ML19332E977: Difference between revisions
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GPU Nuclear Corporation N IO8pm u One Upp6r Pond Road Parsippany, New Jersey 07054 201 316-7000 4 | |||
GPU Nuclear Corporation | TELEX 136-482 Writer's Direct Dial Number: | ||
N IO8pm u One Upp6r Pond Road Parsippany, New Jersey 07054 201 316-7000 4 | |||
TELEX 136-482 | |||
Writer's Direct Dial Number: | |||
1^ | 1^ | ||
November 30, 1989 U.S. Nuclear Regulatory Commission C000-89-0786 Attention: Document Control Center Mail Station P1-137 Washington, DC 20555 Gentlemen: | November 30, 1989 U.S. Nuclear Regulatory Commission C000-89-0786 Attention: Document Control Center Mail Station P1-137 Washington, DC 20555 Gentlemen: | ||
| Line 41: | Line 34: | ||
6 $. / [M - | 6 $. / [M - | ||
. L. Long icePresiden[t & Director Planning & Nuclear Safety JL/cjg(89-21A) | . L. Long icePresiden[t & Director Planning & Nuclear Safety JL/cjg(89-21A) | ||
Attachment cc Mr. William T. Russell, Administrator | Attachment cc Mr. William T. Russell, Administrator Region I U.S. Nucloar Regulatory Commission | ||
Region I | |||
U.S. Nucloar Regulatory Commission | |||
: p. 475 Allendale Road g King of Prussia, PA 19406 bA NRC Resident Inspector | : p. 475 Allendale Road g King of Prussia, PA 19406 bA NRC Resident Inspector | ||
'O Oyster Creek Nuclear Generating Station Forked River, N.J. 08751 | 'O Oyster Creek Nuclear Generating Station Forked River, N.J. 08751 | ||
| Line 52: | Line 41: | ||
-y- Mail Station P1-137 g g Washington, D.C. 20555 g | -y- Mail Station P1-137 g g Washington, D.C. 20555 g | ||
.m44 g g GPU Nuclear Corporation is a subsidiary of General Public Utilities Corporation | .m44 g g GPU Nuclear Corporation is a subsidiary of General Public Utilities Corporation | ||
,. . w -. | ,. . w -. | ||
+ , | + , | ||
cq 7 7 ATTACHMENT OYSTER CREEK NUCLEAR GENERATING STATION UNRESOLVED SAFETY ISSUES' USI NUMBER TITLE APPLICABILITY' STATUS /DATE* REMARKS A-1 Water Hammer All C 3/89 See Note on page 3. | |||
cq 7 7 ATTACHMENT | |||
OYSTER CREEK NUCLEAR GENERATING STATION UNRESOLVED SAFETY ISSUES' | |||
USI NUMBER TITLE APPLICABILITY' STATUS /DATE* REMARKS A-1 Water Hammer All C 3/89 See Note on page 3. | |||
A-2 Asymmetric Blowdown PWR NA -Applies only to PWR Plants. | A-2 Asymmetric Blowdown PWR NA -Applies only to PWR Plants. | ||
Loads on Reactor Primary Coolant. Systems A-3 Westinghouse Steam W-PWR NA Applies only to Westinghouse Plants.~ | Loads on Reactor Primary Coolant. Systems A-3 Westinghouse Steam W-PWR NA Applies only to Westinghouse Plants.~ | ||
| Line 77: | Line 54: | ||
A-8 Mark II Containment Mark II-BWR NA Applies only.to Mark II Poc. Dynamic Loads Containments A-9 Anticipated Transients All C A diverse and independent Alternate Without Scram Rod Injection System installed in 1988. | A-8 Mark II Containment Mark II-BWR NA Applies only.to Mark II Poc. Dynamic Loads Containments A-9 Anticipated Transients All C A diverse and independent Alternate Without Scram Rod Injection System installed in 1988. | ||
Standby Liquid Control System modification completed in 1988. Recirculating Pump Trip installed in 1979. | Standby Liquid Control System modification completed in 1988. Recirculating Pump Trip installed in 1979. | ||
A-10 BWR Feedwater Nozzle BWR .C Committed'to continued inspections.. , | A-10 BWR Feedwater Nozzle BWR .C Committed'to continued inspections.. , | ||
Cracking of.the Feedwater and Control Rod Drive | Cracking of.the Feedwater and Control Rod Drive | ||
