ML20128E312
| ML20128E312 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 05/24/1985 |
| From: | Mittl R Public Service Enterprise Group |
| To: | Butler W Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8505290345 | |
| Download: ML20128E312 (20) | |
Text
.
O O PS G Company Putsc Servce Electnc and Gas 80 Park Plaza, Newark, NJ 07101/ 201430-8217 MAlLING ADDRESS / P.O. Box 570, Newark, NJ 07101 Robert L. Mitti General Manager Nuclear Assurance and Regulation May 24, 1985 Director of Nuclear Reactor Regulation U.S.
Nuclear Regulatory Commission 7920 Norfolk Avenue Bethesda, MD 20814 Attention:
Mr. Walter Butler, Chief Licensing Branch 2 Division of Licensing Gentlemen:
SAFETY EVALUATION REPORT OPEN AND CONFIRMATORY ITEM STATUS HOPE CREEK GENERATING STATION DOCKET NO. 50-354 is a current list which provides a status of the open and confirmatory items identified in Sections 1.7 and 1.8 of the Safety Evaluation Report (SER).
Items iden-tified as " complete" are those for which PSE&G has provided responses and no confirmation of status has been received from the staff.
We will consider these items closed unless notified otherwise.
In order to permit timely resolution of items identified as " complete" which may not be resolved to the staf f's satisf action, please provide a specific description of the issue which remains to be resolved.
Enclosed for your review and approval (see Attachment 3) are the resolutions to the SER items listed in Attachment 2.
This information will be incorporated, as required, into Amendment 11 of the HCGS FS AR.
Should you have any questions or require any additional information on these items, please contact us.
Very truly yours, 8505290345 850524 PDR ADOCK 05000354
/-
E PDR Attachments The Energy People I
95 4312 (4M) 7 83
Director.of! Nuclear-Reactor ~ Regulation.
2.-
5/24/85 C.
- D.
H. Wagner USNRC Licensing Project Manager (w/ attach.)
A.
R. ~ Blough USNRC Senior Resident-Inspector (w/ attach.)
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M P84 154/04 1/2 a
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Date:
5/ 2dV85 ATTACHMENT 1 R.
L. Mitti to A.
Schwencer
'I te No.
Subject Status-ltr.-dated
- OI Riverborne Missiles Completed 1/31/85, 2/22/85, 5/8/85 OI-2 Equipment Qualification Partial Response 2/1/85, 2/20/85, 2/28/85, 3/1/85
- OI-3 Preservice Inspection Program Partial Response 2/14/85 & 3/19/85 OI-4 GDC 51 Compliance Completed 3/12/85 OI-52 Solid-State Logic Modules Open
- OI-6 Postaccident Monitoring Completed 5/14/85 Instr umentation OI-7 Minimum Separation Between Completed 4/4/85 Non-Class'IE Conduit and
-Class IE Cable Trays OI-8
. Control of Heavy Loads Closed 1/18/85 (SSER 1)
OI-9 Alternate and Safe Shutdown NRC Action OI-10 Delivery. of Diesel Generator Closed Amendment 8 Fuel Oil and Lube Oil (SSER 1)
OI-ll Filling of _ Key Management Open Positions 01-12 Training Program Items (a)
Initial Training Program Completed 1/7/85 (b)
Requalification Training Completed 12/28/84, 4/26/85-Program (Revised Program)
(c)
Replacement Training Completed
-1/7/85 Program (d)
TMI Issues I.A.2.1, Completed 1/7/85 I.A.3.1, and II.B.4 (e)
Nonlicensed Training Completed 1/7/85 Program OI-13' Emergency Dose Assessment Completed 1/7/85 Computer Model M P85 27/10 1-mr i
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.=
2 R.
L. Mittl to A. Schwencer Itsm No.
Subject Status ltr. dated OI-14 Procedures Generation Package Partial Response 1/28 /85 & 4/10/85 OI-15 Human Factors Engineering Completed 4/10/85
-C-1 Feedwater Isolation Check Open
' Valve Analysis C-2 Plant-unique Analysis Report Completed 1/8/85 & 1/31/85 C-3 Inservice Testing of Pumps and Open Valves 7C-4.
