ML18038A771
| ML18038A771 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 11/28/1989 |
| From: | Terry C NIAGARA MOHAWK POWER CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| RTR-NUREG-0612, RTR-NUREG-612, TASK-***, TASK-OR GL-89-21, NMP1L-0459, NMP1L-459, NUDOCS 8912050090 | |
| Download: ML18038A771 (92) | |
Text
ACCELERATED D UTION DEMONSTRATION SYSTEM
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REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)
ACCESSION NBR:8912050090 DOC.DATE: 89/ll/28 NOTARIZED: NO DOCKET FACIL:50-220 Nine Mile Point Nuclear Station, Unit 1, Niagara Powe 05000220 AUTH. NAME AUTHOR AFFXLIATXON TERRY,C.D. Niagara Mohawk Power Corp.
RECIP.NAME RECIPXENT AFFILIATION Document Control Branch (Document Control Desk) R
SUBJECT:
Responds to Generic Ltr 89-21 for request for info re status of implementation of USI requirements.
DXSTRIBUTION CODE: A012D COPXES RECEIVED:LTR ENCL SXZE: D TITLE: Generic Ltr 89-21 Response,Xmplementation of Unresolved Safety Issue NOTES
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RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD1-1 PD 1 1 MARTIN,R. 1 1 D,
INTERNAL: ACRS 1 1 B RBER G PTSB 1 1 NUDOCS-ABSTRACT 1 1 G L 01 1 1 D WESSMAN,R PD13 1 1 EXTERNAL: LPDR 1 1 NRC PDR 1 1 S, NSXC 1 1 R
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D I
NOTE TO ALL "RIDS" RECIPIENTS:
PLEASE HELP. US TO REDUCE WASIZt CONTACT THE. DOCUMENT CONTROL DESK, ROOM Pl-37 (EXT. 20079) TO ELIMINATEYOUR NAME FROM DISTRIBUTION LISIS FOR DOCUMENTS YOU DON'T NEEDt TOTAL NUMBER OF COPIES REQUIRED: LTTR 10 ENCL 10
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al 7 NIAGARA H O MOHAWK NIAGARAMOHAWKPOWER CORPORATION/301 PI&INFIELDROAD, SYRACUSE, N.Y. 13212/TELEPHONE (315) 474-1511 November 28, 1989 NHPlL 0459 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555 Re: Nine Mile Point Unit 1 Docket No. 50-220 DPR-63 Gentlemen:
On October 19, 1989, the Nuclear Regulatory Commission issued a request for information concerning status of implementation of Unresolved Safety Issue (USI) requirements (Generic Letter 89-21). to this letter tabulates the requested status of implementation of USIs for which a final technical resolution has been achieved and which are applicable to Nine Mile Point Unit 1. The status provided is based on Niagara Mohawk's judgement of what actions were necessary to resolve the technical issue's for Unit 1 and is not meant to imply compliance with all referenced guidance documents.
Very truly yours, NIAGARA HOHANK PO CORPORATION r
R C. D. Terr Vice President Nuclear Engineering and Licensing JMT/mjd 7998G xc: Regional Administrator, Region I Mr. R. AD Capra, Director Mr. R. E. Hartin, Project Manager Hr. N. A. Cook, Resident Inspector Records Management S912050090 89112S PDR ADOCK,05000220 P PDC
A Il II
ENCLOSURE 1'SI/MPA NUMBER TITLE -REF. DOCUMENT APPL I CAB I LITY STATUS/DATE* REMARKS A-1 Water Hammer SECY 84-119 All 'MI training Item I.A.2.3 NUREG-0927, Rev. 1 incorporated (TMI Item NUREG-0993, Rev. Status Letter 4/8/89).
'ore 1
NUREG-0737 Item Spray System Water I.A.2.3 Hammer Issue still open.
SRP revisions Niagara Mohawk submitted an evaluation on July 6, 1989 (NMP1L 0418).
Awaiting NRC SER.
'A-2/ Asymmetric Blowdown NUREG-0609 PWR NA NMPl is a BWR MPA D-10 Loads on Reactor Primary GL 84-04, GDC-4 Coolant Systems V A-3 Westinghouse Steam NUREG-0844 W-PWR NMPl is a BWR.
Generator 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 NMP1 is a BWR.
Integrity SECY 88-272 GL 85-02 (No requirements)
- C COMPLETE NC NO CHANGES NECESSARY NA NOT APPLICABLE I INCOMPLETE E EVALUATING ACTIONS REQUIRED 7998G
0
)
USI/MPA NUMBER TITLE REF. DOCUMENT APPLICABILITY STATUS/DATE* REMARKS A-5 BKH Steam Generator NUREG-0844, SECY 86-97 BE(H-PHR NMPl is a BWR.
Tube Integrity SECY 88-272 GL 85-02 (No requirements)
E A-6 Mark I Containment NUREG-0408 Mark I--BHR C/1-22-85 See NRC's SER dated Short-Term Program 1/22/85.
A-7/ Mark I Long-Term NUREG-0661 Mark I-BHR C/1-22-85 The Commission's letter D-01 Program NUREG-0661 Suppl. 1 dated 1/22/85 documented~
GL 79-57 their review and accepta~
of the NMP1 Mark I containment program.
A-8 Mark II Containment NUREG-0808 Mark II-BHR NA NMP1 is a Mark I BHR.
Pool Dynamic Loads NUREG-0487, Suppl. 1/2 NUREG-0802 SRP 6.2.1.1C GDC 16 A-9 Anticipated Transients NUREG-0460, Vol. 4 Al 1 Completed during 1988/89 Without Scram 10 CFR 50.62 outage.
- C COMPLETE NC NO CHANGES NECESSARY NA NOT APPLICABLE I INCOMPLETE E EVALUATING ACTIONS REQUIRED 7998G
4 USI/MPA NUMBER TITLE REF. DOCUMENT APPLI CAB I LITY STATUS/DATE* REMARKS A-10/ BHR Feedwater Nozzle NUREG-0619 BHR C/6/84 'ee Niagara Mohawk's MPA B-25 Cracking Letter from DG Eisenhut response to NUREG-0619 dated 1 1/13/80 commitments dated GL 81-11 December 29, 1980.
'iagara Mohawk made modifications to feed-water,low flow control system to reduce/
eliminate thermal cycling on feedwater nozzles in 1984.
'iagara Mohawk is maki operational changes to further reduce/eliminate thermal cycling on feed-water nozzles beginning with the startup from the current outage.
A-11 Reactor Vessel Material NUREG-0744, Rev. 1 Al 1 Vessel material properties Toughness 10 CFR 50.60/ inspected in accordance 82-26 with Tech Spec 4.2.2.b.
A-12 Fracture Toughness of NUREG-0577, Rev. 1 PHR NA NMPl is a BHR.
Steam Generator and SRP Revision Reactor Coolant Pump 5.3.4 Supports
- C COMPLETE NC NO CHANGES NECESSARY NA NOT APPLICABLE I INCOMPLETE E EVALUATING ACTIONS REQUIRED 7998G
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USI/MPA NUMBER TITLE REF. DOCUMENT APPLICABILITY STATUS/DATE* REMARKS C
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A-17 Systems Interactions Ltr: DeYoung to All NC Generic Letter 89-18 was licensees-9/72 issued for information only.
NUREG-1174, NUREG- No specific action or 1229, NUREG/CR-3922, written response was NUREG/CR-4261, NUREG/ required. No action has CR-4470, GL 89-18 been taken.
(No requirements)
A-24/ Qualification of Class NUREG-0588, Rev. 1 Al 1 Safety Evaluation docu-MPA B-60 lE Safety-Related SRP 3.11 menting NMP1 complies with Equipment 10 CFR 50.49 10CFR50.49 issued on GL 82-09, GL 84-24 1/10/85.
GL 85-15 A-26/ Reactor Vessel Pressure DOP Letters to PWR NA NMPl is a BWR.
MPA B-04 Transient Protection Licensees 8/76 NUREG-0224 NUREG-0371 SRP 5.2 GL 88-11 A-31 Residual Heat Removal NUREG-0606 All OLs After NA NMP1's Operating License Shutdown Requirements RG 1.113, 01/79 was issued prior to 1/79.
RG 1.139 SRP 5.4.7 A-36/ Control of Heavy Loads NUREG-0612 All Safety Evaluation C-10, Near Spent Fuel SRP 9.1.5 documenting NMPl complie C-15 GL 81-07, GL 83-42, with NUREG-0612, Section GL 85-11 5.1.1 and 5.3 (Control of Letter from DG Heavy Loads-Phase I) issued Eisenhut dated on 3/5/85.
12/22/80
- C COMPLETE NC NO CHANGES NECESSARY NA NOT APPLICABLE I INCOMPLETE E EVALUATING ACTIONS REQUIRED 7998G
USI/MPA NUMBER TITLE REF. DOCUMENT APPLICABILITY STATUS/DATE* REMARKS A-39 Determination of SRV NUREG-0802 C/1-22-85 For Hark" I Containments; Pool Dynamic Loads NUREGs-0763,0783,0802 SRV acceptance criteria is and Pressure Transients NUREG-0661 presented in NUREG-0661 SRP 6.2.1.1.C and dealt with as part of USI A-7.
