ML19332F362
| ML19332F362 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 12/01/1989 |
| From: | Parker T NORTHERN STATES POWER CO. |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| TASK-***, TASK-OR GL-89-21, NUDOCS 8912140325 | |
| Download: ML19332F362 (10) | |
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Nos1hern States Power Comparty 414 Nicollet Mall
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Minneapolis, Minnesota 55401 1927
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December 1. 1989 Response to f
Generic Letter.89 21 rwer,
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USNuclearRegulatorycomad.,N.y on Washington DC 205f5 PRAIRIE ISLAND NUCLEAR GENERATING PLANT Docket No. 50 282 License No. DPR-42 l
50-306 DPR-60 l
Response to Generic Letter 89 21 Unresolved Safety Issues Imolementation Status i
Generic Letter 89 21 requested that we provide the implementation status of Unresolved Safety Issues. The purpose of this letter is to provide that
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status.- Enclosure 1 is a list showing the status of each Unresolved Safety
(
Issue. Enclosure 2 contains remarks on Unresolved Safety Issues.
Please contact us if further information is required.
Thomas M Parker Manager Nuclear; Support Services j
c: Regional Adminittrator - III, NRC Sr Resident Inspector, NRC Sr Project Manager, NRC G Charnoff Attachments s
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ENCII)SURE 1 - - PRAIRIE ISIAND UNRESOLVED SAFETY ISSUES FOR WHICH A FINAL TECHNICAL RESOLUTION HAS BEEN ACHIEVED USI/MPA NUMBER TITLE REF. DOCUME E APPLICABILITY STATUS /DATE*
REMARKS A-1 Water Hammer SECY 84-119 All C
- See Enclosure-2.
NUREG-0927, Rev. 1 NUREG-0993, Rev. 1 NUREG-0737 Item I.A.2.3 SRP revisions A-2/
Asymmetric Blowdown NUREG-0609 PWR C
See Enclosure 2 MPA D-10 Loads on Reactor Primary GL 84-04, GDC-4 Coolant Systems A-3 Westinghouse Steam NUREG-0844 W-PWR C
See Enclosure 2 Cenerator Tube Integrity SECY 86-97 SECY 88-272 GL 85-02 (No requirements)
A-4 CE Steam Generator Tube - NUREG-0844, SECY 86-97 CE-PWR NA See Enclosure 2 Integrity SECY 88-272 CL 85-02 (No requirements)
A-5 B&W Steam Generator NUREG-0844, JECY 86-97 B&W-PWR NA
. See Enclosure 2 Tube Integrity SECY 88-272 GL 85-02 (No requirements)
A-6 Mark I Containment-NUREG-0408 Mark I-BUR NA See Enclosure 2 Short-Term Program
- C - COMPLETE NC - NO CHANGES NECESSARY NA - NOT APPLICABLE I - INCOMPLETE E - EVALUATING. ACTIONS REQUIRED g
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UST/MPA NUMBER TITLE REF. DOCUMENT APPLICABILITY STATUS /DATE*
REMARKS
-l A-7/
Mark I Long-Tern NUREG-0661 Mark I-BWR NA See Enclosure 2 D-01 Program NUREG-0661 Supp1'. 1 i
GL 79-57 A-8 Mark II Containment NUREG-0808 Mark II-BWR NA See Enclosure 2 Pool Dynamic Ioads NUREG-0487, Suppl. 1/2 NUREG-0802 SRP 6.2.1.1C GDC 16 A-9 Anticipated Transients NUREG-0460, Vol. 4 All I
See Enclosure 2 Without Scram 10 CFR 50.62 A-10/
