ML18038A702
| ML18038A702 | |
| 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 | |
| TASK-***, TASK-OR GL-89-21, NMP2L-1216, NUDOCS 8912130189 | |
| Download: ML18038A702 (90) | |
Text
>-CCFLZ.M'M3 S~('Ri&0'ii Gi~ D~XGi<~".i~.'~'1u5 REGULARLY INFORMATION DISTRIBUTI~SYSTEM (RIDS) r 'I ACCESSION NBR:8912130189 DOC.DATE: 89/11/28 NOTARIZED: NO DOCKET FACIL:50-410 Nine Mile Point Nuclear Station, Unit 2, Niagara Moha 05000410 AUTH. NAME AUTHOR AFFILIATION TERRYiC.D. Niagara Mohawk Power Corp.
RECIP.NAME RECIPIENT AFFILIATION Document Control Branch (Document Control Desk)
SUBJECT:
Responds to Generic Ltr 89-21 re status of implementation of USI requirements.
DISTRIBUTION CODE: A012D TITLE: Generic Ltr COPIES RECEIVED:LTR / ENCL / SIZE:
89-21 Response, Implementation of Unresolved Safety Issue I D
NOTES:
S RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME PD1-1 PD BENEDICT,R LTTR 1
1 1
(
ENC~~I MARTIN,R.
NAME LTTR ENCL 1 1 h
INTERNAL: ACRS 1 1 BARBERiG PTSB 1 1 NUDOCS-ABSTRACT 1 1 REG FILE 01 1 1 WESSMAN,R PD13 1 1 EXTERNAL: LPDR 1 1 NRC PDR 1 1 NSIC 1 1 h
TOTAL NUMBER. OF COPIES REQUIRED: LTTR 11 ENCL 11
I H
el Y NIAGARA O MOHAWK t
NIAGARAMOHAWKPOWER CORPORATION/301 PLAINFIELDROAD, SYRACUSE. N,Y, 13212/TELEPHONE (315) 474-1511 November 28, 1989 NMP2L 1216 I
L U.S. Nuclear Regulatory Commission Attn: Document Control Desk Hashington, D.C. 20555 Re: Nine Mile Point Unit 2 Docket No. 50-410 NPF-69 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 89-2i).
Enclosure 1 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 2.
Very truly yours, NIAGARA MOHAHK PPHER CORPORATION g~/ j~~
C. D. Terry Vice President Nuclear Engineering and Licensing TPS/mjd 8017G Enclosure xc: Regional Administrator, Region I Mr. R. A. Capra, Director Mr. R. E. Martin, Project Manager Mr. H. A. Cook, Resident Inspector Records Management 89121%0189-891128 PDR ADOCI/, 050004i0 P PNU
ENCLOSURE 1 USI/MPA NUMBER TITLE REF. DOCUMENT APPLI CAB I LI TY STATUS/DATE* REMARKS A-1 Water Hammer SECY 84-119 Al 1 NC NMP2 was built/licensed NUREG-0927, Rev. 1 with design features
, NUREG-0993, Rev. 1 consistent with require-NUREG-0737 Item ments of NUREG-0927 to I.A.2.3 minimize the potential of SRP revisions water hammer, as referenced in USAR Section 1.13, system sections (5.4.6,~
5.4.7,. 6.3) of USAR and~
Training per TMI Action I.A.2.3 is incorporated per USAR Section 1.10.
A-2/ Asymmetri c Blowdown NUREG-0609 PWR NA NMP2 is a BWR.
MPA D-10 Loads on Reactor Primary GL 84-04, GDC-4 Coolant Systems A-3 Westinghouse Steam NUREG-0844 H-PHR NA NMP2 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 NMP2 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 8016G
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USI/HPA NUMBER TITLE REF. DOCUMENT APPLICABILITY STATUS/DATE* REMARKS A-5 BIIW Steam Generator NUREG-0844, SECY 86-97 BEW-PWR NHP2 is a BWR.
Tube Integrity SECY 88-272 GL 85-02 (No requirements)
E A-6 Hark I Containment NUREG-0408 Hark I-BWR NMP2 has a Mark II Short-Term Program Containment A-7/ Mark I Long-Term NUREG-0661 Hark I-BWR NA NMP2 has a Mark II D-01 Program NUREG-0661 Suppl. 1 Containment GL 79-57 A-8 Mark II Containment NUREG-0808 Mark II-BWR NC NHP2 meets the requ>rements Pool Dynamic Loads NUREG-0487, Suppl. 1/2 of NUREG 0808, 0487, Suppl.