A-40 Seismic Design SRP Revisions, NUREG/ All USI A-40 being resolved Criteria CR-4776, NUREG/CR-0054, as -part of A-46.
NUREG/CR-3480, NUREG/
CR-1582; NUREG/CR-1161, NUREG-1233, NUREG/4776 NUREG/CR-3805 NUREG/CR-5347 NUREG/CR-3509 A-42/ Pipe Cracks in Boiling NUREG-0313, Rev. 1 BHR I/Next Refuel NMPl's response to.Generic MPA B-05 Hater Reactors NUREG-0313, Rev. 2 Outage Letter 88-01 is found in GL 81-03, GL 88-01 letters dated 7/28/88, 8/25/89 and 9/6/89.
Requirements of GL 88-01 are scheduled to be incorporated into the ISI
-program prior to the next refuel outage per the 7/28/89 letter. The NMPl Technical Specifications were revised to conform to the positions delineated in GL 88-01 by Amendmen No. 107. The Safety Evaluation of Implementation of NUREG-0313, Rev. 1 was issued on 6/6/84.
- C COMPLETE NC NO CHANGES NECESSARY NA NOT APPLICABLE I INCOMPLETE E EVALUATING ACTIONS REQUIRED 7998G:
4' USI/MPA NUMBER TITLE REF. DOCUMENT APPLICABILITY STATUS/DATE* REMARKS A-43 Containment Emergency NUREG-0510, Al 1 NC 'o requirements.
Sump Performance NUREG-0869, Rev. 1 NUREG-0897, R.G. 1.82 (Rev. 0), SRP 6.2.2 GL 85-22 (No requirements)
A-44 Station Blackout RG 1.155 Al 1 'rocedural Chan es-NUREG-1032 within 12 months of NUREG-1109 notification in accordance Modifications ~
'atter10CFR50.63(c)(3).~
with (As required) within 24 f
months. of noti i cati on in accordance with 10CFR50.63(c)(3).
'osition documented in letter dated 4/13/89.
A-45 Shutdown Decay Heat SECY 88-260 Al 1 Development of an NMP1 IPE Removal Requirements NUREG-1289 is scheduled for July 30, NUREG/CR-5230 1993.
SECY 88-260 (No requirements)
A-46 Seismic Qualification NUREG-1030 Al 1 NMPl is a member of the of Equipment in NUREG-1211/ Seismic Qualification Operating Plants GL 87-02, GL 87-03 Utility Group (SQUG)-US~
A-46 is being resolved as group solution with NMPl as the pilot plant of the process.
- C COMPLETE NC NO CHANGES NECESSARY NA NOT APPLICABLE I INCOMPLETE E EVALUATING ACTIONS REQUIRED 7998G
wP C
4 USI/MPA NUMBER TITLE REF. DOCUMENT APPLICABILITY STATUS/DATE* REMARKS A-47 Saf ety Imp 1 i cati on NUREG-1217, NUREG- All NMP1 is currently of Control Systems 1218 evaluating the requirements GL 89-19 of GL 89-19 and is expected to provide its response by 3/19/90.
A-48 Hydrogen Control 10 CFR 50.44 All, except 'ontainment inerted.
Measures and Effects SECY 89-122 PHRs with 'ontainment Atmosphere of Hydrogen Burns large dry Dilution System (CAD) added on Safety Equi pment containments per modification Nl-72-0 .
'afety Evaluation documenting NMP1 complie with 10CFR50.44(c)(3)(ii) issued on 4/29/85.
A-49 Pressurized Thermal RGs- 1.154, 1.99 PWR NA NMPl is a BWR.
Shock SECY 82-465 SECY 83-288 SECY 81-687 10 CFR 50.61/
- C COMPLETE NC NO CHANGES NECESSARY NA NOT APPLICABLE I INCOMPLETE E EVALUATING ACTIONS REQUIRED 7998G
'PLANT NMP- I DOCKET NO(S). 50-220 PROJECT MANAGER Robert E. Martin TECHNICAL CONTACT A. Serkiz USI NO. A-I TITLE Hater Hewer MPA NO. ~NA TAC NOS.
ISSUES
SUMMARY
This Unresolved Safety Issue (USI) was resolved in March 1984, with the publication of NUREG-0927, "Evaluation of Water Hammer in Nuclear Power Plants
- Technical Findings Relevant to Unresolved Safety Issue A-l." Also on March 15, 1984, the EDO sent the Commissioners SECY 84-119 titled, "Resolution of Unresolved Safety Issue A-l, Water Hammer."
In SECY 84-119, the staff concluded that the frequency and severity of water hammer occurrences had been significantly reduced through (a) incorporation of design features such as keep-full systems, vacuum breakers, J-tubes, void detection systems, and improved venting procedures; (b) proper design of feed-water valves and control systems; and (c) increased operator awareness and training. Therefore, the resolution of USI A-I did not involve any hardware or design changes on existing plants. It did involve Standard Review Plan (SRP) changes (forward fits) and a comprehensive set of guidelines and criteria to evaluate and upgrade utility training programs (per TMI Task Action Plan Item I.A.2.3). In addition, the assumption was made that for BWRs with isolation condensers ( ICs) a reactor-vessel high water-level feedwater pump trip was in place or being installed. This was necessary because calculated values had postulated an IC failure by water hammer that opened a direct pathway to the environment.
IMPLEMENTATION AND STATUS
SUMMARY
PLANT SPECIfIC):
By letter dated April 13, 1978, the licensee responded to the Comission's letter dated February I, 1978 regarding the need for a feedwater pump trip on reactor high water level. By letter dated December 15, 1978, the staff informed the licensee that after review of their April 13, 1978 response and subsequent verbal communication with members of the licensee's organization, the staff had determined that a trip of the motor driven feedwater pumps on high reactor water level was not necessary to assure safe operation of Nine Mile Point Unit 1.
However, several years later, in response to the issue of the hi-level trip as a TMI Action Item the licensee committed by letter dated April 1, 1982, to install, prior to star tup from the next refueling outage, a reactor vessel high level trip of the motor driven feedwater pumps. For the purpose of USI A-1, the issue was completed prior to the July 24, 1984 startup by the installation of a reactor vessel high level trip of the moter driven feedwater pumps.
In addition to the scope of the issue dealt with by USI A-l, a concern was raised during Safety System Funtional Inspection 88-201 regarding the potential for water hammer during star tup of the core spray system during a LOCA.
4 By letter dated March 28, 1989, the licensee forwarded to the NRC their analysis for the potential for water hammer during startup of the core spray system during a LOCA.
By letter dated April 18, 1989, the licensee forwarded to the NRC the status of TMI action plan items. TMI issue number I.A.2.3 (Administration of Training Program) is shown as being completed.
By letter dated July 6, 1989 Niagara Mohawk submitted a response to a staff request for additional information for Safety System Functional Inspection Unresolved Item 88-201-2C regarding the potential for Waterhammer during startup of the core spray system during a LOCA. The staff has reviewed the licensee's response and found it acceptable.
4 I
REFERENCES:
NMP-1 A-1
- 1. RE UIREMENT DOCUMENTS:
TITLE NUDOCS NO. DATE Letter from Denton to Utilities, 8403150310 03/05/84 "Notice of Issuance and Availability NUREG-0927 Rev. 1, Safety Issue A-1" NUREG-0927 "Evaluation of Water 8306060413 05/31/83 Hammer in Nuclear Power Plants-Technical Findings Relevant to Unresolved Safety Issue A-1" NUREG-0993 Rev. 1 8306060418 March 1984 "Regulatory Analysis for for USI A-l, Water Hammer" SRP Sections: 3.9.3, 3.9.4, 5.4.6, 5.4.7, 6.3, 9.2.1, 9.2.2, 10.3, and 10.4.7 SECY-84-119, "Resol ution 03/15/84 of Unresolved Safety A-l, Water Hammer"
- 2. IMPLEMENTATION DOCUMENTS:
TITLE NUDOCS NO. DATE Letter from T. Ippolito (NRC) 7812290049 12/15/78 to D. P. Disc (NMPC)
Letter from D. P. Disc (NMPC) 8204060097 04/01/82 to D. Eisenhut (NRC)
Letter from C. D. Terry (NMPC) 8904050067 03/28/89 to NRC Letter from L. Burkhardt (NMPC) 8904260240 04/18/89 to NRC Letter from C. D. Terry (NMPC) 8907110345 07/06/89 to NRC
- 3. VERIFICATION DOCUMENTS:
TITLE NUDOCS NO. DATE
PLANT NMP-1 DOCKET NO(S). 50-220 PROJECT MANAGER Robert E.Martin TECHNICAL CONTACT J. Kudrick.
