regulator
. Recirculation flow controller l
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Item # 47 (cont'd)
Category D Not involved directly or indirectly in event or its effects.
. Instrumentatien or power electric busses
. Environmental control systems
. Component service water systems The impact of Category A and B failures is discussed below.
Category C & D equipment performance does not significantly impact event severity.
Reactor Water Level Trip (L 8)
If this trip system, an anticipator of level, pressure and heat sink problems fails then the turbine generator moisture / vibration monitor would initiate a turbine trip. In addition, the Reactor Protection System is a fully safety grade backup to the Level 8 trip. General Electric has generically analyzed the impact of Level 8 trip failure upon transient event severity. -
The results of these studies have been discussed in meetings with the NRC both generically and on plant specific dockets. It has been concluded that the delta MCPR impact consequence of the L8 trip faflure is sufficiently small to justify its continued use in tran-sient analyses.
l In order to provide further assurance of L8 trip operability an I addition will be made to the Shoreham Technical Specification to provide for formal surveillance.
Relief Valve Operation Should the relief function of the SRVs fail to operate, the valves would open automatically in the safety mode (fully safety grade).
l There is no difference in event impact between the relief and safety functions since the MCPR reaches its lowest values before opening of the SRVs.
Main Turbine Bypass System The main turbine steam bypass system provides a momentary relief function for certain events. T;A most limiting transient event which takes credit for the turbine bypass system is the feedwater controller failure. Analysis indicates that a delta MCPR increase of approximately 0.08 applies to the transient without a functioning main turbine bypass system. In light of bypass system reliability and the very low probability of feedwater controller failure, this delta MCPR increase is not considered to be large enough to justify l a change in the present Chapter 15 transient analyses.
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Item # 47 (cont'd)
In order to provide further assurance of turbine bypass operability, an addition will be made to the Shoreham Technical Specifications to provide for formal surveillance.
Rod Block Monitor The Reactor Manual Control System implements a rod block if an erroneous rod withdrawal is attempted. The rod withdrawal error transient is evaluated utilizing the mitigating effect of the Rod Block Monitor (RBM).
General Electric met with the NRC on January 22, 1981 to demonstrate that the RBM is highly reliable having many redundant and self-testing features and that credit for its operation should be allowed in tran-sient analyses.
The NRC indicated tentative approval of the design and transient analysis with the addition of periodic Technical Specification testing to assure system operability.
The Shoreham Technical Specifications will be amended to include this requirement.
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l SNPS Item # 48 - High Energy Line Breaks The Environmental Qualification Program is currently being implemented at Shoreham. It describes the program by which Class lE equipment will be qualified in accordance with NUREG-0588 to certain defined environmental limits. Those limits are established on the basis of analysis of worst case events, including high energy line breaks. However, nonsafety related equipment does not require environmental qualification. The failure of these components in an accident environment will not affect the ability to safely shut down the plant and do not result in consequences more severe than those of Chapter 15 analyses or beyond the capability of operators or safety systems. A more detailed discussion of the significance of control system failures is provided in our response to SER Open Items numbers 46 and 47.
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Item 55 - Q-List i
SECTION A
- 1. Biological shielding within Primary Containment, Reactor Building, Control Building.
Response: Table 3.2.1-1, Section XLII, Structures, has been modified to incorporate item 1. .
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- 2. Missile Barriers within Primary Containment, Reactor Building and Control Building.
Response: Table 3.2.1-1, Section XLIII, Structures, has been modified to incorporate the missile barriers within the Reactor Building and Control Building. There are no missile barriers located within the Primary Containment.
- 3. Combustible Gas Control System Response: Item 3 is part of the Primary Containment Atmosphere Control System. Item XXVIII of Table 3.2.1-1.
- 4. Engineered Safety Feature Actuation System
- Response
- The engineered safety systems and their actuation signals are previously called out in Table 3.2.1-1 as part of the
, system (s) in which they are located.
- 5. Sampling System Sa Containment isolation valves 5b Piping within containment isolation valves Response: For the Post-Accident Sampling System, the corresponding isolation valves (5a above) and the piping within the valves (5b above) are designated as part of the system (s) being sampled, thus they are previously discussed in Table 3.2.1-1.
- 6. Containment Spary Response: Containment spray is part of the 1WR system which is addressed in Section IX of Table 3.2.1-1.
- 7. Onsite Power Systems (Class lE) 7a Diesel generator package including auxiliaries...
Response: The diesel generator package, including auxiliaries, are addressed in Section XXVII, Onsite Power-Systems, Sub-section (a) Diesel Emergency Power System, item 6, diesel generators.
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- ______ __ ______ __________ -__ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ~ _ _ _ _ _ _ _ _ . _ _ _ _ _ ._ . _ _ _ _ _ _ _ _ _ _ _ _ _
1 7b 4160V switchgear 7c 480V load centers 7d 480V motor control centers 7e Instrument, control and power cables...