Fuel Assembly' Accelerations Completed Amendment 8' C-5 Fuel Assembly Liftoff Completed Amendment 8 C-6 Review of Stress Report Open C-7 Use of Code Cases Completed 12/17/84 s
C-8 Reactor Vessel Studs and Completed 5/24/85 Rev. l-Fastners
!C-9 Containment Depressurization NRC Review Analysis C-10 Reactor Pressure Vessel Shield NRC Review Annulus Analysis C-l l -
Drywell Head. Region. Pressure NRC Review Response Analysis C-12 Drywell-to-Wetwell Vacuum NRC Review Breaker-Loads C-13 Short-Term Feedwater System Complete 4/22/85 Analysis C-14' Loss-of-Coolant-Accident Completed 3/1/85
-Analysis C-15 Balance-of-Plant Testability Completed Amendment 8 Analysis C-16 Instrumentation Setpoints Completed 2/15/85 C-17 Isolation Devices Open M P85 27/10 2-mr i
r.
.9 '
3 R.
L. Mittl to A.
Schwencer Item No.
Subject-Status ltr. dated C-18~
' Regulatory Guide.l.75 NRC Review C Reactor Mode Switch NRC Review C-20 Engineered-Safety Features Open Reset-Controls e
C-21 High Pressure Coolant Injection Open
' Initiation C-22 IE Bulletin 79-27 Completed Amendment 8 C-23 Bypassed and Inoperable Status NRC Review Indication C-24 Logic for Low Pressure Coolant Open Injection-Interlock Circuitry 2 C-25 End-of-Cycle-Recirculation Pump Completed 3/1/85 Trip
'C-26 Multiple Control System Failures NRC Review C Relief Function of Safety / Relief Completed 2/15/85 Valves C-28
~ Main Steam Tunnel Flooding Completed 5/24/85 Analysis C-29 Cable Tray Separation Testing Completed 4/4/85 C-30 Use of Inverter as Isolation Completed 3/7/85 Device C-31' Core Damage Estimate Procedure Open C-32 Continuous Airborne Particulate Open Monitors
'C-33 Qualifications of Senior Radiation Open Protection Engineer C-34
.Onsite Instr ument Information Open
.C-35 Airborne Iodine Concentration Open Instruments M P85 27/10 3-mr i
r-4 R.
L. Mittl--to A..Schwencer-Itcm No.
S ub].ect Status ltr. dated-
' - C-36
' Emergency Plan. Items Partial Response 11/9/84, 1/16/85, 2/7/85 &
.4/4/85.-
' C-3 7.'
.TMI' Item.II.K.3.18 Partial Response 3/1/85 &
4/22/85 MP.85 27/10 3-mr
ATTACHMENT 2
' ITEM'NO SER SECTION
-SUBJECT
- C-8 5.3.1.5 Reactor vessel studs and fasteners
.- C-28 8.3.3.l'.4 Main steam tunnel' flooding-analysis R7D:mr-M P85 27/10 5-mr 3
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ATTACHMENT 3 M P85 27/10 6-mr
a SER ITEM NO. C-8.(SER SECTION 5.3.1.5)
' REACTOR VESSEL' STUDS AND FASTENERS The reactor vessel studs-and fasteners-satisfy most of the recommendations of RG 1.65, " Materials and Inspections for Reactor. Vessel Closure Studs."
The FSAR does.not-discuss the nondestructive examinations of the bolts and nuts, and
'the applicant needs to_ confirm that the Code-specified-inspections were performed.
This is a confirmatory issue.
RESPONSE
' FS AR. Section 5.3.1.7 has been rev ised to prov ide the infor-mation requested above.
The following NSSS item by item assessment of compliance to Regulatory Guide.l.65, Position C.2 is provided for clarification:-
o Purchase Specification 21A9340, Rev ision.1,. Para-graph 5.1.4a, requires testing ' to be ' done af ter heat treatment prior to threading.
o ASME III, 1968 plus' Winter 1969 Addenda Paragraph N-322 (required-by purchase spec.) requires examina-tion in accordance with SA-388.
o Purchase Specification 21A9340, Rev ision 1, Para-graph 5.3.4, requires magnetic particle or liquid penetrant exam to ASME III, Paragraph N-626 or N-627 be done on.the final surface af ter all forming and heat treabment.
Paragraph N-266 and N-267 are equiv alent to NB-2583.
o ASME III, 1968 plus Winter 1969 Addenda Paragraph N-325.4 is equivalent to ASME III, 1974, Paragraph NB-2585.
M P85 29/06 1-cag Rev. 1
a 24 85027734 8 HCGS FSAR g/33.
5.3.1.7 Reactor vessel Fasteners The reactor vessel closure head (flange) is fastened to tho' react c vessel shell flange by multiple sets of threaded stude and nuts.
The lower end of each stud is installed in a threaded Mole in the vessel shell flance.