USI NO. A-6 TITLE Mark I Containment Short Tenn Pro ram MPA NO. TAC NOS.
ISSUES
SUMMARY
This USI was resolved in December 1977 with the publication of NUREG-0408, "Mark I Containment Short-Term Program Safety Evaluation Report."
The objectives of the Mark I short-term program were: (a) to examine the containment system of each BWR facility with a Mark I containment design to verify that it would maintain its integrity and functional capability when subjected to the most probable hydrodynamic loads induced by a postulated design-basis LOCA, and (b) to verify that licensed Mark I BWR facilities could continue to operate safely, without undue risk to the public health and safety until such time as a methodical, comprehensive long-term program is conducted.
The NRC staff used a safety factor of at least two to failure for the weakest structural or mechanical component in the Mark I containment system in judging that containment integrity and functions would be assured under most probable design-basis LOCA-induced hydrodynamic loads.
As indicated in NUREG-0408, the staff required full implementation of the calculation of the hydrodyiiamic loads and structural analysis as an interim measure until complete implementation of the long-term program had been achieved. In NUREG-0408 the staff concluded that the objectives of the Short-Term Program had been satisfied, thus documenting the basis for resolving this safety issue. This issue is considered complete for all affected BWRs.
IMPLEMENTATION AND STATUS
SUMMARY
PLANT SPECIFIC :
By letter dated September 3, 1976, October I, l976, and October 14, 1976, the licensee provided information to the NRC regarding the resolution of USI A-6.
On February 28, 1978 an Exemption from GDCSO was granted to all affected licensees. These exemptions concerned a minimum margin of safety of two in the containment design as part of the short term program. This was deemed an adequate basis for continued operation until the completion of the Long Term Program.
An Order for Modification of License and Grant of Extension of Exemption was issued on January 13, 1981. This Order and Exemption was extended by a further Order dated January 19, 1982 to require completion of modifications to meet the Long Term Program defined in NUREG-0661.
For purposes of documenting a programmatic endpoint of USI A-6 for NMP-1, the Short Term Program, the issuance of the Exemption on February 28, 1978 is utilized.
REFERENCES:
NMP-1 A-6
- 1. RE UIREMENT DOCUMENTS:
TITLE NUDOCS NO. DATE NUREG-0408, "Mark I Containment 10/77 Short Term Program Safety Evaluation Report" (See Table I-2 for letters to BWR licensees requesting action)
- 2. IMPLEMENTATION DOCUMENTS:
TITLE NUDOCS NO. DATE G. Rhode (NMPC) to G. Lear (NRC) 10/14/76 Exemption from GDC 50 on 02/28/78 containment design minimum margin of safety TA Ippolito (NRC) to DP Disc (NMPC) issuing Order and Extension of Exemption 01/13/81 D. Vassallo (NRC) to D.P. Disc (NMPC) issuing Order modifying the 1/13/81 Order 01/19/82
- 3. VERIFICATION DOCUMENTS TITLE NUDOCS NO. DATE
'LANT NMP-I DOCKET NO(S). 50-220 PROJECT MANAGER Robert E. Martin TECHNICAL CONTACT J. Kudrick OSI NO. A-7 TITLE Mark I Loo Term Pro ram MPA NO. TAO NOS. 07942 ISSUES
SUMMARY
This USI was resolved in August 1982 with the publication of Supplement 1 to NUREG-0661, "Safety Evaluation Report, Hark I Containment Long-Term Program" and Standard Review Plan Section 6.2.1.1.C. For operating BWRs, MPA D-Ol was established for implementation purposes.
The focus of this USI was the suppression pool hydrodynamic loads, associated with a postulated LOCA, which had not explicitly been included in the original Mark I containment design. The issue was identified during large-scale testing of a Mark III containment design. The staff addressed this issue in NUREG-0661, published in July 1980, and in Supplement 1 to NUREG-0661, published in August 1982.
The objective of the long-term program (LTP) was to establish the design-basis loads that are appropriate for the anticipated life of each Mark I BWR facility and to restore the originally intended design-safety margins for each Mark I containment system. The principal thrust of the LTP was the development of generic methods for defining suppression pool hydrodynamic loadings and the associated structural assessment techniques for the Mark I configuration. On the basis of experimental and analytical programs conducted by the Mark I Owners Group, it was determined that the hydrodynamic load definition pro-cedures, with some modifications defined in NUREG-0661, provided a conservative estimate of these loading conditions. Thus, the requirements associated with this USI were concerned with the structural assessment of Hark I containments and related structures to the hydrodynamic loads defined by the staff in the LTP.
In January 1981, the staff issued "Orders For Modification of License and Grant of Extension of Exemptions" to each licensee of a Mark I plant. The orders required the licensees to assess the suppression pool hydrodynamic loads in accordance with General Electric documents and NUREG-0661 on a defined schedule. For some plants, the implementation schedule was extended by a subsequent order.
IMPLEMENTATION AND STATUS
SUMMARY
PLANT SPECIFIC :
On January 19, 1982 the NRC issued an Order extending the completion date for the Mark I Long-Term Program at the Nine Mile Point Unit No. 1 facility to read: "Prior to the start of Cycle 8 at the completion of your Spring 1983 refueling outage."
By letter dated January 5, 1983, the licensee committed to complete Mark I Long-Term Program modifications prior to the start of the Cycle 8 in 1984.
The staff's letter of 1/07/83 provided the staff's position on interpretation of the required completion date for cycle 7 and the beginning of cycle 8 to meet the requirement of the Order. The licensee.has recently orally confirmed that these modifications were completed prior to the start of cycle 8 on 06/13/84.
As additional background information concerning the staff's post-implementation review of the plant unique analysis report, it is noted that the staff forwarded to the licensee the results of its review made with the assistance of Srookhaven National Laboratory and Franklin Research Center.
The staff concluded that the modifications made by the licensee were in accordance with the generic acceptance criteria contained in Appendix A of NUREG-0661. Where deviations from the acceptance criteria specified in NUREG-0661 had been taken, they were found acceptable by the staff.
REFERENCES:
NMP-1 A-7
- 1. RE UIREMENT DOCUMENTS:
TITLE NUDOCS NO. DATE NUREG-0661, "Safety Evaluation Report, Mark I Containment Long Term Program" 07/80 NUREG-0661, Supplement 1 08/82 Orders for Modification to License for Applicable Licensees 1981
- 2. IMPLEMENTATION DOCUMENTS:
TITLE NUDOCS NO; DATE T. A. Ippolito (NRC) to D. P. Disc (NMPC) issuing Order and Extension of Exemption 1/13/81 Letter from D. P. Disc (NMPC) 8112230144 12/11/81 to T. A. Ippolito (NRC)
D.B. Vassallo (NRC) to D. P. Disc (NMPC) issuing Order modifying the 1/13/81 Order 1/19/82 Letter from C. V. Mangan (NMPC) 8301070265 01/05/83 to D. B. Yassallo (NRC)
Letter from D. B. Vassallo (NRC) 8301180521 01/07/83 to G. K Rhode (NMPC)
Letter from D. B. Vassallo (NRC) 8502040083 01/22/85 to B. G. Mooten (NMPC)
Letter C. D. Terry (NMPC) 8912050090 11/28/89 to NRC
- 3. YERIFICATION DOCUMENTS:
TITLE NUDOCS NO. DATE
'PLANT NMP-1 OOCKET NO(S). 50-220 PROJECT MANAGER Robert E. Martin TECHNICAL CONTACT J. Mauck USI NO. A-9 TITLE ATWS er 10 CFR 50.62 MPA NO. A-20 TAC NOS. 67506 ISSUES
SUMMARY
This USI was resolved in June 1984 with the publication of a final rule (10 CFR 50.62) to require improvements in plants to reduce the likelihood of failure of the reactor protection system (RPS) to shut down the reactor following anticipated transients and to mitigate the consequences of an anticipated transient without scram (ATWS) event.
The rule includes the following design-related requirements: 50.62(C)(1),
diverse and independent auxiliary feedwater initiation and turbine trip for all PWRs; 50.62(C)(2), diver se scram systems for CE and B8W reactors; 50 .62(C)(3) alternate rod injection (ARI) for BWRs 50.62(C)(4); standby liquid control system (SLCS) for BWRs; and 50.62(C)(5I, automatic trip of recirculation pumps under conditions indicative of an ATWS for BWRs. Information requirements and an implementation schedule are also specified.
IMPLEMENTATION AND STATUS
SUMMARY
(PLANT SPECIFIC):
By letter dated April 1, 1987, the licensee submitted information to
'demonstrate the adequacy of the ARI, SLCS and RPT. Additional information was submitted by the licensee regarding the ARI and RPT on July 6, 1988.
By letter dated August 31, 1988, the staff forwarded to the licensee a safety evaluation supporting the licensee submittals on compliance with the ATWS rule regarding the ARI and RPT.
On October 31, 1988, Tech Specs Amendment No. 101 was issued (in response to the licensee's application dated March 7, 1988, as supplemented on Apri 1 13, 1988). It revises the Tech Specs for the Liquid Poison System to satisfy the requirements of 10 CFR 50.62.