79 Transformers 71 Protective relays and control panels >
7j AC control power inverters 7k 120V AC vitial bus distribution equipment 71 Containment electrical penetration assemblies ,
7m Other cable penetrations (fire stops)
- Response: 7,b,c,d,e,g,1,j,k,1,m were previously covered in Table 3.2.1-1. Further clarifica' tion of item 7,b,c,d,e,g,1,j, k,1,m is provided in Table 3.2.1-1, in revised Section XXVII, Onsite Power Systems.
7f Conduit and cable trays and their supports...
Response: Item 7f are off the shelf hardware items, thus they will not be included in Table 3.2.1-1.
7h Valve operators have system designation, and are classified as part of, and along with, the system in which they fall.
- 8. DC Power Systems (Class lE) -
8a 125V batteries, battery charges, and distribution equipment 8b Cables 8d Battery racks .
8e Protective relays and control panels Response: Items 8,a,b,d,e are covered in the revised Section XXVII, Onsite Power Systems, Subsection (e) DC Power Systems.
8c Conduit and cable trays and their supports...
Response: Item 8c are co-mercial grade hardware items, thus they will not be included m Table 3.2.1-1.
- 9. Main Steam Isolation Valves Leakage Control Systems Response: Item 9 above has been incorporated in Table 3.2.1-1 see Section XXXIV.
- 10. Radiation Monitoring (fixed and portable)
- 11. Radioactivity Monitoring (fixed and portable)
- 12. Radioactiviry Sampling (air surface and liquid)
r Response: Items 10,11,12 were previously covered in Section VIII, Process Radions Monitors, and Section XXXIX, Area Radiation Monitoring System of Table 3.2.1-1.
- 13. Radioactive Contamination Measurement and Analysis
- 14. Personnel Monitoring Internal and External
- 15. Instrument Storage, Calibration and Maintenance
- 16. Decontamination
- 17. Respiratory Protection, Including Testing
- 18. Contamination Control Response: Items 13-18 are administrative requirements and will not be included in Table 3.2.1-1.
- 19. Radiation Shielding Response: Item 19 is similar to item 1 above and is incorporated in Table 3.2.1-1 Section XLII Structures.
- 20. Waterproof Doors to Safety-Related Buildin'gs .
Response: Section XLII of Table 3.2.1-1, Structures, has been modified to include the waterproof doors of the Control Building, Screenwell and the Diesel Fuel Pump House.
- 21. Site Grading Response: Section XLII of Table 3.2.1-1, Structures, has been
- modified to include that area adjacent to the intake canal called out in the Safety Evaluation Report pg. 2-24.
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- 22. Sediment Measurements in the Intake Canal Response: Item 22 will not be included in Table 3.2.1.-l since it is an administrative requircment. However, this item is a Technical Specification Requirement.
- 23. Meteorological Data Colllection Program Response: Item 23 will not be included in Table 3.2.1-1 since this is an administrative requirement.
- 24. Expendable and Consumable Items Necessary for the Functional
! Performance of Safety-Related Structures, Systems and Components.
Response: Item 24 will not be included in Table 3.2.1-1 since this
! is an administrative requirement.
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- 25. Safety-Related Masonry Walls Response: Section XLII of Table 3.2.1-1, Structures, has been modified to incorporate this item.
- 26. Measuring and Test Equipment Usad for Safety-Related Structures, Systems and Components.
Response: Item 26 is, an administrative requirement and will not be incorporated in Table 3.2.1-1.
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SECTION B
- 1. Primary containment atmospheric control system - The hydrogen recombiners and associated containment isolation valves and piping within containmant isolation valves should be under the controls of the operational QA program.
Response: This system has been OA qualified in Table 3.2.1-1, Section XXVIII, and all administrative QA responsibilities :
have been initiated. .
- 2. Reactor building closed loop cooling water system - The piping within the containment isolation valves should be under che controls of the operational QA program.
Response: This system has been QA qualified in Table 3.2.1-1, Section XXVIII, and all administrative QA responsibilities have been initiated.
- 3. Identify the safety-related instrumentation and control systems and components to the same scope and level of d.; tail provided in Chapter 7 of the F,SAR. .
Response: The instrumentation and control system components have a been classified in Table 3.2.1-1 as part of the systems in which they fall.
- 4. Clarify that charcoal filters are included in the building standby ventilation system and the control room ventilation system.
Response: Section XXIX, Reactor Building Standby Ventilation ~
Systems and Section XXXVIII, Miscellaneous Ventilating Systems of Table 3.2.1-1 have been modified to include item 4.
- 5. Clarify that the floodproofing of the seismic category I civil structures listed in item XLII of Table 3.2.1-1 meets the QA requirements of 10 CFR 50, Appendix B.
Response: Section XLII, Structures, has been modified to incorporate all items in 5 above.
- 6. Provide a section in Table 3.2.1-1 for Effluent Radiation Monitors.
Response: Section VIII, Process Radiation Monitors includes all airborne and effluent monitors.
- 7. Radwaste System tank, atmospheric, must meet Regulatory Guide 1.143.
Response: Section XVIII Radwaste System of Table 3.2.1-1 has been modified to incorporate item 7.
l 1 - __ . . - _ _
l ...
- l.
- 8. Ducting and Isolation Valves should be classified under ASME III-2.