BWR Feedwater Nozzle NUREG-0619 BWR NA See Enclosure 2 MPA B-25 Cracking Letter from DG Eisenhut dated 11/13/80 GL 81-11 A-11 Reactor Vessel Material NUREG-0744, Rev. 1 All C
See Enclosure 2 Toughness 10 CFR 50.60/
82-26 A-12 Fracture Toughness of NUREG-0577, Rev. 1 PWR NC See Enclosure 2 Steam Generator and SRP Revision Reactor Coolant Pump 5.3.4 Supports A-17 Systems Interactions Ltr: DeYoung to All I
See Enclosure 2 licensees - 9/72 NUREG-1174. NUREG-1229, NUREG/CR-3922, NUREG/CR-4261, NUREG/
CR-4470, GL 89-18 (No requirements)
A-24/
Qualification of Chss NUREG-0588, Rev. 1 All C
See Enclosure 2 MPA B-60 1E Safety-Related SRP 3.11 Equipment 10 CFR 50.49 GL 82-09, CL 84-24 CL 85-15 a
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.. USI/MPA NUMBER TITLE' REF. DOCUMENT-APPLICABILITY STATUS /DATE*
REMARKS A-26/
Reactor Vessel Pressure DOR Letters to~
PWR G
See Enclosure 2-MFA B-04 Transient Protection Licensees 8/76 FUREG-0224 NUREG-0371~
SRP 5.2 GL 88-11 A-31 Residual Heat Removal NUREG-0606 All OLs After NA See Enclosure 2 Shutdown Requirements RG 1.113, 01/79 RG 1.139 SRP 5.4.7 A-36/
Control of Heavy loads NUREG-0612 All C
See Enclosure 2 G-10, Near Spent Fuel SRP 9.1.5 C-15 GL 81-07, GL 83-42, GL 85-11 Letter from DG Eisenhut dated 12/22/80 A-39 Determination of SRV NUREG-0802 BVR NA See Enclosure 2 Pool Dynamic inads NUREGs-0763,0783,0802 and Pressure Transients NUREG-0661 SRP 6.2.1.1.G A-40 Seismic Design SRP Revisions, NUREC/
All I
See Enclosure 2 Criteria CR-4776. FUREG/CR-0054, NUREG/CR-3480,'NUREG/
CR-1582, NUREG/CR-1161, i
NUREG/CR-3905 NUREG/CR-5347 NUREG/CR-3509 j
A-42/
Pipe Cracks in Boiling NUREG-0313. Rev. 1 BVR NA See Enclosure 2 MPA B-05 Vater Reactors NUREG-0313, Rev. 2 l
GL 81-03, CL 88-01.
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USI/MPA
- NUMBER TITLE REF. DOCUMENT' APPLICABILITY
. STATUS /DATE*
REMARKS A-43 Containment Emergency NUREG-0510 All NA See Enclosure 2 Sump Performance NUREG-0869, Rev. 1 NUREG-0897, RG 1.82 (Rev. 0), SRP 6.2.2 GL 85-22 (No requirements)
A-44 Station Blackout RG 1.155 All 1
See Enclosure 2 NUREG-1032 NUREG-1109 10 CFR 50.63 A-45 Shutdown Decay Heat SEGY 88-260 All I
See Enclosure 2 Removal Requirements NUREG-1289 NUREG/CR-5230 SECY 88-260 (No requirements)
A-46 Seistsic Qualification NUREG-1030.
All I
See Enclosure 2 of Equipment in NUREG-1211/
Operating Plants GL 87-02, GL 87-03 A-47 Safety implication NUREG-1217, All E
See Enclosure 2 of Control Systems NUREG-1218 GL 89-19 A-48 Hydrogen Control 10 CFR 50.44 All, except NA.
See Enclosure 2 Measures and Effects SECY 89-122 PWRs with of Hydrogen Burns large dry on Safety Equipment containments A-49 Pressurized Thermal RGs 1.154, 1.99 PWR C
See Enclosure 2 Shock SECY 82-465 SECY 83-288 SECY 81-687 10 CFR 50.61/,
GL.88-11
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j USI Remarks A-1 Vater Hammer The Prairie Island Operator Training Program covers water hammer events.