A nut and washer are installed on the upper end of each stud.
The proper amount of preload can be applied to the studs by a sequential tensioning using
~ hydraulic tensioners.
The design and analysis of this area of the vessel is in full compliance with all ASME B&PV Code,Section III, Class I requirements.
The material for studs, nuts, and washers is SA-540 Grade 324.
The maxisus, reported ultimate tensile strength for the bolting material is less than the 170,000 psi. limitation in Regulatory Guide 1.65.
Also the Charpy impact test recommendations in Paragraph IV.A.4 of Appendix G to 10 CFR 50 were not specified in the vessel order since the order was placed prior to. issuance of Appendiz G to 10 CFR 50.
However, impact data from the certified materials report shows G ingsetm stegenst of that all bolting satorials have met the Append es. The, nondedrash sammeegdsiwss prope 87
- H esses ced a
f
//
- den wer*
A phosphate coating was applied to threaded areas of studs, nuts and bearing areas of nuts, and washers to act as a rust inhibitor i
and to assist in retaining lubricant on these surfaces.
5.3.1.8 SRP Rule Review l
5.3.1.8.1 Acceptance Criterion II.2 SRP 5.3.1 acceptance criterion II.2 requires that the reactor vessel and its appurtenances be fabricated and installed in accordance with ASME B&PV Code,Section III, Paragraph NS-4100.
The manufacturer or installer of such components is required to certify, by application of the appropriate Code symbol and completion of an appropriate data report in accordance with ASME B&PV Code,Section III, Paragraph NA-8000, that the materials used comply with the requirements of N5-2000, and that the fabrication or installation comply with the requirements of NS-4000.
The HCGS RPV and appurtenances were sanufactured in accordance with the 1968 edition of the ASME B&PV Code,Section III, which does not have NB-designated subarticles.
In light of RCGS's compliance with 1968 ASME B&PV Code,Section III, and information 5C4 #E#
5.3-11 Amendment 1 s.e
INSERT 1 TO PAGE 5.3-11 Regulatory Guide 1.65 defines acceptable materials and testing procedures with regard to reactor vessel closure and stud bolting for light-water-cooled reactors.
The design and analysis of these reactor vessel bolting materials is in
-full compliance with ASME Code Section III, 1968 Edition Class 1 requirements which do not have NB-designated subarticles.
In relationship to regulatory position C.2 of Regulatory
. Guide 1.65, the bolting materials were ultrasonically examined in accordance with ASME Section III, N-322 after final heat treatment and prior to threading.
The specified requirement for examination according to SA-388 was complied with.
Straight beam examination was performed on 100 percent of cylindrical surfaces, and from both ends of each stud.using a 3/4 maximum diameter transducer.
In addition to the Code required notch, the reference standard for the radial scan contains a 1/2-inch diameter flat bottom hole with a depth of 10 percent of thickness, and the end scan standard contains a 1/4 diameter flat bottom hole 1/2 inch deep.
.Also, angle beam examination was performed on the outer cylindrical surface in both axial and circumferential directions.
Any indication greater than the indication from the applicable calibration feature is unacceptable.
A distance-amplitude correction curve per N-325 was used for the longitudinal wave examination.
Surface examinations 1were performed on the studs and nuts after final heat treatment and threading, as specified in the regulatory guide, in accordance with N-626 or N-627 of the applicable ASME Code.
Specifications for ordering replacement / spare nuts and studs are in compliance with Regulatory Guide 1.65.
MP85 105/05 1-az
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'SER ITEM NO.I C-28 (SER SECTION 8.3.3.1.4)-
MAIN STEAM TUNNEL FLOODING ANALYSIS.
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By letter dated September' 28, 1984, the applicant provided
~
.the results of.a separation. analysis that addressed protec-tion.of.-all Class lE: equipment from' external: hazards.
On the basis of the results of this analysis,.the staff con -
cludes'that all_ Class 1E equipment i;s protected-from exter-
- nal. hazards,umeets GDC 2, 4, and 17, and is acceptable with
- one exception.
The one exception involves a. flooding hazard in the main steam tunnel.
A feedwater line break will cause
~ flooding and may'cause~ failure of some Class.lE motor-operated, valves and temperature elements.
For each of these items, the _ applicant has stated that. there is a primary and backup protective device located:in the hazard-free area.
The L applicant-has committed to perform an analysis to verify that', af ter both the primary l and backup protective devices open as a result of failure,of the nonprotected equipment together;with the worst case single failure, the plant can j
be safely shut down.
If the analysis shows the plant cannot be safely' shutdown'the applicant has committed to provide i.