NUREG-0802 1/2 and 802 per USAR SRP 6.2.1.1C Section 1.13, Appendix 6A GDC 16 and NRC SER, SSER3,
. (NUREG 1047) Sections 3.9.3.1, 6.2.1.7.
Anticipated Transients NUREG-0460, Vol. 4 Al 1 NC Compliance with 10CFR50.62 Without Scram 10 CFR 50.62 is described in USAR Section 15.8. Also refer-enced in NUREG 1047, SSER 2 and 3, and Safety Evaluation dated 3/24/
from NRC.
- C COMPLETE NC NO CHANGES NECESSARY NA- NOT APPLICABLE I INCOMPLETE E FVALUATING ACTIONS REQUIRED 8016G
0 l USI/MPA NUMBER TITLE REF. DOCUMENT APPLICABILITY STATUS/DATE* REMARKS A-10/ BWR Feedwater Nozzle NUREG-0619 BWR NC NMP2 is in compliance with MPA 8-25 Cracking Letter from DG Eisenhut NUREG-0619 as identified dated ll/13/80 in USAR Sections 1.12 GL 81-11 (issues 2 & 36) and 1.13 and as discussed in NUREG 1047 (SER), Section 3.9.3.1.
A-11 Reactor Vessel Material NUREG-0744, Rev. 1 All NC NMP2 is in compliance with Toughness 10 CFR 50.60/ 10CFR50, Appendix G as 82-26 identified in USAR Sectio~
1.13, Section 5.3 and Appendix 5A. This is concurred with in NRC SER (NUREG 1047)
A-12 Fracture Toughness 'of NUREG-0577, Rev. 1 PWR NA NMP2 is a BWR.
Steam Generator and SRP Revision Reactor Coolant Pump 5.3.4 Supports A-17 Systems Interactions Ltr: DeYoung to Al 1 NC Per Guidance of GL 89-21, licensees-9/72 no actions required.
NUREG-1174, NUREG-1229, NUREG/CR-3922, NUREG/CR-4261, NUREG/
CR-4470, GL 89-18 (No requirements)
- C COMPLETE NC NO CHANGES NECESSARY NA NOT APPLICABLE I - INCOMPLETE E EVALUATING ACTIONS REQUIRED 8016G
Cr US I/HPA NUMBER TITLE REF. DOCUMENT APPL ICAB1LITY STATUS/DATE* REMARKS A-24/ f Qual i i cation of Class NUREG-0588, Rev. 1 Al 1 NC NMP2 meets or MPA B-60 1E Safety-Related SRP 3.11 exceeds the requirements Equipment 10 CFR 50.49 of NUREG-0588 as identified GL 82-09, GL 84-24 in USAR Section 1 ~ 13. The GL 85-15 EQ Program is described in the NMPC EQD (Equipment Qualification Document) dated April, 1985, and is in compliance with 10CFR50.49 as identified in letter~
NMP2L 0822 and 0833 date~
8/18/86 and 8/21/86 respectively and as discussed in NRC SSER4 (NUREG-1047).
A-26/ Reactor Vessel Pressure DOP Letters to NMP2 is a BHR 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 NC NMP2 is in compliance with Shutdown Requirements RG 1.113, 01/19 requirements as identified RG 1.139 in USAR Section 1.13, Sec ion SRP 5.4.7 5.4.7 and as discussed i the NRC SER (NUREG-1047).
C COMPLETE NC NO CHANGES NECESSARY NA NOT APPLICABLE I - INCOMPLETE E EVALUATING ACTIONS REQUIRED 8016G
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USI/MPA NUMBER TITLE REF. DOCUMENT APPLICABILITY STATUS/DATE* REMARKS A-36/ Control of Heavy Loads NUREG-0612 Al 1. NC NMP2 is in compliance with C-IO, Near Spent Fuel SRP 9.1.5 the requirements of NUREG C-15 GL 81-07, GL 83-42, 0612 as identified in USAR GL 85-11 Section 1.13 and Appendix Letter from DG 9C and as discussed in the Eisenhut dated NRC SER (NUREG 1047) 12/22/80. Section 9.1.5.