Prairie Island responded to NRC letter dated May 12, 1975 on May 29, 1975, r
September 2,1975 and submitted a report on January 29, 1976 entitled,
" Analysis of PWR Secondary System Fluid Flow Instability."
Prairio Island responded to NRC letter dated September 2, 1977 on December 30, 1977. The j-December 30, 1977 submittal refers to the January 29, 1976 submittal. The NRC issued a Safety Evaluation Report on September 13, 1979.
A-2 Asymmetric Blowdown Loads in RCS On October 24, 1984 Prairie Island submitted Westinghouse Report WCAP 10640,
" Technical Bases for Eliminating Large Primary Loop Pipe Rupture as the Structural Design Basis for Prairie Island Unit 1."
On October 21, 1985 Prairie Islar.d submitted Westinghouse Reports WCAP 10928 and 10930, " Technical Bases for Eliminating Large Primary Loop Pipe Rupture as the Structural Design Bases for Prairie Island Unit 2" and " Toughness Criteria for Thermally Aged Cast Stainless Steel"-respectively.
The NRC issued a Safety Evaluation Report on December 22, 1986.
A-3. A-4.
and A-5 Steam Generator Tube Interrity Prairie Island provided steam generator inspection program information in response to Generic Letter 85 02, dated June 19, 1985.
A6 Mark I Containment Short Term Program This is a BWR issue, A-7 Mark I Lone Term Program This is a BWR issue.
A-8 Mark II Containment Pool Dynamic Loads This is a BWR issue.
A-9 ATUS ner 10 CFR 50.62 The NRC issued a Safety Evaluation Report dated August 17, 1988 on the Prairie Island AMSAC design. Modification was completed in March 1989 for
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Unit 2. and will be completed in February 1990 for Unit 1.
NRC Inspection Report No. 88022 dated August 31, 1989 verified ATWS completion for Unit 2.
1 Ai10 BWR Feedwater Norzle Crackine h
This'is a BWR issue.
A-11 Remet'or Vessel Materials Tourhness l
Prairie Island submitted its Reactor Vessel Material Surv-111ance Program on October 31, 1977.
A-12 Potential of Low Fracture Tourhness and immellar Tearine in PVR SG and
- RCP Sunnorts 1
. Prairie Island responded to NRC letter dated September 14, 1977 with a report on November 11, 1977 entitled, "An Examination of Reactor Coolant Pump and Steam Generator Supports"; the report was updated on December 27, 1977. A review of the report indicates that the Prairie Island equipment supports are free of the design and material deficiencies described in the'NRC letter.
A-17 Systems Interactions in Nuclear Power Plants Prairie Island is preparing a PRA which sould_ resolve this issue.,, The schedule for PRA/IPE completion is attached to NSP letter dated October 30, i
1989, " Response to Individual Plant Examination'for Severe Accident i
Vulnerabilities Ceneric Letter No. 88 20, Supplement No.
1."
The Prairie Island PRA should be completed in February 1993.
A-24 Oualification of Class 1E Eauioment L
i The NRC transmitted a Safety Evaluation Report and a Technical Evaluation
' Report on April 25, 1983. NSP responded on May 5, 1983, May 19, 1983 and May 27, 1983. The NRC transmitted a Safety Evaluation Report on June 3,1983.
NSP provided additional information on August 10, 1983.
The NRC issued a final Safety Evaluation Report on March 25, 1985 NRC Inspection Report Nos. 50 282/86012 and 50 306/86014 dated December 29, i,
1986 documents the Environmental Qualification inspection to verify compliance with the Environmental Qualification rule.
A-26 Reactor Vessel Pressure Transient Protection e
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Prairie Island responded to NRC letter dated August 13, 1976 on September 3, 1976 and December 10, 1976. The Prairie Island response to NRC letter dated January 10, 1977 on February 25, 1977, stated that the modifications would be completed by the end of 1977, and that a backup air supply would be installed in Unit I in the Spring 1978 outage and in Unit II in the Fall 1977 outage.-
Prairie Island responded to NRC letter dated February 11, 1977 on April 4, 1977.