' Systems in accordance with NUREG-0619. | ' Systems in accordance with NUREG-0619. | ||
e (a9-21) | e (a9-21) | ||
E _ _ _ . _ _ _ . .. _u _ __ _a_ __ _ __ ___ . .-. _ _ . . | E _ _ _ . _ _ _ . .. _u _ __ _a_ __ _ __ ___ . .-. _ _ . . | ||
N | N | ||
[pd | [pd | ||
| Line 95: | Line 65: | ||
- ;w | - ;w | ||
~ | ~ | ||
USI NUMBER TITLE APPLICABIITY STATUS /DATK* REMARKS A-ll Reactor Vessel Material All C'3/88 Technical' Specification pressure-temperatur.4 operating limits meet both Appendix G t:e 10 CFR 50 and RG. .l.99, > | USI NUMBER TITLE APPLICABIITY STATUS /DATK* REMARKS A-ll Reactor Vessel Material All C'3/88 Technical' Specification pressure-temperatur.4 operating limits meet both Appendix G t:e 10 CFR 50 and RG. .l.99, > | ||
.' Rev. 2. License Amendment No. 120,1 dated 3/21/89. | .' Rev. 2. License Amendment No. 120,1 dated 3/21/89. | ||
| Line 101: | Line 70: | ||
Steam Generator and Reactor Coclant Pump Supports A-17 Systems Interactions All HC No Action Required by G.L. | Steam Generator and Reactor Coclant Pump Supports A-17 Systems Interactions All HC No Action Required by G.L. | ||
A-24 Qualification of Class All C 11/30/85 In compliance with 10 CFR 50.49 lE Safety-Related Equipment A-26 Reactor Vessel Pressure PWR NA Applies only to PWR Plants. | A-24 Qualification of Class All C 11/30/85 In compliance with 10 CFR 50.49 lE Safety-Related Equipment A-26 Reactor Vessel Pressure PWR NA Applies only to PWR Plants. | ||
Transient Protection | Transient Protection A-31 Residual Heat Removal All GLs After 01/79 NA POL issued 4/69. No action required A-36 Control of Heavy Loads All C 2/15/85 Committed to continued corrpliance Near Spent Fuel with Heavy Lcad requirements. | ||
A-39 Determination of SRV BWR C Addressed via USI A-7 " Mark I Pool Dynamic Loads 1.cng Term Program". | A-39 Determination of SRV BWR C Addressed via USI A-7 " Mark I Pool Dynamic Loads 1.cng Term Program". | ||
and Pressure Transients A-40 Seismic Design All I Member of SQUG. .USI A-40 will be Criteria addressed as part of SQUG talkdowns associated with USI A-46. Walkdowns | and Pressure Transients A-40 Seismic Design All I Member of SQUG. .USI A-40 will be Criteria addressed as part of SQUG talkdowns associated with USI A-46. Walkdowns scheduled for 15R, based on NRC schedule of 6/90 for supplemental SER. | ||
A-42 Pipe Cracks in Boiling BWR C Continuing inspectione in accordance Water Reactors with IGsCC Plan. | |||
scheduled for 15R, based on NRC | |||
schedule of 6/90 for supplemental SER. | |||
A-42 Pipe Cracks in Boiling BWR C Continuing inspectione in accordance | |||
Water Reactors with IGsCC Plan. | |||
A-47 | A-47 | ||
* Containment Emergency All NC No licensee action required. | * Containment Emergency All NC No licensee action required. | ||
, Sump Performance | , Sump Performance | ||
, . c , , - | , . c , , - | ||
w f,3 4tg | w f,3 4tg | ||
, y .+ | , y .+ | ||
USI NUMBER TITLE' APPLICABIITX STATUS /DATE* Ef5 MARKS a - | USI NUMBER TITLE' APPLICABIITX STATUS /DATE* Ef5 MARKS a - | ||
A-44 Station Blackout All I Responsa under *iRC review. | A-44 Station Blackout All I Responsa under *iRC review. | ||
Modifications scheduled for fall of - | Modifications scheduled for fall of - | ||
| Line 130: | Line 88: | ||
A-46 Seismic Qualification All I Member of SQUG with walkdowns scheduled for 15R, based on NRC-schedule of 6/90 for. supplemental SER. | A-46 Seismic Qualification All I Member of SQUG with walkdowns scheduled for 15R, based on NRC-schedule of 6/90 for. supplemental SER. | ||