Response: ASME III-2 does not cover ventilation ducting, thus, there is no impact on Table 3.2.1-1.
- 9. Provide a Section in Table 3.2.1-1 for ESF Filtration Systems.
J Response. Table 3.2.1-1 already classifies ESF filtration systems in the Reactor Building Standby Ventilation Systems,Section XXIX.
1 i
l
. . _ - . ~ .
SECTION C
- 1. Plant Safety-Parameter Display Console
Response
Table 3.2.1-1 modified to incorporate this item, see Section XLIV.
- 2. Reactor Coolant System Vents Response: Not applicable to Shoreham. No plant related change required.
- 3. Plant Shielding Response: Not applicable to Shoreham. No plant related changed required.
- 4. Post Accident Sampling Response: Table 3.2.1-1 modified to incorporate post accident, sampling system - Section XLV.
- 5. Valve Position Indication Response: Table 3.2.1-1 modified to incorporate valve position -
indication, Section XLVI.
- 6. Dedicated Hydrogen Penetration Response: Item 6 is previously covered in Section XLIII, Primary Containment Structure, item 3, Penetrations, in Table 3.2.1-1.
- 7. Containment Isolation Dependability Response: Containment isolation dependability la a system by system responsibility and the related systems are already classified and included in Table 3.2.1-1.
S. Accident Monitoring instrumentation Response: Incorporated into Table 3.2.1-1,Section XLVII, Accident Monitoring Instrumentation System.
- 9. Instrumentation for Detection for Inadequate Core-Cooling Response: Implementation of changes will be accomplished on a system by system basis, hence any modifications will be in Table 3.2.1-1 on a system by system basis.
i
[
- 12. Challenges to and Failure of Relief Valves Response: There is no hardware changes related to item 12, thus no modification to Table 3.2.1-1 is required.
- 13. ADS Actuation Response: T%ere is no hardware change related to this item, thus, no modification to Table 3.2.1-1 is required.
- 14. Restart of Qpre Spray and LPCI Response: There is no hardware change related to this item, thus, no modification to Table 3.2.1-1 is required.
- 15. RCIC Suction Response: There is no hardwa're change related to this item, thus, no modification to Table 3.2.1-l is required.
- 17. Power on Pump Seals Response: There is no hardware change related to this item, thus, mo modification to Table 3.2.1-1 is required.
- 18. Common Reference Levels Response: There is no hardware change related to this item, thus, no modification to Table 3.2.1-1 is required.
- 19. ADS Valves, Accumulators and Associated Equipment and Instru-mentation.
Response: There is no hardware change related to this item, thus, no modification to Table 3.2.1-1 is required.
- 20. Emergency Plans Response: Item 20 is an administrative requirement and has no impact on Table 3.2.1-1.
- 21. Emergency Support Facilities Response: Item 21 incorporated in Table 3.2.1-1, Section XXXIII, Permanent Emergency Support Facilities.
- 22. Inplant I2 Radiation Monitoring Response: Components of the Inplant I Radiation Monitoring System 2
that are used in conjuntion with the Procese Radiation Monitoring System are classified as part of the Process Radiation Monitoring System,Section VIII, of Table 3.2.1-1.
- 23. Control Room Habitability Response: Systems responsible for monitoring control room habitability are identified in Table 3.2.1-1.
l l
_ _ _ - . - ~ _ . - -- - -
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pni.v. i . 2.1-i (
EOlllPMENT CI.ASSIFICATION E
N l
C Quality (*al LILCOt*td U
i Scopetal Group Quality Purchau'8*3 Princi-cf C1.sssifi- Assurance Seismic (63 Order 3.a1883 l jf Principal Conax>nent t il Susiply incation(33 ca t le an Catmor.y Cateoory fut e= (b. t . . C.me,i..n t : s a n 4 1. Heactor System
- 1. heactor vessel GE PC A f I 2/47 PfME 111 A, IWs Q
I I 2/67 teinti r ut.
ASMF. 111 O
port. skitt A, 19 (. '.
] Einte r 6b I.
- 3. Reactor vessel CE PC A I I 12/b4 PSME III appurtenancer, F. , l '> u's pressure retaining Winteg t.u ,
tortionu !
! 88 CRL ha,using su[1 sorts GE PC -
1 I X j
- 5. Reactor lutersial GE PC A I I X f strinctutes, essji- !
. neered safety teatures 6 2/b7
- 6. Coa ts supiort stzuc- GE PC - 1 1 X i
tures
- 7. Ott.er internal struc-tures .
.. Shtond head & GE PC . - 1 I X
}
separator asseinbly I
. h. Dryers. GE PC -
II I X (9)
- 8. Control rods GE PC , I I X
- 9. Control tod drives GE ic -
1 I X
- 10. Power rouge detector GE PC la 1 I t.SH k. 114 -2 .
hardware
- 11. D.31 inssesublies GE PC -
I I X f
- 12. heactor vessel GE PC -
1 I Y. I stabiliztr
- 13. Reactor vessel star P It -
1 I X
. t r :ss 116 Fe. actor vessel in- GE PC -
II HA y sula tions .