The licensee states orally that it considers the implementation of 10 CFR 50.62/USI A-9 to be complete by February 28, 1990. This is the date by which the licensee has verified the mixing concentration for the Liquid Poison System in order to implement the requirements of License Amendment no. 101.
'EFERENCES: NMP-1 A-9
- 1. RE UIREMENT DOCUMENTS:
TITLE NUDOCS NO. DATE NUREG-0460, and Supplements, , 03/80 "Anticipated Transients Without Scram for Light Water Reactors" Federal Register Notice 49 FR 26045 (10 CFR 50. 62) 06/26/84
- 2. IMPLEMENTATION DOCUMENTS:
TITLE NUDOCS NO. DATE Letter from T. E. Lempges (NMPC) 8704070373 04/01/87 to NRC Letter from C. D. Terry (NMPC) 8807120406 07/06/88 to NRC Letter from M. C. Haughey (NRC) 8809060097 08/31/88 to C. V. Mangan (NMPC)
Letter from M. F. Haughey (NRC) 8811090428 10/31/88 to C. V. Mangan (NMPC)
Letter from C. D. Terry (NMPC) 8912050090 11/28/89 to NRC
- 3. VERIFICATION DOCUMENTS:
TITLE NUDOCS NO. DATE
PLANT NMP-1 DOCKET NO(S). 50-220 PROJECT MANAGER Robert E. Martin TECHNICAL CONTACT K. Wichman US I NO. A-10 TITLE BWR Feedwater Nozzle Crackin MPA NO. 8-25 TAC NOS; 08499 and 72944 ISSUES
SUMMARY
This issue was resolved in November 1980 with the publication of NUREG-0619, "BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking." MPA B-25 was established by NRC's Division of Licensing for implementation purposes.
Inspections of operating BWRs conducted up to April 1978 revealed cracks in the feedwater nozzles of 20 reactor vessels. It was determined that cracking was due to high-cycle fatigue caused by fluctuations in water temperature within the vessel in the nozzle region.
By letter dated November 13, 1980, Darrell G. Eisenhut provided licensees with a copy of NUREG-0619. The 1'etter stated that NUREG-0619 provided the resolu-tion of the staff's generic technical activity USI A-10, which resulted from the inservice discovery of cracking in feedwater nozzles and control rod drive return line nozzles. NUREG-0619 describes the technical issues, General Electric and staff studies and analyses, and the staff's positions and require-ments. Licensees were required to respond, pursuant to 10 CFR 50.54(f), that they would meet implementation dates indicated in NUREG-0619.
Generic Letter 81-11 was subsequently issued to provide technical clarification to the November 13, 1980 letter, to clarify that it had been sent to PWR licensees for information only, and that no response was required from PWR licensees.
IMPLEMENTATION AND STATUS
SUMMARY
(PLANT SPECIFIC):
By letter dated December 29, 1980 the licensee responded to NUREG-0619 to confirm that they would meet the implementation dates indicated in NUREG-0619.
By letter dated July 10, 1981, the NRC accepted the December 29, 1980 proposed actions regarding the implementation of A-10 per NUREG-0619 and requested additional information. By letter dated October 29, 1981, the Commission notified the licensee that no further correspondence was necessary on the subject of A-10 implementation and the Commission was satisfied with the licensee's Commitment to meet the intent of NUREG-0619. The licensee indicates orally that the initial implementation of its responses to NUREG-0619 was completed on June 4, 1983 when the feedwater low flow control modification no. Nl-8269 was completed. This modification was directed at reducing the thermal cycling on the feedwater nozzles.
( ~
J NHP-1 A-10 In addition to the scope of the issue dealt with by USI A-10 the following actions were taken.
(a) By letter dated December 23, 1986, the licensee forwarded a report of the inservice inspection of the feedwater nozzles.
(b) By letter dated triarch 21, 1989, as supplemented tray 5, 1989, and by information presented in an April 18, 1989 meeting the licensee requested a deferment to the NUREG-0619 commitment to remove a feedwater nozzle sparger.
(c) By letter dated September 26, 1989, the staff forwarded to the licensee a safety evaluation in which it was concluded that there is reasonable assurance that the facility can be safely operated during the next two cycles with the feedwater nozzle "A" in its current condition.
(d) In addition, the licensee states that operational changes are being made to further reduce/eliminate thermal cycling on feedwater nozzles beginning with the startup from the current outage.
r
'EFERENCES'. NMP-1 A-10 RE UIREMENT DOCUMENTS:
TITLE NUDOCS NO. DATE Letter from D. Eisenhut transmitting NUREG-0619, "BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking,"
resolution of A-10 to licensees 11/13/80 Generic Letter 81-11, "BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking (NUREG-0619)" 02/20/81
- 2. IMPLEMENTATION DOCUMENTS:
TITLE NUDOCS NO. DATE Letter from N. P. Disc (NMPC) 8101050087 12/29/80 to D. G. Eisenhut (NRC)
Letter from T. A. Ippolito (NRC) 8107220366 07/10/81 to D. P. Disc (NMPC)
Letter from T. A. Ippolito (NRC) 8111200008 10/29/81 to D. P. Disc (NMPC)
Letter from C.Y. Mangan (NMPC) 8701020108 12/23/86 to T. E. Murley (NRC)
Letter from NMPC to NRC 8903280112 03/21/89 Letter from C. D. Terry (NMPC) 8905190123 05/05/89 to NRC Letter from M. L. Slosson (NRC) 8910030451 09/26/89 Letter C. D. Terry (NMPC) 8912050090 11/28/89 to NRC
- 3. YER IF I CATION DOCUMENTS:
TITLE NUDOCS NO. DATE
0
'LANT NMP-I DOCKET NO(S). 50-220 PROJECT MANAGER Robert E. Martin TECHNICAL CONTACT B. Elliott-USI NO. A-11 TITLE Reactor Vesse1 Materials Toe hoess MPA NO. A007 TAC NOS. 07445 ISSUES
SUMMARY
This. USI was resolved in October 1982 with the publication of NUREG-0744, "Pressure Vessel Material, Fracture Toughness.". NUREG-0744 was issued by Generic Letter 82-26 and provided only a methodology to satisfy the require-ments of 10 CFR Part 50, Appendix G. No licensee response to Generic Letter 82-26 was required.
Because of the remote possibility that nuclear reactor pressure vessels designed to the ASME Boiler and Pressure Vesse'l Code would fail, the design of nuclear facilities does not provide protection against reactor vessel failure.
Prevention of reactor vessel failure depends primarily on maintaining the reactor vessel material fracture toughness at levels that will resist brittle fracture during plant operation. At service times and operating conditions typical of current operating plants, reactor vessel fracture toughness properties provide adequate margins of safety against vessel failure; however, as plants accumulate more and more ser vice time, neutron irradiation reduces the material fracture toughness and initial safety margins.
Appendix G to 10 CFR Part 50 requires that the Charpy upper shelf energy throughout the life of the vessel be no less than 50 ft-lb unless demonstrated that lower values will provide margins of safety against failure it is equivalent to those provided by Appendix G of the ASME code. USI A-ll was initiated to address the staff's concern that some vessels were projected to have beltline materials with Charpy upper shelf energy less than 50 ft-lb.
NUREG-0744 provides a method for evaluating reactor vessel materials when their Charpy upper shelf energy is predicted to fall below 50 ft-lb. Plants will use the prescribed method when analysis of irradiation damage predicts that the charpy upper shelf energy is below 50 ft-lb.
IMPLEMENTATION AND STATUS
SUMMARY
PLANT SPECIFIC:
Licensee states that vessel material properties are inspected in accordance with TS 4.2.2.b. Licensee has not identified a condition approaching the 50 ft-lb level and therefore has no need to utilize NUREG-0744 methods.
) ~
REFERENCES:
NMP-1 A-11
- 1. RE UIREMENT DOCUMENTS:
TITLE NUDOCS NO. DATE NUREG-0744, Revision 1, "Pressure 10/82 Vessel Material Fracture Toughness" Generic Letter 82-26, "Pressure Vessel Material Fracture Toughness" 11/12/82
- 2. IMPLEMENTATION DOCUMENTS:
TITLE NUDOCS NO. DATE Letter from C. D. Terry (NMPC) 8912050090 11/28/89 to NRC
- 3. VERIFICATION DOCUMENTS:
TITLE NUDOCS NO. DAVE
r I ~
PLANT NMP-1 DOCKET NO(S). 50-220 PROJECT MANAGER Roberr E. Martin TECHNICAL CONTACT D. Thatcher OSI NO. A-17 TITLE S stems Interactions in Nuclear Power Plants NPA NO. TAC NOS.
ISSUES
SUMMARY
Generic Letter (GL) 89-18, dated September 6, 1989, was sent to all power reactor licensees and constitutes the resolution of USI A-17. The generic letter did not require any licensee actions.