A-47 Safety Implication All E 3/90 Evaluating GL 89-19 to determine appropriate response. | A-47 Safety Implication All E 3/90 Evaluating GL 89-19 to determine appropriate response. | ||
A-48 Hydrogen Control 'All, except PWRs with | A-48 Hydrogen Control 'All, except PWRs with C Inerted Containscnt required by Measures and Effects large dry containments Technical Specification. Mark I Containments required to be inerted for resolution of USI A-48. | ||
C Inerted Containscnt required by Measures and Effects large dry containments Technical Specification. Mark I Containments required to be inerted for resolution of USI A-48. | |||
A-49 Pressurized Thermal Shock PWR NA Applies to PWR Plants. | A-49 Pressurized Thermal Shock PWR NA Applies to PWR Plants. | ||
SC - COMPLETW NC - NO CHANGES NECESSARY NA - NOT 7,PPLICABLE I - INCOMPLETE E - EVALUATING ACTIONS REQUIRED NOTE: A reactor-vessel high water-level feedwater pump trip is not provided at Oyster Creek. The'feedwater control syst(m has been modified by the addition of a Reactor Level Setdown (RLS) program and by revising the existing feedwater pump runcut protection logic. The modification automatically lowers the effective feedwater control level setpoint when reactor water level decreases to the RLS initiation setpoint (137" TAF) and a full scram signal is present. Th'a RLS program in conjunction with the revised runout protection logic act to limit the m uitude of the post' trip high water level transient to avoid flooding the isolaaion condenser steamlines. The main turbine trips on reactor-vessel high water-level and, as a result, the reactor. scrams. The RLS initiation setpoint is selected to control water level in such en event. | SC - COMPLETW NC - NO CHANGES NECESSARY NA - NOT 7,PPLICABLE I - INCOMPLETE E - EVALUATING ACTIONS REQUIRED NOTE: A reactor-vessel high water-level feedwater pump trip is not provided at Oyster Creek. The'feedwater control syst(m has been modified by the addition of a Reactor Level Setdown (RLS) program and by revising the existing feedwater pump runcut protection logic. The modification automatically lowers the effective feedwater control level setpoint when reactor water level decreases to the RLS initiation setpoint (137" TAF) and a full scram signal is present. Th'a RLS program in conjunction with the revised runout protection logic act to limit the m uitude of the post' trip high water level transient to avoid flooding the isolaaion condenser steamlines. The main turbine trips on reactor-vessel high water-level and, as a result, the reactor. scrams. The RLS initiation setpoint is selected to control water level in such en event. | ||
; The plant emergency operating procedures provide the operators with guidance in the event of abnormal | ; The plant emergency operating procedures provide the operators with guidance in the event of abnormal reactor-vessel water level. Prior to reaching 170" TAF, the operators are instructed to trip all the feedwater | ||
reactor-vessel water level. Prior to reaching 170" TAF, the operators are instructed to trip all the feedwater | |||
' pumps. If the reactor-vessel water level exceeds 180" TAF, the guidar.ce is to isolate the isolation condensers, | ' pumps. If the reactor-vessel water level exceeds 180" TAF, the guidar.ce is to isolate the isolation condensers, | ||