. protection.
On_the basis of these commitments, the staff concludes that.the design meets the protection requirements of GDC'2, 4, and 17 and is acceptable.
This item is confirmatory pending receipt and review of the applicant's analysis.
Response
The' attached report entitled, " Main Steam Tunnel Flooding-Analysis," for; Hope Creek Generating Station, dated April 1985,- documents the above hazards analysis and concludes that none of the components which are flooded and.are not
. qualified for_ submergence are required for safe _ shutdown of
- the plant, nor will their failure : prevent safe shutdown.
Because_the_ equipment / systems that are required to safely shutdown-the plant.are_ single-failure proof, no single failure can prevent safe shutdown.
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MAIN STEAM TUNNEL FLOODING ANALYSIS i
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HOPE CREEK GENERATING STATION PUBLIC SERVICE ELECTRIC & GAS CO.
O APRI L 1985 PREPARED BY BECHTEL POWER CORP.
txinnrmmrumras 'Iv'ry PURPOSE
' f~')
The Hope Creek Safety Evaluation Report (S ER) requires submittal for NRC staf f review a hazards analysis for ef fect of a feedwater line break in the main steam tunnel (MST) on the plant's ability to safely shut down.
(SER Section 8.3.3.1.4 - Confirmatory Item No, 28)
METHODOLOGY The analysis was conducted in the following manner:
1.
Identify all Class lE equipment and components in the main steam tunnel, Room 4316, that will be subject to the worst case submergence which results from a break in a main feed-water line.
(Flood level is Elevation 126' of this roon).
2.
Identify the safety channel and safety function / system of each equipment or component identified.
3.
Determine if the equipment or camponent is qualified for subme rgence.
If not, determine if the equipment or compo-nent circuit has a primary and backup protective device located in a hazard-free area.
4.
Determine if the plant can be safely shut down af ter both
{m the primary and backup protective device open as a result of the failure of the unprotected equipment or canponent s
together with the worst case single failure.
ANALYSIS The Class 1E equipment and components which will be subjected to flooding are identified on the attached table.
This table also provides information on safety channel, safety function / system, submergence qualification and location of the primary and backup protective devices.
The evaluation of safe shutdown af ter loss of the equipment / components not qualified for submergence is as follows:
A.
Motor Operated Valves (MOVs) 1.
LAB-HV-F071 - Main Steam drain line isolation downstream of the outboard MSIVs.
This valve is not required to mitigate the consequences of a feedwater line break or any other pipe break which could cause flooding of the main steam tunnel, nor is it required for safe shutdown.
2.
1 KP-HV-5 8 29 A& B, -5834A&B, -5835A&B, -5836A&B, and 5837 A&B -
MSIV sealing system gas supply valves.
These valves are only required to mitigate the consequences of a LOCA.
Should any valve (s) spuriously open, upstream piping is O'
protected from overpressurization by check valves (See FS AR Figure 6.7-1).
These valves have no safe shutdown function.
K68/15-2
Afil 13 '86 6 8 2 d.I, 4 0 /
A.
Motor Operated Valves (MOVs) - Cont'd l
I 3.
l AE-HV-F0 39 - RWCU return to f eedwater.
This valve is powered f rom AC motor control center 10B242, Channel D.
The supply line to the RWCU system has a containment inboard isolation valve (1BG-HV-F001) powered from the Channel A source and an outboard isola-tion valve (IBG-HV-F004') powe red f rom 10B242.
Neither these valves nor their power supplies are located in the MST.
There is no single failure which can prevent both supply isolation valves f rom closing.
They will close automatically on low level in the RPV.
B.
Solenoid Operated Valves 1KL-PDV-582 5 A& B - MSIV sealing system supply valves.
See discussion of item A.2 above.
C.
The nnoccu ple s The thennoccuples provide input to the MSIV isolation logic which closes the MSIVs on high temperature in the MST.
Clos ure of the MSIVs is not required to mitigate the consequences of a feedwater line break or to safely shut down the reactor.
Closure of the MSIVs will not prevent safe shutdown of the reactor.
()
CONC LUSION As discussed above, none of the camponents which are flooded and are not qualified for submergence are required for safe shutdown of the plant, nor will their failure prevent safe shutdown.
Because the equipment / systems that are required to safely shut-down the plant are single-f ailure proof, no single f ailure can prevent safe shutdown.