GL 89-18 had two enclosures which (a) outlined the bases for the resolution of USI A-17, and (b) provided five general lessons learned from the review of the overall systems interaction issue. The staff anticipated that licensees would review this information in other programs, such as the Individual Plant Examination ( IPE) for Severe Accident Yulnerabilities. Specifically, the staff expected that insights concerning water intrusion and flooding from internal sources, as described in the appendix to NUREG-1174, would be considered in the IPE program. Also considered in the resolution of this USI was the expectation that licensees would continue to review information on events at operating nuclear power plants in accordance with the requirements of TMI Task Action Plan Item I.C.5 (NUREG-0737).
IMPLEMENTATION AND STATUS
SUMMARY
PLANT SPECIFIC:
Per guidance of GL 89-21 on the status of USIs, no licensee actions were required in response to GL 89-18 and accordingly, the licensee stated that no actions were taken.
As background is noted that information By letter on flooding issues beyond the scope of GL 89-18 dated August 3, 1972, the licensee was requested to it review their facility following an event at guad Cities Unit 1 where flooding caused degradation of some of the engineered safety features. By letter dated September 29, 1972, the licensee responded to the Augsut 3, 1972 letter and stated that there was no flooding potential for existing engineered safety system.
V I
REFERENCES:
NMP-1 A-17
- 1. RE UIREMENT DOCUMENTS:
TITLE NUDOCS NO. DATE Generic Letter 89-18 09/06/89 NUREG-1174 "Evaluation of May 1989 Systems Interactions in Nuclear Power Plants" NUREG-1229 "Regulatory Analysis August 1989 for Resolution of USI A-17" NUREG/CR-3922 "Survey and January 1985 Evaluation of System Interaction Events and Sources" NUREG/CR-4261 "Assessment of June 1986 System Interaction Experience in Nuclear Power Plants" NUREG/CR-4470 "Survey and August 1986 Evaluation of Vital Instrumentation and Control Power Supply Events" NRC Letters to Licensees 9/72 Informing Licensees of Staff Concerns Regarding Potential Failure of Non-Category I Equipment
- 2. IMPLEMENTATION DOCUMENTS:
TITLE NUDOCS NO. DATE Letter from T. J. Brosnan (NMPC) 09/29/72 to D. J.Skovholt (NRC)
Letter from C. D. Terry, (NMPC) 8912050090 11/28/89
- 3. VERIFICATION DOCUMENTS:
TITLE NUOOC NO. DATE
PLANT NMP-1 DOCKET NO(S). 50-220 PROJECT HANAGER Robert E. Martin TECHNICAL CONTACT P. Shemanski USI NO. A-24 TITLE gualification of Class 1E E ui ment HPA NO. TAC NOS. 42476 ISSUES
SUMMARY
This USI was resolved in July 1981 with the publication of NUREG-0588, Revision 1, "Interim Staff Position on Environmental gualification of Safety-Related Electrical Equipment." Part I of the report is the original NUREG-0588 that was issued for comment; that report, in conjunction with the Division of Operating Reactor (DOR) Guidelines, was endorsed by a Commission Memorandum and Order as the interim position on this subject until "final" positions were established in rule making. On January 21, 1983 the Commission amended 10 CFR 50.49 (the rule), effective February 22, 1983, to codify existing qualification methods in national standards, regulatory guides, and certain NRC publications, including NUREG-0588.
The rule is based on the DOR Guidelines and NUREG-0588. These provide guidance on (a) how to establish environmental service conditions, (b) how to select methods which are considered appropriate for qualifying the equipment in different areas of the plant, and (c) such other areas as margin, aging, and documentation. NUREG-0588 does not address all areas of qualification; supplement, in selected areas, the provisions of the 1971 and 1974 versions of it does IEEE Standard 323. The rule recognizes previous qualification efforts completed as a result of Commission Memorandum and Order CLI-80-21 and also reflects different versions IEEE 323 dependent on the date of the construction permit Safety Evaluation Report (SER . Therefore, plant-specific requirements may vary in accordance with the rule.
In summary, the resolution of A-24 is embodied in 10 CFR 50.49. A measure of whether each licensee has implemented the resolution of A-24 may therefore be found in the determination of compliance with 10 CFR 50.49. This was addressed by 72 SERs for operating plants issued shortly after publication of the rule and subsequently in operating license reviews pursuant to Standard Review Plan Section 3.11. This was further addressed by the first-round environmental qualification inspections conducted by the NRC.
IMPLEMENTATION AND STATUS
SUMMARY
(PLANT SPECIFIC):
By letter dated May 31, 1984, the licensee forwarded to the NRC the status of the Environmental gualification Program for NMP-1.
On January 10, 1985, the NRC issued a Safety Evaluation stating that "NHPC's Equipment gualification Program is in compliance with the requirements of 10 CFR 50.49, that the proposed resolution for each of the environmental qualification deficiencies identified for Nine Mile Point, Unit No. 1 is acceptable and that continued operation of Nine Mile Point, Unit No. 1 will not present undue risk to the public health and safety."
The schedule for implementation of 50.49 requirements for certain items was extended by several NRC letters including the one of March 15, 1985 which addressed the emergency condensen isolation valve actuators. The licensee states orally that initial implementation of all 50.49 requirements was complete on July 9, 1986 when these valves were made operable.
REFERENCES:
NMP-1 A-24
- 1. RE UIREMENT DOCUMENTS:
TITLE NUDOCS NO. DATE DOR "Guidelines for Evaluating Environmental qualification of Class 1E Electrical Equipment in Operating Reactors" NUREG-0588, "Interim Staff Position on Environmental gualification of .
Safety Related Electrical Equipment" 12/79 Commission Memorandum and Order, CLI-80-21, on DOR Guidelines and NUREG-0588 05/23/80 NUREG-0588, Revision 1 07/81 10 CFR 50.49 (48 FR 2730-2733) 01/21/83 Standard Review Plan 3.11, Environmental qualification of Mechanical and Electrical Equipment 07/81
- 2. IMPLEMENTATION DOCUMENTS:
TITLE NUDOCS NO. DATE Letter C. V. Mangan (NMPC) 05/31/84 to D. B. Yassallo (NRC)
Letter D. B. Vassallo (NRC) 8502010710 01/10/85 to B. G. Hooten (NMPC)
Letter H. R. Denton (NRC) to B. G. Hooten (NMPC) 8504030322 3/15/85
- 3. VERIFICATION DOCUMENTS:
TITLE NUDOCS NO. DATE
.4 PLANT NMP-1 DOCKET NO(S). 50-220 PROJECT MANAGER Robert E. Martin TECHNICAL CONTACT J. Wermiel USI NO. A-36 TITLE Control of Heav Loads Phases I St II MPA NO. C-10 C-15 TAC NOS. 08063 ISSUES
SUMMARY
This .USI was resolved in July 1980 with the publication of NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants," and Standard Review Plan (SRP) Section 9.1.5. The staff established MPAs C-10 and C-15 for the implementation of Phases I and II, respectively, of the resolution of this issue at operating plants.
In nuclear power plants, heavy loads may be handled in several plant areas.
these loads were to drop in certain locations in the plant, they may impact If spent fuel, fuel in the core, or equipment that may be required to achieve safe shutdown and continue decay heat removal. USI A-36 was established to systematically examine staff licensing criteria and the adequacy of measures in effect at operating plants, and to recommend necessary changes to ensure the-safe handling of heavy loads. The guidelines proposed in NUREG-0612 include definition of safe load paths, use of load handling procedures, training of crane operators, guidelines on slings and special lifting devices, periodic inspection and maintenance for the crane, as well as various alternatives.
By Generic Letters dated December 22, 1980, and February 3, 1981 (Generic Letter 81-07), all utilities were requested to evaluate their plants against the guidance of NUREG-0612 and to provide their submitta'ls in two parts: Phase I (six month response) and Phase II (nine month response). Phase I responses wer e to address Section 5.1.1 of NUREG-0612 which covered the following areas:
- 1. Definition of safe load paths
- 2. Development of load handling procedures
- 3. Periodic inspection and testing of cranes qualifications, training and specified conduct of operators
- 5. Special lifting devices should satisfy the guidelines of ANSI N14.6.6.
- 6. Lifting devices that are not specially designed should be installed and used in accordance with the guidelines of ANSI B30.9
- 7. Design of cranes to ANSI B30. 2 or CMAA-70 Phase II responses were to address Sections 5.1.2 thru 5.1.6 of NUREG-0612 which covered the need for electrical interlocks/mechanical stops, or alternatively, single-failure-proof cranes or load drop analyses in the spent fuel pool area (PWR), containment building (PWR), reactor building (BWR), other areas and the specific guidelines for single-failure-proof handling systems.
As stated in Generic Letter 85-11, "Completion of Phase II of 'Control of Heavy Loads at Nuclear Power Plants' NUREG-0612," all licensees have completed the requirement to perform a review and submit a Phase I and a Phase II report.