- 11. Nucle.ar lloller Synst em
- 1. Vessels, inst r ionen- GE PC A 1 2 N t'l. 111-1 tation compdesising cham.hus s
- 2. vessels, air aceinu- P PC Is 1 1 F ;; tit iII-e 1.it ora 1 ot 2l =
SNPS-1 FSAlt M h aClC Taut.E 1.2.1-1 (Cot 4T *D1
- I Qualityt*aD LI!404*bt Purchaset** P inck-w Scopetan Groep Quality of Classifi- Assurance Seismict63 Orde r lule n 6 Principal Cm ponenttaa supply locationsa8 cation c,tegory category _ pa t e. n nh- co m ntn**D
- 3. Piping, relief valve P IC C I I ASHE III-3 discharge (including rams! wad and supports)
GE PC A I 11/69 Is31.1.0 g
- 4. Piping, main steam within outer I
Q isolation valve
- 5. Pipe supports, maist GE PC A I i 1/75 ASHE III O ste m within outer isolation valve g!
PC =- I I X l
- 6. Pipe whip restraints, P main steam ASME III-1 I
Isolation valves
- 8. Othcr primaty cool- P IC A I I A:2tE III-1 ant pretssure loundary piping within iso-1-tion valves
- 9. Piping, instrumen- P RIl D II HA See 140te Is tation beyond outer-
- anst isolation valves
- 12. Valves, feedwater P PC, idl A I I ASilE III-1 (10) isolation valves i
afhI within A ASME III-1
- 11. Valves, other, iso- P PC, Ril I I f;
- 1. Lion valves ,uid within
- 14. Valves, instrumen.- P R15 D II 14A See Natu ti tation beyond outermost isola-tion valves X
- 15. Electrical modules GE PC -
1 I with sately function ,
- 16. Cable, with safety P - -
I NA X function III. Fecinculation Syst.ne
- 1. Piping GE PC A I I 10/69 1> 11.1. 0
2 og 21 1:n viulian 4 - Fra tisas y 1977
i
. ; i' SNPS-1 YSAR ;
TAtiLL 3.2.1-1 tCONT*D1 Scopetaa Qualityt*al Group LILCOt*b8 Quality Purchase 468 Princa-d of (;1 an S1J i- Assurance Seismict&D Order paltF8 ,
Principal Onnr.onent488 Eupply I.ocition485 catnon cateinry_ tategory pa r er rode a onsi.en g sas ,,,,,,,
- 3. Pipe restraints, CE PC - I I
.j a
recarculation line
- 4. Ptanps
- 5. Valves GL GE PC PC A
A I I I I
11/bv 10/63 ASML Ille ASME VIII/stss spoo M
- 6. Pump motors GE PC -
II I 10/b9 X
- 7. Electrical sedules, GE ku, et -
I I X with saluty
+ tunctaon
- u. Cable with satety P - -
I HA X tunction IV. CRD llydrauh c SvStem
- 1. Valves, isolation, P PC, Idi A I I ASML 111-1 .
water return line
- 2. Valves, scram dis- GL Rb b 1 1 12/bb B31.1.0 charge volume i lines
- 3. Valves insert and P Rn b I I ASME 111-2 (11) witfulcaw lines
- 4. Valves, other P Rb D II NA b31.1.0
- 5. Paping, water return P PC A g
't I ALMr. 111-1 line withan isola-r.lon valves
- 6. Piping, scrami dis- P Ru 3 I I ASML 131-2 ch.arge volume lines
- 7. Piping, insert and P l C, RB is I I ASML III-2 withdraw lant:s
- u. Paping other P Ru D 11' NA b31.1.0
- 9. CRD pumps, tilters, t.E Ru D II HA 1970 K ,
and strainers
- 10. Ily traulac control GE Rh - I I speciaL (12) unit
- 11. Cable, with safety P - -
1 NA X iunction V. Lt.awaby Laquad Control M utem
- 2. Pump GE Ru b 1 1 te/b3 b31.1.0 a.
IIIS
- 3. Pump neotor GL Hu -
I I 12/b3 1
- 4. V.alves, explosivo GL 4t u b I I 1//6.7 %P P '8'11-1 3 ot 21 hevinton 19 - La sit e:: t.s c 19m0
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1 0
4 TADT.E 3.2.1-1 (CONTeng s
QualityO43 LILCOt*b3 Scopetal Group Quality Purchaset*3 Princi-4 ' i Principal Commonentt a 3 of gupp)_g_- LocatIonO3 catinn Classifi- Assurance Seismictos Order Catenory _ category Date paltT3 Cote Cossentst#3 g
ASME III-C em l 4. Pump motors CE RB -
I I 7/69 X
- 6. Valves, beyond P outermost isola-tion valves RB, R 1 X M
- 7. Electrical modules GE -
I with saf ety itsic-tion
- 8. Cable, with safety P - -
I HA X Q
function - h XI. HPCT Syatnu i 1. Piping, within P PC A I , I ASME III-1 outernest isola-
! tion valves
tion valves
- 3. Piping return test P O D , II NA B31.1.0 line to condensate _ ,
atorage tank laryond reactor insildis.g
- 4. Vacuum punto dis- P RD B I I ASHE III-2 cl6stge lino f . Punip GE RD h I I 6/69 D31.1.0/
ASHE III-C
- 7. Valves, return tcat P RB B I I Art!E III-2 h line to condensate utorage
- 8. Valves, other P BB B I I ASHE III-2 4
- 9. Turbine GE Rin . -
1 1 6/69 (lii)
- 10. Electrical pedules CE RD j d I I A with outely f unc- !