, thereby preventing water from entering the condensers. | , thereby preventing water from entering the condensers. | ||
_ , _ _ _ _ _ _ - . _ _ . - - :-_ _ - - . _}} | _ , _ _ _ _ _ _ - . _ _ . - - :-_ _ - - . _}} | ||
Revision as of 15:41, 31 January 2020
| ML19332E977 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 11/30/1989 |
| From: | Long R GENERAL PUBLIC UTILITIES CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| TASK-***, TASK-OR C-89-786, C000-89-0786, GL-89-21, NUDOCS 8912130223 | |
| Download: ML19332E977 (4) | |
Text
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'I I
GPU Nuclear Corporation N IO8pm u One Upp6r Pond Road Parsippany, New Jersey 07054 201 316-7000 4
TELEX 136-482 Writer's Direct Dial Number:
1^
November 30, 1989 U.S. Nuclear Regulatory Commission C000-89-0786 Attention: Document Control Center Mail Station P1-137 Washington, DC 20555 Gentlemen:
Subjects Oyster Creek Nuclear Generating Station (OCNGS)
Docket No. 50-219 Generic Letter 89-21 Pursuant.to Generic Letter 89-21, please find attached the tabular information for DCNGS Unresolved Safety Issues as requested in Enclosure 1 to the Generic Letter. The status is based on our understanding of the issues as described in the Generic Letter and on '
discussione with and information provided by NRC Project Manager, Alex Dromerick. The information in the remarks column provide the basis for our understanding of the implementation status. The projected implementation dates do not represent a change to commitments previously provided by GPUN, and, however, are subject to change. If s there are any questions, pleas 4 contact M. W. Laggart at 201-316-7968.
Very truly yours, 1 -
6 $. / [M -
. L. Long icePresiden[t & Director Planning & Nuclear Safety JL/cjg(89-21A)
Attachment cc Mr. William T. Russell, Administrator Region I U.S. Nucloar Regulatory Commission
- p. 475 Allendale Road g King of Prussia, PA 19406 bA NRC Resident Inspector
'O Oyster Creek Nuclear Generating Station Forked River, N.J. 08751
'. QW
- g Mr. Alex Dromerick n< U.S. Nuclear Regulatory Commission
-y- Mail Station P1-137 g g Washington, D.C. 20555 g
.m44 g g GPU Nuclear Corporation is a subsidiary of General Public Utilities Corporation
,. . w -.
+ ,
cq 7 7 ATTACHMENT OYSTER CREEK NUCLEAR GENERATING STATION UNRESOLVED SAFETY ISSUES' USI NUMBER TITLE APPLICABILITY' STATUS /DATE* REMARKS A-1 Water Hammer All C 3/89 See Note on page 3.
A-2 Asymmetric Blowdown PWR NA -Applies only to PWR Plants.
Loads on Reactor Primary Coolant. Systems A-3 Westinghouse Steam W-PWR NA Applies only to Westinghouse Plants.~
Generator Tube Integrity A-4 CE Steam Generator Tube CE-PWR NA Applies only to Combustion Engineering Integrity Plants.
A-5 B&W Steam Generator B&W-PWR NA Applies only to Babcock & Wilcox Plants.
Tube Integrity A-6 Mark I containment M C h I- T.;A C 12/79 Resolved via NUREG-0408.
Short-Term Program A-7 Mark I Long-Term Mark I-BWR C 5/89 Modifications completed 5/89.
Program SER. dated 1/3/84.
A-8 Mark II Containment Mark II-BWR NA Applies only.to Mark II Poc. Dynamic Loads Containments A-9 Anticipated Transients All C A diverse and independent Alternate Without Scram Rod Injection System installed in 1988.
Standby Liquid Control System modification completed in 1988. Recirculating Pump Trip installed in 1979.
A-10 BWR Feedwater Nozzle BWR .C Committed'to continued inspections.. ,
Cracking of.the Feedwater and Control Rod Drive
' Systems in accordance with NUREG-0619.
e (a9-21)
E _ _ _ . _ _ _ . .. _u _ __ _a_ __ _ __ ___ . .-. _ _ . .