Based on the improvements in heavy loads handling obtained from implementation of NUREG-0612 (Phase I), further action was not required to reduce the risks associated with the handling of heavy loads. Therefore, a detailed Phase II review of heavy loads was not necessary and Phase II was considered completed.
While not a requirement, NRC encouraged the implementation of any actions identified in Phase II regarding the handling of heavy loads that were considered appropriate.
IMPLEMENTATION AND STATUS
SUMMARY
PLANT SPECIFIC):
The licensee responded to the NRC December 22, 1980 letter by several letters between May 22, 1981 and November 25, 1985. For the purpose of USI A-36, the issue was completed by the last submittal made by the licensee on January 18, 1985. The staff provided acceptance of the licensee's program by a Safety Evaluation issued on March 5, 1985.
REFERENCES:
NMP-1 A-36
- 1. RE UIREMENT DOCUMENTS:
TITLE NUDOCS NO. DATE Letter, Darrell G. Eisenhut, NRC, to all licensees, applicants for OLs and holders of CPs transmitting NUREG-0612 and staff positions 12/22/80 Generic Letter 85-11, Hugh L.
Thompson, NRC, to all licensees for Operating Reactors, "Completion of Phase II of 'Control of Heavy Loads at Nuclear Power Plants'UREG-0612" 06/28/85
- 2. IMPLEMENTATION DOCUMENTS:
TITLE NUDOCS NO. DATE Letter D. P. Disc (NMPC) 8108040118 07/28/81 to D. G. Eisenhut Letter D. P. Disc (NMPC) 109290460 9/22/81 to D. G. Eisenhut Letter T. E. Lemoges (NMPC) 8208050470 08/01/82 to D. Eisenhut Letter T. E. Lemoges (NMPC) 8205040529 04/29/82 to D. Ei,senhut Letter T. E. Lemoges (NMPC) 8208050470 08/01/82 to D. Eisenhut Letter T. E. Lemoges (NMPC) 8310040597 09/30/83 to'. Eisenhut Letter C. Y. Mangan (NMPC) 8407310223 07/26/84 to D. C. Vassallo (NRC)
Letter T. E. Lempges (NMPC) 8501230433 01/18/85 to Vassallo (NRC)
Letter D. B. Vassallo (NRC) 8503210434 03/05/85 to B. G. Hooten (NMPC) 3 VERIFICATION DOCUMENTS:
TITLE NUDOCS NO. DATE
0 C~
PLANT NMP-1 OOCKET NO(S). 50-220 PROJECT MANAGER Robert E. Martin TECHNICAL CONTACT J. Kudrick OSI NO. A-39 TITLE Determination of SRV Pool Dynamic Loads and Tem erature Limits NPA NO. TAC NOS.
ISSUES
SUMMARY
This USI was resolved with the publication of Standard Review Plan (SRP)
Section 6.2.1.1.C, in October 1982. In addition, NUREGs 0763, 0783 and 0802 were issued for Mark I, Mark II, and Mark III containments, respectively.
BWR plants are equipped with safety/relief valves (SRVs) to protect the reactor from overpressurization. Plant operational transients, such as turbine trips, will actuate the SRV. Once the SRV opens, the air column within the partially submerged discharge line is compressed by the high-pressure steam released from the reactor. The compressed air discharged into the suppression pool produces high-pressure bubbles. Oscillatory expansion and contraction of these bubbles create hydrodynamic loads on the containment structures, piping, and equipment inside containment.
NUREG-0802 presents the results of the staff's evaluation of SRV loads. The evaluation, however, is limited to the quencher devices used in Mark II and III containments. With respect to Mark I containments, the SRV acceptance criteria are presented in NUREG-0661, "Safety Evaluation Report, Mark I Containment and Long-Term Program," and are dealt with as part of USI A-7.
SRP Section 6.2.1.1.C addresses the applicable review cr iteria, since all Mark II and III containment designs are understood to have completed their operating license (OL) reviews subsequent to resolution of this USI and reflection of the resolution in the SRP.
IMPLEMENTATION AND .STATUS
SUMMARY
PLANT SPECIFIC):
As stated above, the SRV acceptance criteria are presented in NUREG-0661 for Nine Mile Point Unit 1, (Mark I containment) and are dealt with as part of USI A-7.
t4
REFERENCES:
NMP-1 A-39
- 1. RE UIREMENT DOCUMENTS:
TITLE NUDOCS NO. DATE SRP 6.2.1.1.C, Pressure Suppression Type BWR Containments NUREG-0802, "Safety/Relief Valve quencher Loads: Evaluation for 8WR Mark II and III Containments, Generic Technical Activity A-39" 1982 NUREG-0661, "Safety Evaluation Report- 7/80 Mark I Long Term Program"
- 2. IMPLEMENTATION DOCUMENTS:
TITLE NUDOCS NO. DATE Letter C. D. Terry (NMPC) 8912050090 11/28/89 to NRC
- 3. VERIFICATION DOCUMENTS:
TITLE NUDOCS NO. DATE
- The applicable SRP revision number would depend on the date of the evaluation for each specific plant.
~ ~
I+,
~ ~
PLANT NMP-I DOCKET NO(S). 50-220 PROJECT MANAGER Robert E. Martin TECHNICAL CONTACT W. Koo--
USI NO. A-42 TITLE Pi e Cracks in Roi1in Water Reactors NPA NO. 8-06 TAC NOS. 69147 ISSUES
SUMMARY
This USI was resolved in February 1981 with the publication of NUREG-0313, Revision 1, "Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping." That NUREG document was issued to all holders of BWR operating licenses or construction permits and to all applicants for BWR operating licenses. The staff established MPA B-05 for implementation of the resolution at operating plants.
Pipes have cracked in the heat-affected zones of welds in prima y system piping in BWRs since mid-1960. These cracks have occurred mainly in Type 304 stainless steel, which is the type used in most .operating BWRs. The major problem is recognized to be intergranular stress corrosion cracking (IGSCC) of austenitic stainless steel components that have been made susceptible to this failure by being "sensitized," either by post-weld heat treatment or by sensitization of a narrow heat affected zone near welds.
"Safe ends" that have been highly sensitized by furnace heat treatment while attached to vessels during fabrication were found to be susceptible to IGSCC in the late 1960s. Most of the furnace-sensitized safe ends in older plants have been removed or clad with a protective material, and only a few BWRs still have furnace-sensitized safe ends in use. Most of these, however, are in smaller diameter lines.
Cracks reported before 1975 occurred primarily in 4-inch-diameter recirculation loop bypass lines and in 10-inch-diameter core spray lines. Cracking is most often detected during inservice inspections using ultrasonic test techniques.
Some piping cracks have been discovered as a result of primary coolant leaks.
NUREG-0313, Revision 1 provided the NRC staff's revised acceptable methods for reducing the IGSCC susceptibility of BWR code class 1, 2, and 3 pressure boundary piping of sizes identified above and safe ends. In addition, it provided the requirements for augmented inservice inspection of piping with nonconforming materials.
As a result of further IGSCC degradations in larger piping, the staff provided licensees with additional requirements in several NRC communications (i.e.,
Bulletins 82-03, 83-2, and 84-11). The long-term resolution of IGSCC in BWR piping (including the scope of A-42) was provided in NUREG-0313, R'evision 2 which was transmitted to all holders of BWR operating licenses via Generic Letter 88-01.
I NMP-1 A-42 IMPLEMENTATION AND STATUS
SUMMARY
(PLANT SPECIFIC):
By letter dated July 1, 1981, July 8, 1981, August 6, 1982, December 20, 1982, and February 17, 1983, the licensee provided information in response to Generic Letter 81-04.
The safety evaluation on implementation of NUREG-0313, Rev. 1 was issued by the staff on 06/06/84. For the purpose of responding to the USI A-42, as defined by the scope of GL 81-04/NUREG 0313, Rev. 1, this safety evaluation is considered to reflect the resolution of the USI by the licensee. The SE notes that a certain NRC contractor's review concluded that not all of the NUREG-0313, Rev 1 guidelines had been found to be met. However that review did not encompass later plant specific actions taken in at least four areas.
Acknowledging these actions and the replacement of recirculation piping, the staff's June 6, 1984 transmittal letter then provided the evaluation of NUREG-0313 Rev. 1 compliance for information and assistance purposes in the licensee's preparation of responses to staff initiatives (namely GL 84-11) beyond the scope of USI A-42. On this basis the staff's June 6, 1984 safety evaluation is considered to reflect the programmatic conclusion of USI A-42 issues by the licensee and the staff. This date also corresponds to the resolution of MPA B-05 on 06/84.
Going well beyond the scope of USI A-42, it is noted that the licensee responded to GL 88-01 by letter dated July 28, 1988 as supplemented by letter dated August 25, 1989 and September 6, 1989. Technical Specification Amendment No. 107 which incorporates the requirements of GL 88-01 was issued on July 7, 1989. The licensee plans to implement the requirements of GL 88-01 prior to the next refueling outage per the July 28, 1989 letter. The licensee's submittals are under review by the staff.