~
tion
- 11. Cat,1c, with saf ety P - -
I EA X iunction ---
{
f e b fe Hf 21 1.
-~- *
- ,,$_J ... ye
- s -
'O
,'~2'\ *4
- i. . . -f
\. , / ('. . -
- te1*T.-t Y:LA1
- A !
70015.112.1-1._(R'!3T1P1 E
ammmmmam Oisall' y(
- a D 1.ILC04
- bI scope 82) G a ou, pu.nlity Purchaset*n Princi-pal (73
@ l of Cla6J111- ASmtrance SelGm1CO) Order Date Code comu.wint s * *
- p h lgal coyponautt*D suinel y location 885 caticn tintenary_ (3 dory XII. >> Q C fys_t.gn g lt A I I ASilE III-1
- 1. Piping, within u:terroot inoja-P g
- i. ion valven I. I p1 ny , 1: ys ad P I:u u I I ASitE III-2 Cl
...r. ..
s . . i . ,-
t ient valtJa D31.1.0 P o ti II tan
- 3. Piping, rotursi test line to con-densate storage tank beyond re.tctor building ASf!E III-2
- 4. Vacutua pump din- P lin It I I
- . :n g.! 1i: e iktam v sitiui g m ia t<>
cont,s it.au aL 1 u31a- ,
t tusi v.ilunu S. Ivy GE hu in I I 6/69 183 1 . 1 . 0 /
ASME III-C A 1 ASt4E III-8 L. V i t .s.a , 1 esclat. ins P 10,* litt 1
.;*1 within P I:11 9 I I ASME III-2
- 1. 'h t v o , re: tut n test.
I line to condon eat e ntas090 AstlE III-2
- 8. Valv.ra, othar P 1:li si 1 1
- 9. 'Iuthi no GE Ill),
1 1 (./6 *l ( 1 84 )
j His - I X
- 10. 1.l tet ri c4 L vulules , GE j I
with auf ety f unction X
- 18. Cable, with salcty P - -
I tiA t unct ion XIII. Fuel Service tituigenent t X
- 1. Fuel preparatlimi GE im - i
.i chine X
- 2. General purpose GE pin -
1 1 gr.npple j i
l l
l ,..i
.I !
SNPS-1 FSAR M
r TABLE 3.2.1-1 (CONTepg l' I.
w hw Qualityt*al LIIrut*ht
"a me Scopetas Group , Quality Purchasetel Princi-Principal Cranswinent a n t al classifi-Supiity _ Incationtal cation Assurance Soissaicial order CategorL Category D.a t e ghalt**
Caxle Q>osmentsten FJ XIV. Reactor __ Vessel Service Ennisment
- 1. Steam line plugs Gt; RB - . I I X
I I X sling and head g strongtuck
- 3. Drywell head litting P RB -
I I X Q rig XV. In-Vessel Service Enulpment
- 1. Control rod grapple GE RD -
I I X XVI. Hefueling Equ a issent t
I I 4/71 AISC
II tm X drywell
II See Note X (15) l reactor cavity (15) '
II HA X stand XV11. Storage Equirrae nt
- 1. ruel storage racks GE Hu -
I I X
1 I X a.;e container
- 3. Spent tuel pool, P Ra -
1 I X dryer /sep. geol, hx ,
c.vity liners ;
XVIII. Ra.twast e Syst em ,
- 1. Tanks, atmunspheric P RW D II HA X Reg,l'as..lo /./f3
- 3. Piping, cont.nitument P PC 15 I I ASME III-2
- I '
isolation I
- 5. Piping, other P Pit,0 T.RW D II HA lill.1.0
- 6. Punips P Ril, RW D II HA X
- 7. Valves, 11nw contzul P kW D II IR 3t11.1.0 and f ilte*r F'/Stt*In Ill.l.o (46) bl. Valvec, ot her P tut, 3:W D II HA l l
8 ut 21 1:a v n:. a. ai 4. It ey I'977
TfNISlH0 H00d N\ *
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=
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M I T e.c o M i
M e MM MM 8 s M
e M.e .I. e.e M M CE M M M e ne oM me M oo Me MM
,l 8
-e a M M M M M *= M e we >> e e e= *
= me me M U a. g . e
cw e
-e ** 9 W gam
- WE M
E W e fa3 cme eW
- % h[ O M e e L3 W ici f.3 ECEE M L1 M bi M to M L1 L12 f.f{ mM MQ **f'*
WM MM I
he G.1aQO 4 4 4 4 m. "3 eC eC 4 M 2 = eC 4 4 4 a
e w
I. c e S t 4 e 4 he O 4 4 4 j
9 f'* \ \ N
- U Wt G.
h* a N N es ce a m *= ,= e-3 he 1
,g Ce O C.