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MAIN STEAM T1NNEL FIDODING ANALYSIS Page 2 of 5 CLASS lE EQUIPMENT AND CD@OiENIS ANALYZED EQUIP!ENT/
SAEETY SAFETY FUNCTIOW OUALIFIED FOR ILCATION & PRIMAHf/BMXUP COMRJNENT No.
1 CHANNEL SAFETY SYb" FEM SUBMERGENCE PIUfECTI\\E IEVICE 10C611 - Primary (1)
ISK-TE-N010B RPS X Main Steau Timnel high tenp.
No 10C411 - Backup (Thennoccuple) trip irput to NSSSWStean Leak Detection l
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10C611 - Primary (1) l 1SK-TE-N012B RPS X Main Stean Tunnel high tenp.
No 10C411 - Backup (Thennoccuple) trip irput to ESSWStean Led Detection l
10C609 - Primary (1)
'1SK-TE-N010C RPS Y Main Stean Ttnnel high tenp.
No 10C410 - Backup (Thennoccuple) trip irput to tESSWStean Leak Detection l.
10C609 - Primary (1)
ISK-TE-N012C RPS Y Main Steau Ttnnel high tenp.
No 10C410 - Backup (Thennoccuple) trip irput to tESSWStean Led Detection l
l 10C611 - Primary (1) 1SK-TE-N010D RPS Z Main Stean Timnel high tenp.
No 10C411 - Backup (Thennoccuple) trip irput to NSSSWStean Leak Detection l
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ISK-TE-N012D RPS Z Main Stean Ibnnel high tenp.
No 10C611 - Primary (Thermocouple) trip irput to tESSWstean 10C411 - Backup 11) i Leak Detection l
l 10B232 N 1AB-lW-F071 C
Main Stean lines downstrean (Motor cperated drain isolation i
valve) l l
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v Lj MAIN STEAM HJNNEL FLOODING ANALYSIS Page 4 of 5 CLASS lE EUJIPMENT AND COMPONENTS ANALYZED EQUIPENT/
SAFEIY bAEETf FUNCTIOt4/
QUALIFIED FOR LEATION CF--PRIMAlW/BAQ(UP CUMREENT No.
CHANNEL SAFETY SYb"1TM SUBMERGENCE PICIECTIVE DEVICE 1KL-PW-5825A D
EIV Inboard Seal Gas Supply No lYF404 (2) l (Solenoid Valve) i 1KP-W-5829A D
EIV Inboard Seal Gas Supply No 10B242 (2)
(Motor q>erated
valve) l l
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I IKP-W-5834A D
EIV Inboard Seal Gas Supply No 10B242 (2)
(Motor q>erated I valve) l IKP-W-5835A D
EIV Inboard Seal Gas Supply No 10B242 (2)
(Motor cperated I valve) l l
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l 1KP-W-5836A D
EIV Inboard Seal Gas Supply No 10B242 (2)
(Motor cperated l valve)
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1KP-W-5837A D
EIV Inboard Seal Gas Supply No 10B242 (2) y, (Motor cperated l 3
N valve l
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w 1AE-W-F039 D
NWCU Discharge to Feedwater No 108242 (2) a co (Motor q>erated !
Isolation M
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K68/15-7
3 MAIN STEAM 'I1NNEL FIDODING ANALYSIS Page 5 of 5 1
CLASS lE EQUIPMENT AND 00@ONENTS ANALYZED t
EQUIPENT/
SAFETY SAEETI FUNCTIOr4/
QUALIFIED FOR LEATION T PRIMAN/ BACKUP COMPONENT No.
CHANNEL SAFETY SYSTEM SUBMERGENCE PROTECTIVE DEVICE e
1&W-F019 D
Stem Lines Drain Outboard Yes 10B242 (3)
(Motor cperated Isolation valve) l 1
1 1&W-F067A D
Main Steau Line A Outboard Yes 10B242 (3)
(Motor cperated Drain valve l
l 1&W-F067B D
Main Steau Line B Outboard Yes 10B242 (3)
(Motor cperated l Drain valve) l l
1&W-FM -
D Main Stean Line C Outboard Yes 10B242 (3)
(Motor >
ve tW, !
Drain j
valve) j J
1&W-F067D D
Main Stean Line D Outboard Yes 10B242 (3)
(Motor cperated Drain valve EU i
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(1) Opening the backup protective device de-energizes the associated RPS diannel Wiich may result in a reactor trip. This is a safe condition.
3 N
(2) Opening the primary and backup protective devicesdoes not affect any canponent other co than the identified conponent.
w-
-.N (3) For qualified cperators, only gimary gotection cbvice location is govided.
CD 9
K68/15-8
-