Prairie Island submitted a report on July 22, 1977 entitled, " Pressure Mitigating Systems Transient Analysis Results."
Prairie Island responded to the NRC request for Tech Specs dated May 1, 1978 on August 4,,1978.
The NRC issued a Safety Evaluation Report with License Amendment Nos. 38 and 32-on August 3, 1979.
A-31 RHR Shutdown Recuirements Not Applicable. License received prior to January 1979.
A-36 control of Heavy Loads. Phase I and II Prairie Island responded to the NRC request dated May 17, 1978 for information on control of heavy loads near spent fuel on July 21, 1978.
.The NRC c}osed out " Control of Heavy loads - Phase I" in a Safety Scaluation Report dated June 6,1983. The Technical Evaluation Report was sent to NSP on May 31, 1984. Generic Letter 85-11, dated June 28, 1985 closed out " Control of. Heavy Loads Phase II. "
A-39 Determination of SRV Pool Dynamic Loads and Temocrature Limits
- This is a BUR issue.
'A 40 Seismic Desien Criteria l'
Prairie Island is participating in the Seismic Qualification Utility Group (SQUG).
See USI A 46.
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-A 4,2 Pine Cracks in Boiline Water Reactors This is a BWR issue.
A-43 Containment Emercenev Sumo Performance A-43 is applicable to new construction only.
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-Enclosure 2 i
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t A-44 Station'B1'ackout
. Prairie Island provided information required by 10 CFR Part 50 Section 50.63(c)(1), loss of all alternating current power,-on April 13, 1989.. The
. modifications and procedure changes. identified;in the April 13, 1989 submittal 1lJ
- will be completed prior to-the starty of Cycle 15 for Unit 2 (1992)..
A-45 Shutdown Decay Heat Removal Re;; tekents Prairie Ieland is preparing a PRA which sould resolve this issue.
The fccheduleitur PRA/IPE completion is attached to NSP letter dated October-30,
.1989, " Response to Individual. Plant Examination for Severe Accident Vulnerabilities Generic Letter No. 88-20, Supplemert do.
1."
The Prairie
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' Island.PRA should~be completed in February 1993.
A-46 Seismic O alification of Eatiinment in Ooeratine Plants L
Prairie Island response to Generic Letter 87-02, on October 6, 1988 stated
'that Prairie Island plans to have seismic. verification plant walkdowns within 60 days.following' conclusion of the third refueling outage after receipt of tbs iinal staff Safety Evaluation Report supplement and resolution oftall open issues and completion of walkdown team training.
' A-47 Safety Imnlication of Control Systems in LUR Nuclear Power Plants
' Prairie Island's response to Generic Letter 89-19 is due March l5,1990.
A-48 Hydroren Contro1~Mee.sitres and Effects of Hydromen Burns on Safety Eauloment Since-Prairie Island is a PWR with a large dry containment Unresolved Safety
. Issue'A-48 does not apply.
HowcVer, in response to NUDEG-0737, Prairie Island
- hydrogen recombiners were installed; Unit I in thi 19E? refueling outage and Unit 2 in tne 1983 refueling outage. High point vents aere installed in 1981
-refueling outage for Unit 1 and in the 1982 refueling outage for Unit 2.
Tech'nical Specifications were requested on July 1,1983.
License Amer.dment Nos. 68 and 62 were issued on February 21, 1984.
. A,-f 4 Pressuri::ed Thermal Shog s
. Prairie Island letter daced January 10, 1986 submitted an Assessment of Pressurized Thermal Shock Reference Temperature in Accordance with 10 CFR Part 13f'.
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1 Page.5-SC, Section 50.61. - The NRC issued Safety Evaluation Reports for Unit 1 on 4
. August 18 1986 and for Unit 2 on June 23, 1986..
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