=>
.ae U be e n R5 oc ng
= M M M d
= M MM gggyy M-MM Ma gg Mw 94 f4 Q f a g A U>
e >e 0 W i =M40
- Q -e se : M ,e M
- J G -e = = M M M me =e >.a w me me M a* =e M me me me == M ** M O 3 4 .9 u M e me c = 1 m , l= >3 Ce d 4 -
<- 8 g x
- ' .? n C
~
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w .Ie e I we C
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- t I ee a: Co W.. e O he see eq e' ".J G Q G in:
- e.:, ne w
=l C 3 C
s,
- 3. x
.c- e
- 44 9 *t =3 c f: C3 C C =c 3 3 c "3 = == "3 =
%= 8 e t gr. = a % Cr. M =E E a E a J.r. .f. a n n bl i
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- 3 Princi- (
of Clanniti- Assuranco Solonnic(88 Order pal (73 '
' prhtclpal ('rwegngjit t a ) Supply Loc.it i onens cation Sq ngory Citogory Dato code i n===nta E
- 8 XXIII. f aecaly.2 nobs
- 1. Elect.rical auxtules, a. P IJB - 1 I X win.h safoty func- 12 GE RIl - I I X
- t. l t -i s
- 2. C.ible, wit h safety P - - I tiA X i na.it ion X
h 8 J. 14 'te cimt anun GE Ru -
I 1 ,
- g. sel ,
XXIV *. Offgas Systene e, . , , .
- 1. Atmos'heric p glycol P T D II 14A X t.unks
- 2. Itent ext-hanger n P T is I I, llA ASilE VIII
- 1. P'H ug P T, I(W D 11 14A D31.1.0 t; . V..i n.>, flow P T, hil D I1 14A D11.1.0
, ro itr c1 te . V.15.ni, uth r P T, hW la II flA 1531.1.0
- 6. Ste.es jet stir P T 13 11 14 % ASitE VIII
, ej :: tors *
, 1. Cisa ca.al vennela P ItW In II IIA ASHE VIII U. Ik e. uabi neru P T is II 14A ASHE VIII
- 9. F i l *..r t. a. P PW D II 1.A ASilE VIII ,
IAY* E.dY.If*_* dil! t.T jjy:s tcia
- 1. Pig.t s;ig, Saf ety P tus ,0, P, it c I I ASHE III-3 related
- 2. Pipin.J, other P - la II t4A h31.1.0
- 3. Pua,*a P P c I I ASHE III-3 85 . Puna motora P P -
1 1 X s
- 5. Valves, luol.a t ion P P. E c I I ASHE III-3
- 6. v.stveu, other P T,0,P D II :A B31.1.0
- 7. Elei:t rie nt sioduluu, P It , P -
I I X i
ui t ti u.ticty itinct.iori .
I
- 8. Cattle, teltti ua tet y P - -
I I4A X it.nction ,
XXVI. Q.fpres emil Alt fiyr.t nis
- 1. Ve:::sein, .ac cuninl a - P PC, lus C I 1 ASHE. III-3 t .u s, support inig uisoty-tel.ited c'/:s t ems t
m of 2 !
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TAllI.F 3.2.1-1 (COW 'Pl.
l %
f Qualityt*al LILCot*bt l Scopetal Group Quality Purchaset** Princi- """" t principel Conconent t
- D of supply classif1-Incat ion t 3 3 c.ition Assurance Seismicss3 Order Cateoory Category Date pal (88 Cove Cohewnt s t
- 8 Q
me f
, ,' 3. Valvus in lines P PC, RB C 1 I ASHt III-3 1
. between accumula-I tors and safety-related systems Q
isolation
(, . Electrical modtales P PC, RB, R -
I I X with s.sfety faniction
- 7. Cables with safety P - -
I NA X function
- 8. Valves and piping, P - D II NA bit.1.0 otler
_ XX V// _ pnt,d<. Po">e Sys k m5 q, Di esel Dnergency Tuwer Systems .
- 1. Day tanks P k, O -
I I - P.SEE III-3
- 2. Piping, fuel oil P H, O -
I I ASME III-3 system
- 3. Valves, fuel oil P R, O -
I I , ASPIE III-3 system
) 4. Punip5, fuel oil P RO -
I I PSEE IIt .i sys t eem
- 5. Ptap motors, f uel P R, O -
1 I X oil system
- 6. Diesel-generators P R, O -
I I X
- 7. Electrical miodules P H, O -
1 I X witta saf ety ituac-tions
- 8. Cosble, with safety P H, O -
I HA X tunctions
- 9. Diesel fuel storage P - -
I I ASHt,111-3
- t. inks
- 10. Diesel air compressoas ? H -
I I X i
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TART-E 3.2.1-1 (CONT
- DI M i i
Qualityt?al LILoot*b3 t
, Scopesas Group Quality Purchamet*3 Princi-1 of Classifi- Assurance Seismict*S Order p. sitz) l Principal componenstM Supply Tocationaal ration C.itegory Category Date Code Co.een int e n ,
, XXVIII. Primary containment Atmospheric =
4 a:ontrol System l g j
- 1. Piping
- 2. Valves P
P R8 RB I
I I I
ASHE III-2 ASME III-2 O ,!