To recap the discussion presented above for the 24 GWRs that were operating when GL 81-04 was issued (February 1981, the implementation date for this USI has been determined to be the date of the letter transmitting the staff's evaluation of the licensee's response to GL 81-04. Hence, the date this USI was implemented for NMP-1 is June 6, 1984.
'EFERENCES: NMP-1 A-42
- 1. RE UIREMENT DOCUMENTS:
TITLE NUDOCS NO. DATE NUREG-0313, Revision 1, "Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping," 07/80 Generic Letter 81-04, "Implemen- 02/26/81 tation of NUREG-0313, Rev. 1 for Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping (Generic Task A-42)"
- 2. IMPLEMENTATION DOCUMENTS:
TITLE NUDOCS NO. DATE Letter D. P. Disc (NMPC) to 8107070307 07/01/81 D. G. Eisenhart (NRC)
Letter T. E. Lempges (NMPC) 8208130195 08/06/82 to D. G. Eisenhart (NRC)
Letter C. V. Mangan (NMPC) 8212270265 12/20/82 to D. B. Vassallo (NRC)
Letter C. V. Mangan (NMPC) 8302230216 02/17/83 to D. 8. Vassallo (NRC)
Letter D. B. Vassallo (NRC) 8406200403 06/06/84 to B. G. Hooten (NMPC)
Letter C. D. Terry (NMPC) 8808040296 07/28/88 to NRC Letter L. Burkhardt (NMPC) 8909010320 08/25/89 to NRC Letter L. Burkhardt (NMPC) 8909200156 09/06/89 to NRC Letter C. D. Terry (NMPC) 8912050090 11/28/89 to NRC
- 3. VERIFICATION DOCUMENTS:
TITLE NUDOCS NO. DATE
I C 4
a a
PLANT NMP-1 DOCKET NO(S). 50-220 PROJECT MANAGER Robert E. Martin TECHNICAL CONTACT P. Gill USI NO. A-44 TITLE Station Blackout NPA NO. A-022 TAG NOS. 66570 ISSUES <<
SUMMARY
This USI was resolved in June 1988 with the publication of a new rule (10 CFR 50.63) and Regulatory Guide 1. 155.
Station blackout means the loss of offsite ac power to the essential and nonessential electrical buses concurrent with turbine trip and the unavailability of the redundant onsite emergency ac power systems. HASH-1400 showed that station blackout could be an important risk contributor, and operating experience has indicated that the reliability of ac power systems might be less than originally anticipated. For these reasons station blackout was designated as a USI in 1980. A proposed rule was published for comment on March 21, 1986. A final r,ule, 10 CFR 50.63, was published on June 21, 1988 and became effective on July 21, 1988. Regulatory Guide 1.155 was issued at the same time as the rule and references an industry guidance document, NUMARC-8700. In order to comply with the A-44 resolution, licensees will be required to:
maintain onsite emergency ac power supply reliability above a minimum level develop procedures and training for recovery from a station blackout determine the duration of a station blackout that the plant should be able to withstand use an alternate qualified station blackout ac power source, if available, to cope with a evaluate the plant's actual capability to withstand. and recover from a station blackout backfit hardware modifications if necessary to improve coping ability Section 50.63(c)(1) of the rule required each licensee to submit a response including the results of a coping analysis within 270 days from issuance of an operating license or the effective date of the rule, whichever is later.
IMPLEMENTATION AND STATUS
SUMMARY
(PLANT SPECIFIC):
By letter dated April 13, 1989, the licensee submitted a response as required in Section 50.63(c)(i) of the rule. In its response the licensee stated that the Class IE batteries evaluated in accordance with NUMARC 8700 guidelines were determined to be inadequate to meet station blackout loads for four hours.
The licensee is performing an analysis to determine if additional load stripping will allow Nine Mile Point Unit I to meet station blackout loads for four hours or whether additional capacity will be required. The licensee has not yet finalized its plans and therefore, the results and plans for meeting station blackout loads are not yet available.
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'EFERENCES: NMP-1 A-44
- 1. RE UIREMENT DOCUMENTS:
TITLE NUDOCS NO. DATE 10 CFR 50.63, "Loss of A11 A1ternating Current Power" 06/21/88 Regu1atory Guide 1.155, "Station 81ackout" 08/88
- 2. IMPLEMENTATION DOCUMENTS:
TITLE NUDOCS NO. DATE Letter C. D. Terry (NMPC) 8904240053 04/13/89 to NRC
- 3. VERIFICATION DOCUMENTS:
TITLE NUDOCS NO. DATE
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en,
'LANT NMP-1 DOCKET NO(S). 50-220 PROJECT MANAGER Robert E. Martin TECHNICAL CONTACT P. Y. Chen USI NO. A-46 TITLE Seismic qualification of Equipment in Operating Plants NPA NO. 8-106 TAG NOS. 69461 ISSUES
SUMMARY
USI A-46 was resolved with the issuance of GL 87-02 on February 19, 1987, which endorsed the approach of using the seismic and test experience data proposed by the Seismic gualification Utility Group (SHRUG) and Electric Power Research Institute (EPRI). This approach was endorsed by the Senior Seismic Review and Advisory Panel (SSRAP) and approved by the NRC staff.
The scope of the review was narrowed to equipment required to bring each affected plant to hot shutdown and maintain it there for a minimum of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
The review includes a walkthrough of each plant which is required to inspect equip-ment. Evaluation of equipment will include: (a) adequacy of equipment anchorage; (b) functional capability of essential relays; (c) outliers and deficiencies (i.e., equipment with non-standard configurations); and (d) seismic systems interation.
As an outgrowth of the Systematic Evaluation Program (SEP), the need was identified for reassessing design criteria and methods for the seismic quali-fication of mechanical equipment and electrical equipment. Therefore, the seismic qualification of the equipment in operating plants must be reassessed to ensure the ability to bring the plant to a safe shutdown condition when subject to a seismic event. The objective of this issue was to establish an explicit set of guidelines that could be used to judge the adequacy of the seismic qualification of mechanical and electrical equipment at operating plants in lieu of attempting to backfit current design criteria for new plants.
Generic Letter 87-02 with associated guidance, required all affected utilities to evaluate the seismic adequacy of their plants. The specific requirements and approach for implementation are being developed jointly by SgVG and the staff on a generic basis before individual member utilities proceed with plant-specific implementation.
IMPLEMENTATION AND STATUS
SUMMARY
PLANT SPECIFIC :
For NMP-1, the licensee performed the seismic review and trial walkdown of NMP-1 in early 1988. By letter dated September 23, 1988, the licensee advised NRC that it will complete the work scope contained in Revision 0 of the GIP by the end of the next planned refueling outage which is dependent on NMP-1 date of return to operation (scheduled for early 1990).
The next step in the resolution of this issue is dependent on staff action.
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'EFERENCES: NMP-1 A-46
- 1. RE UIREMENT DOCUMENTS:
TITLE NUDOCS NO. DATE Generic Letter 87-02, "Verifi-cation of Seismic Adequacy of Mechanical and Electric Equipment in Operating Reactors" 02/19/87 NUREG-1211, "Regulatory Analysis for Resolution of Unresolved Safety Issues A-46..." 02/87 NUREG-1030, "Seismic Qualification of Equipment in Operating Plants, Unresolved Safety Issue A-46" 02/87 Letter attached with "Generic Safety Evaluation Report on SQUG GIP, Revision 0," from L. Shao (NRC) to Neil Smith (SQUG) 07/29/88 "Generic Implementation Procedure (GIP) for Seismic Verification of Nuclear Plant Equipment," Revision I 12/88 "Generic Implementation Procedure (GIP for Seismic Verification of Nuclear Plant Equipment," Revision 0 06/88
- 2. IMPLEMENTATION DOCUMENTS:
TITLE NUDOCS NO. DATE Letter C. D. Terry (NMPC) to NRC 2809280263 09/23/88
- 3. VERIFICATION DOCUMENTS:
TITLE NUDOCS NO. DATE
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PLANT NMP-I DOCKET NO(S). 50-220 PROJECT MANAGER Robert E. Martin TECHNICAL CONTACT J. Mauck USI NO. A-47 TITLE Safety Imp1ication of Controi Systems in LNR Nuclear Power Plants MPA NO. 6113 TAC NOS. 74966 ISSUES
SUMMARY
USI A-47 was resolved September 20, 1989, with the publication of (GL) 88-19.
Generic'etter The generic letter states:
"The staff has concluded that all PWR plants should provide automatic steam generator overfill protection, all BWR plants should provide automatic reactor vessel overfill protection, and that plant procedures and technical specifications for all plants should include provisions to verify periodically the operability of the overfill protection and to assur e that automatic overfill protection is available to mitigate main feedwater overfeed events during reactor power operation. Also, the system design and setpoints should be selected with the objective of minimizing inadvertent trips of the main feedwater system during plant startup, normal operation, and protection system surveillance. The Technical Specifications recommenda-tions are consistent with the criteria and the risk considera-tions of the Commission Interim Policy Statement on Technical Specification Improvement. In addition, the staff recommends that all BWR recipients reassess and modify, if needed, their operating procedures and operator training to assure that the operators can mitigate reactor vessel overfill events that may occur via the condensate booster pumps during reduced system pressure operation."