N
[pd
~ '
- ;w
~
USI NUMBER TITLE APPLICABIITY STATUS /DATK* REMARKS A-ll Reactor Vessel Material All C'3/88 Technical' Specification pressure-temperatur.4 operating limits meet both Appendix G t:e 10 CFR 50 and RG. .l.99, >
.' Rev. 2. License Amendment No. 120,1 dated 3/21/89.
A-12 Fracture Toughness of PWR NA Applies only to PWR Plants.
Steam Generator and Reactor Coclant Pump Supports A-17 Systems Interactions All HC No Action Required by G.L.
A-24 Qualification of Class All C 11/30/85 In compliance with 10 CFR 50.49 lE Safety-Related Equipment A-26 Reactor Vessel Pressure PWR NA Applies only to PWR Plants.
Transient Protection A-31 Residual Heat Removal All GLs After 01/79 NA POL issued 4/69. No action required A-36 Control of Heavy Loads All C 2/15/85 Committed to continued corrpliance Near Spent Fuel with Heavy Lcad requirements.
A-39 Determination of SRV BWR C Addressed via USI A-7 " Mark I Pool Dynamic Loads 1.cng Term Program".
and Pressure Transients A-40 Seismic Design All I Member of SQUG. .USI A-40 will be Criteria addressed as part of SQUG talkdowns associated with USI A-46. Walkdowns scheduled for 15R, based on NRC schedule of 6/90 for supplemental SER.
A-42 Pipe Cracks in Boiling BWR C Continuing inspectione in accordance Water Reactors with IGsCC Plan.
A-47
- Containment Emergency All NC No licensee action required.
, Sump Performance
, . c , , -
w f,3 4tg
, y .+
USI NUMBER TITLE' APPLICABIITX STATUS /DATE* Ef5 MARKS a -
A-44 Station Blackout All I Responsa under *iRC review.
Modifications scheduled for fall of -
1992.
A-45 Shutdown Decay Heat T11' NC Subsumed into Individual Ple.nt Removal Requirements Exami'1ation Program.
A-46 Seismic Qualification All I Member of SQUG with walkdowns scheduled for 15R, based on NRC-schedule of 6/90 for. supplemental SER.
A-47 Safety Implication All E 3/90 Evaluating GL 89-19 to determine appropriate response.
A-48 Hydrogen Control 'All, except PWRs with C Inerted Containscnt required by Measures and Effects large dry containments Technical Specification. Mark I Containments required to be inerted for resolution of USI A-48.
A-49 Pressurized Thermal Shock PWR NA Applies to PWR Plants.
SC - COMPLETW NC - NO CHANGES NECESSARY NA - NOT 7,PPLICABLE I - INCOMPLETE E - EVALUATING ACTIONS REQUIRED NOTE: A reactor-vessel high water-level feedwater pump trip is not provided at Oyster Creek. The'feedwater control syst(m has been modified by the addition of a Reactor Level Setdown (RLS) program and by revising the existing feedwater pump runcut protection logic. The modification automatically lowers the effective feedwater control level setpoint when reactor water level decreases to the RLS initiation setpoint (137" TAF) and a full scram signal is present. Th'a RLS program in conjunction with the revised runout protection logic act to limit the m uitude of the post' trip high water level transient to avoid flooding the isolaaion condenser steamlines. The main turbine trips on reactor-vessel high water-level and, as a result, the reactor. scrams. The RLS initiation setpoint is selected to control water level in such en event.
- The plant emergency operating procedures provide the operators with guidance in the event of abnormal reactor-vessel water level. Prior to reaching 170" TAF, the operators are instructed to trip all the feedwater
' pumps. If the reactor-vessel water level exceeds 180" TAF, the guidar.ce is to isolate the isolation condensers,
, thereby preventing water from entering the condensers.
_ , _ _ _ _ _ _ - . _ _ . - - :-_ _ - - . _