- 3. Fans P RB -
I I X
1 I X b1pers
- I 5. Electrical modules P RB -
I 1 X i wath safety j
iunctions 4.. Cables with safety P Q
i function I MA X k
XXIX . heactor Ruilding Standby ventflation Svetem
- 1. Ducting and isola- P RB -
I I X tion valves with safety function *
- b')
, 2. Blowers D
- N";P ha -
I I X
- 3. Unit coolers P RB -
1 I X
- 4. Chilled water system P RB, R -
(11) ;
I I X (18) )
- 5. Electrical modules P RB -
I I X '
with a safety function
- 6. Cable with a safety P - -
I NA X function ;
XXX. Primary containment Purge .
System l 1. Contaisonent isolation P PC,hD D I I ASME III-2 volves and associ4 ted piping .
- 2. All other components P kB -
II HA X i XXXI. Inwer conversion System i
- 1. Main steam piping P RO, T B I I ASME III-2 ;
laetween outennost isolation valves .
up to turbine stop v.ilves l a
fi i
12 of 21 Ite vi ss inn 18. - f.p.' i l 1974 I i
l
, \
(
i 4
SNPS-1 FSAR I
TABT.E 3.2.1-1 tcONT*D1 ,
I a
' M t Quality 4*al LILCOt*b3 i
Scopetal Group Quality Purchaset*3 Princi-of Classifi- Assurance Seismicts3 Order pa1873 supp1v category cateoory Dato code cuisur nt nis t principal e s..entaal Incationsal cation
- mem**
- 2. Main steam branch P T D I I ASME III-2 %
p 3 i
piping to 1st valve
' capable of timely actuation
- 3. Main turbine bypass P T D I I ASME III-2 Q .
piping up to by-pass valve
'l P T b I I ASME III-2
- 4. First valve that is l either normally closed or capable M .
- of automatic closure .
in branch pipinq connected to munin g !
steam and turbine
- bypass piping WA S t ecial (19)
- 5. Turbine stop valves, P T D II ,
turbine control A valves and turbine bypass valves Special (19)
T D II WA to . Main steam leads P l irem turbine con-4 trol valve to turbine casing RB, T D II HA B31.1.0 (10) :
- 7. Feedwater and con- P 1 densate system beyond 3rd isola-tion valve XXXII. co_n,densate Storace and l tr.insfer System
- 2. Piping, suction line to llPCI, PCIC B31.1.0
- 3. Piping & Valves other P O D II HA O D II NA X ,
- 4. Other components P l
XXXIII. g d . ..i s.w .. c [ t.n . i . . j - -
if c+ *'i n o. ,1 e ,, , .-. . . .. _ . .. .. -.. - .. -
) vfl 5 p .a.e , s,L l ,
l,)e2ssIlm.i.,)) ,
i i
13 of 21 Reviolon 16 - April 1979
i I
l l SNPS-1 ISAR k
=:c
!l TABT.E 3.2.1-1 (coffr*fu
, Qualitytea3 LILCO(*bt
' Scopeta Group Quality Purchaset*8 Princi- %
j
( Principal Crumponentant of supply
. Classifi-inca t ion E 8 8 cation Assurance Seismic 881 Grder Category Category pa1883 Code Conenent s t e n g
E._' t e XXx1v Ah m W .i i w b i,,,.i
. Q
'if \
M IXalnas cubal .
g 313 _ .. _ . _ . _ _ _ _ - . _ _ _ _ _ . . . . . _ , _ , . . . . . . _ _ _ . . _ _ . _ _ . _
O L<s
- XXXV. Mi(scellaneous cosnoonentswie m<d e-8 2. ait 44)_ _ . . _ _ . _ . _ _ Q
- 1. Reactor building polar crane P RB -
I I X k
, XXXVI. Reactor Building Closed
! Loop Cooling 14ater System
- 1. Pumps and heat P RB C I I ASHE 111-3 exchangers
isolation i
- 4. Pumps and piping P R, C D II HA B31.1.0 for notor generator NG set coolers piping, other P PC, RB D II NA D31.1.0
( /alves, other P PC, RB D II HA B31.1.0 XXXVII. plu.113 naut and Floor pipinage Systems
- 2. Pumps P RD,T,kW D II HA X
- 5. Cable, with a safety P - -
I NA X function
- 6. Piping, other P RD,T,RW D II HA B31.1.0
- 7. Valves, other P RD,T,RW D II HA D31.1.0 14 of 21 Revision ib - F.pril 1979
( .