Also, page 2 of the generic letter provides for additional actions for CE and B&W plants. The generic letter provides amplifying guidance for licensees.
The generic letter requires that licensees provide NRC with their schedule and commitments within 180 days of the letter's date. The implementation schedule for actions on which commitments are made should be prior to startup after the first refueling outage, but no later than the second refueling outage, beginning 9 months after receipt of the letter.
IMPLEMENTATION AND.STATUS
SUMMARY
(PLANT SPECIFIC :
NMP-1 is current1y evaluating the requirements of GL 89-19 and is expected to provide its response by March 19, 1990.
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C 'EFERENCES: NMP-1 P
A-47
- 1. RE UIREMENT DOCUMENTS TITLE NUDOCS NO. DATE Generic Letter 89-19 09/20/89 "Request for Action Related to Resolution of USI A-47" NUREG-1217 "Evaluation of Safety June 1989 Implications of Control Systems in LWR Nuclear Power Plants" NUREG-1218 "Regulatory Analysis July 1989 for Resolution of USI A-47"
- 2. IMPLEMENTATION DOCUMENTS:
TITLE NUDOCS NO. DATE Letter C. D. Terry (NMPC) 8912050090 11/28/89 to NRC
- 3. VERIFICATION DOCUMENTS:
TITLE NUDOCS NO. DATE
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'LANT NMP-1 DOCKET NO(S). 50-220 PROJECT MANAGER Robert E. Martin TECHNICAL CONTACT J. Kudrick USI NO. A-48 TITLE Hydrogen Control Measures and Effects of Hydrogen Burns on Safety E ui ment MPA NO. TAC NOS.
ISSUES
SUMMARY
The NRC staff concluded April 19, 1989, that USI A-48 is resolved, as stated in SECY 89-122.
USI A-48 was initiated's a result of the large amount of hydrogen generated and burned within containment during the Three Mile Island (TMI) accident.
This issue covers hydrogen control measures for recoverable degraded core accidents for all BWRs and those PWRs with ice condenser containments.
Extensive research in this area has led to significant revision of the Com-mission's hydrogen control regulations, given in 10 CFR 50.44, published December 2, 1981.
10 CFR 50.44 requires inerting of BWR Mark I and Mark II containments as a method for hydrogen control. The BWR Mark I and Mark II reactor containments have operated for a number of years with an inerted atmosphere (by addition of an inert gas, such as nitrogen) which effectively precludes combustion of any hydrogen generated. USI A-48 with respect to BWR Mark I and II containments is not only resolved but understood to be fully implemented in the affected plants.
The rule for BWRs with Mark III containments and PWRs with ice condenser containments was published on January 25, 1985. The rule required that these plants be provided with a means for controlling the quantity of hydrogen produced, but did not specify the control method. In addition, the task action plan for USI A-48 provided for plant-specific reviews of lead plants for reactors with Mark III and ice condenser containments. Sequoyah was chosen as the lead plant for ice condenser containments and Grand Gulf for Mark III containments. Both of the lead plant licensees chose to install igniter-type systems which would burn the hydrogen before it reached threatening concentra-tions within the containment. Final design igniter systems have been installed not only in both lead plants, Sequoyah and Grand Gulf, but in all other ice condenser and Mark III plants as well. The staff's safety evaluations of the final analyses required to be submitted by these licensees by the rule are scheduled for completion in 1989.
Large dry PWR containments were excluded from USI A-48 because they have a
, greater ability to accommodate the large quantities of hydrogen associated with a recoverable degraded core accident than the smaller Mark I, II, III and ice condenser containments. However, this issue has continued to be considered and, in 1989, hydrogen control for large dry PWR containments was identified as a high-priority Generic Issue (GI) 121. The resolution of GI 121 is being actively pursued in close coordination with more recent research findings.
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NMP-I A-48 ISSUES
SUMMARY
(CONT.):
The NRC staff has concluded that USI-A-48 is resolved as stated in SECY 89-122. If interested, the report should be consulted for further details regarding the relationship of A-48 to other ongoing hydrogen activities.
IMPLEMENTATION AND STATUS
SUMMARY
(PLANT SPECIFIC):
Containment inerted. Containment Atmosphere Dilution System (CAD) added per modification Nl-72-03. Safety Evaluation documenting NMP-1 complies with 10 CFR 50.44(c)(3)(ii) issued on April 29, 1985.
The licensee states orally that the capability to inert NMP-1 has existed since the beginning of its operating life. Therefore the date of the Provisional Operating License, August 22, 1969, is utilized as the initial implementation date for inerting of the containment.
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REFERENCES:
- 1. RE UIREMENT DOCUMENTS:
TITLE NUDOCS NO. DATE 10 CFR 50.44, Standards for 12/81 Combustible Gas System in Light-Mater-Cooled Power Reactors SECY-89-122, Resolution of USI A-48, "Hydrogen Control Measures and Effects of Hydrogen Burns on Safety Equipment" 04/19/89
- 2. IMPLEMENTATION DOCUMENTS:
TITLE NUDOCS NO. DATE Letter C. D. Terry, (NMPC) to NRC 11/28/89
- 3. VERIFICATION DOCUMENTS:
TITLE NUDOCS NO. DATE
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~Pa~~~ No. 1 02/06/90 LISTING OF INCOtlPLETE USI DATA FOR INPUT FRON PROJECT tlANAGERS ISSUE ISSUE DESCRIPTIVE HAttE IttPLENEHT INPLEtlENT LICEHSEE COtNEHT STAFF CONNENT HUNBER DATE STATUS lf PLANT MANE: HINE NILE POIHT 1 A-01 MATER HAtlNER / / NC SSFI A-02 ASYNNETRIC BLQMDOMN LOADS OH / / H/A PMR OHLY REACTOR PRIttARY COOLANT SYSTENS A-03 MESTIHGHOUSE STEAN 6EHERATOR TUBE / / N/A MESTIN6HOUSE ONLY INTE6RITY A-04 CE STEAN 6ENERATOR TUBE INTE6RITY H/A CE PLANTS ONLY A-05 B4M STEAN GENERATOR TUBE H/A BRM PLANTS ONLY IHTE6RITY A-06 NARK I SHORT-TERN PR06RAN 02/28/78 C DELTA P CONTROL A-07 NARK I LONG-TERN PR06RAN 06/13/84 C A-08 NARK II CONTAINMENT POOL DYHANIC / / H/A NK II BMR ONLY A-09 A-10 LOADS ATMS LONG-TERM PROGRAN BMR FEEDMATER NOZZLE CRACKIH6 06/04/83 C i% ~
A-11 REACTOR VESSEL NATERIALS / / NC TOUGHHESS A-12 FRACTURE TOU6HNESS OF STEAtl / / H/A CP AFTER 83 ONLY GENERATOR AHD REACTOR COOLANT PUttP SUPPORTS A-17 SYSTENS INTERACTION / / HC HO REQUIRENENTS A-24 OUALIFICATIOH OF CLASS 1E 07/09/86 C EXTENSION GRANTED SAFETY-RELATED EQUIPtlEHT A-26 REACTOR VESSEL PRESSURE TRANSIENT / / H/A PllR ONLY PROTECTION A-31 RHR SHUTDOMN REQUIREMENTS / / H/A HEM PLANTS ONLY.
A-36 CONTROL OF HEAVY LOADS NEAR SPENT 01/18/85 C ENDED SR'L-85-11 FUEL A-39 DETERNIHATIOH OF SAFETY RELIEF / / NC SEE A-07 VALVE POOL DYNANIC LOADS AHD TEttPERATURE LINITS A-40 SEISNIC DESIGN CRITERIA- / / NC SUBSUNNED BY A-46 SHORT-TERN PROGRAN A-42 PIPE CRACKS IH BOILIH6 MATER 06/06/84 C 6L-88-01 H/A REACTORS A-43 COHTAINtiENT ENER6EHCY SUNP / / HC INFO ONLY PERFORNANCE A-44 STATION BLACKOUT 12/31/92 I SER 12/31/90 A-45 SHUTDOMN DECAY HEAT REttOVAL / / NC IPE SUBSUNED BY SEVERE ACC REQUIRENEHTS A"46 SEISNIC OUALIFICATIOH OF / / I REO UNDER DEVEL EOUIPNENT IH OPERATIH6 PLANTS A-47 SAFETY INPLICATIOHS OF COHTROL 03/19/90 E HEM REOUIRENENTS A-48 SYSTEMS 8 ~~/ay<
HYDROGEN CONTROL NEASURES AHD INERTED EFFECTS OF HYDR06EH BURNS ON SAFETY EOUIPllENT A-49 PRESSURIIED THERNAL SHOCK / / N/A PMR ONLY
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