I SNPS-1 FSAR ud f 4 '
TAft!.E 3.2.1-1 (cottf*D1 g Qualityt*al LILC04*bt Purchase 8O Princi-G ammmme Scope (a3 Group Quality of Classifi- Assurance Seisar.ic 4
- 3 Order pa18FD Principal componentes) suroly locationta3 cation Category Category Date Code Cramment e(s )
O XXXVIII. Miscellaneous Ventilation Systems E
I I O
- 1. Battery rooss II & V P R
- 2. Screenwell ptsuphouse P P -
1 1 X H&V
- 3. Relay and emergency P R - I I X switchgear H & V P R - I I X Q
- 4. Control room ai [J7,m q conditioning'
- 5. Diesel generat r rocan P R - 1 I X ventilation XXXIX. g ea Radiation Monitoring System
?.M.M ronga.*tA p (Qg3 , y y K XL. Leak Detection $? stem
- 1. Temperature element GE PC, RD - I I X
- 3. Dif f erential tem-perature switch
- 4. Differential flow GE rc, hD - I I X switch GE PC, RB -
I I X
- 5. Pressure switch
I I X sure switch GE PC, RS -
I I X
- 7. Differential flow sunsaer
- 8. Reactor building P RD .
- II HA X (21) floor drain sumps
- 9. Reactor building P Ru -
II See Note X (22) floor drain pumps (22) and piping 15 of 21 . Revision Ib - April 1979
i
. 1
.( 1 . i a
SNPS-1 FSAR TAh!.E 3.2.1-1 dCONTeng :2",,
C C5 Quality 4*al LILC0(*b3 Scope 4a3 Group Quality Purchase 8*3 Princi- M 3
of Clausifi- Assurance Seisinicts) Order pal (33 !
Principal Conux>nenta n s gupply focationtal cation Category _ Category Date Code Cnimientst*3 XLI. Fire Protection Systens
- 1. Water spray deluge P - -
II HA X -
systems
- 2. Sprinklers, carton dioxide systems P -
II NA X Q
- 3. Portable and wheeled P - -
II HA X b extinguishers XLII. civil Structures t
- 1. Reactor building P Ris -
1 I
- 2. Office and service P - -
11 NA building
- 3. Screenwell P P -
I I 86 . Control building P C - I I
- 5. Turbine building P P
T = -
II NA NA I
h r
ACI-310-71 ACJ-301-t.nL72 (23)
- 6. Intake Canal -
- 7. Discharge tunnel P - -
II HA AISC-70
- 8. Discharge pipe and P - -
II NA diftuser
- 9. Radwaste building P RW -
1 I
- 10. Auxiliary boiler and P - -
II Ith IM act building XLIII.
.Tnse L. aiu svt A o alnwnt y Lona 4/MedSt ruct ut e
- 1. Reinforced concrete P PC -
1 I ACI-301
- 2. Liner P PC -
1 I 8/70 -
(2 7.)
- 3. Penetrations P PC -
I I 8/70 h31."I,1969 .
- 4. Drywell head and P IC .
- I I 8/10 ASnE III-il !
drywell equi tsnent, Sununer 1969 l CRD removal and suppression chamber at.ccas hatches
- 5. Drywell personnel P PC - r I 8/70 ASME III-u hatch Winter 1969
- 6. Personnel hatch P PC -
I I ASHE III-HC f or daywell equip- Winter 1972 nwnt. hatch tuner-gency air lock)
- 7. Ik>wacoracra P PC b I I ASHC III-2 (25)
Winter 1972 16 of 21 lu vi sion th - April 1979
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- b)
Group Quality C3
, Principal consonentt a Claualfi- Purchase 8*8 Ptlaci-
' gn,piv_ focatfon an c n g t.ug Assurance gtngy Seismicts) Order catcoory pa10 3
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SNPS Item 59 - Control of Heavy Loads The information requested to satisfy SER Open Item No. 57, " Control of Heavy Loads" has been specifically delineated in NRC letters October 22, 1980 and March 2, 1981 from D. G. Eisenhut. In accordance with the schedule provided in the referenced generic letters, the information requested in paragraphs 1 and 2 will be submitted by June 22, 1981 for Sections 2.2 and 2.3. Section 2.4 is not applicable to SER's as outlined in the March 2, 1981 letter.
We do not believe that resolution of this generic issue should be carried as an open item on the Shoreham docket.
4 a
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i SNPS Item # 60 - Station Blackout In the very unlikely event that both offsite and onsite alternating current (AC) power is lost, boiling water reactors may use a combination of safety / relief valves and the Reactor Core Isolation Cooling (RCIC) system to remove core decay heat without reliance on AC power.
Emergency procedures will be developed to ensure safe opcration of the plant and restoration of AC power. In addition, operators will be trained to effectively deal with this event. In this light, LILCO intends to perform, as part of its low power test program, several tests verifying RCIC operability upon loss of AC power or other degraded electrical conditions. For more details, refer to our response to NUREG-0737 item I.G.1, " Training During Low Power Testing".
The procedures and most of the training described above will be completed prior to fuel load. Completion of training (low power testing) can not be accomplished until after fuel load (but prior to commercial operation).
A complete assessment of LILCO's planned facility procedures and training programs with respcet to this matter will be forwarded by June 5, 1981. This is in accordance with the letter from Darrel G. Eisenhut to all Licensees of Operating Nuclear Power Reactors and Applicants for Operating Licenses dated February 25, 1981 and reeiived by LILCO on March 6, 1981.
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