A-39 Determination of SRV NUREG-0802 NC NMP2 is in compliance with Pool Dynamic Loads NUREGs-0763,0783,0802 the requirements of NUREG~
and Pressure Transients .NUREG-0661 0783 as identified in USA~i, SRP 6.2.1.1.C Section 1.13 and Appendix 6A and as discussed in the NRC SER, SSER 3, 4, 5 and 6 (NUREG 1047), Sections 3.9.3.1 and 6.2.1.7.
A-40 Seismic Design SRP Revisions, NUREG/ All NA Seismic Design was Criteria CR-4776, NUREG/CR-0054, developed in accordance NUREG/CR-3480, NUREG/ with Reg. Guide 1.60 and CR-1582, NUREG/CR-1161, 1.61 as described in USAR NUREG-1233, NUREG-4776 Sections 3.7 and 1.8.
NUREG/CR-3805 NUREG/CR-5347 NUREG/CR-3509
- C COMPLETE NC NO CHANGES NECESSARY NA NOT APPLICABLE I - INCOMPLETE E EVALUATING ACTIONS REQUIRED H016G
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US I /MPA NUMBER TITLE REF. DOCUMENT APPLICABILITY STATUS/DATE* REMARKS
'-A-42/ Pipe Cracks in Boiling NUREG-0313, Rev. 1 BWR NMP2 is in compliance with.
MPA 8-05 Water Reactors NUREG-0313, Rev. 2 the requirements of GL 81-81-03, 88-01 GL GL
$ 0$ (NUREG 0313, Rev. 1) as incorporated into USAR Section 5.2.3 and discussed in NRC SER (NUREG 1047).
NMP2 has respon e to GL88-01 (NUREG 0313, Rev. 2) in letters NMP2L 1151 and 1213 dated 7/28/88 and 11/1/89 respectively. As~
indicated, an additional response will be provided
~
by 12/15/89.
A-43 Containment Emergency NUREG-0510, Al I NC NMP2 is in compliance with Sump Performance. NUREG-0869, Rev. 1 requirements as discussed NUREG-0897, R.G. 1.82 in NRC SER, Sec. 6.2.2 (Rev. 0) SRP 6.2.2 (NUREG 1047). Design GL 85-22 reviews are also performed (No requirements) against GL 85-22 as part .
of the 10CFR50.59 safety evaluation process.
A-44 Station Blackout RG 1.155 Al 1 NMP2 response to Station NUREG-1032 Blackout was submitted to NUREG-1109 the NRC in Letter NMP2L-10 CFR 50.63 1184 dated 4/13/89.
Procedure changes are required as identified and are projected to be completed within 12 months following NRC notification per 10CFR50.63(C)(3).
- C COMPLETE NC NO CHANGES NECESSARY NA NOT APPLICABLE I INCOMPLETE E EVALUATING ACTIONS REQUIRED 8016G
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US I /MPA NUMBER TITLE REF. DOCUMENT APPL I CAB I L I TY STATUS/DATE* REMARKS A-45 Shutdown Decay Heat SECY 88-260 A11 Per the guidance of GL89-21 Removal Requirements NUREG-1289 this issue will be evalu-NUREG/CR-5230 ated under the IPE program.
SECY 88-260 This program is expected (No requirements) to be completed by July 31, 1992 (NMP2L 1212, dated 10/31/89).
A-46 A-47 Seismic Qualification of Equipment in Operating Plants Safety Implication NUREG-1030 NUREG-1211/
GL 87-02, GL 87-03 NUREG-1217, NUREG-Al 1 Al 1 NA NMP2 was sei smi c'ibui1 criteria.
NMP2 t to censing is currently evalu-t current of Control Systems 1218 ating the requirements of GL 89-19 GL 89-19 and is expected to provide its response by March 19, 1990.
A-48 Hydrogen Control 10 CFR 50.44 Al 1, except NC NMP2 is in compliance with Measures and -Effects SECY 89-122 PHRs with the requirements of of Hydrogen Burns large dry 10CFR50.44 as identified on Safety Equipment containments in USAR Section 1.13, 6.2.5 and as discussed in the NRC SER (NUREG 1047).
A-49 Pressurized Thermal RGs 1.154, 1.99 NMP2 is a BHR.
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 8016G
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PLANT NMP-2 DOCKET NO(S). 50-410 PROJECT MANAGER Robert E. Martin TECHNICAL CONTACT A. Serkiz US I NO. A-I TITLE Mater Kamer NPA NO. ~NA TAC NOS.
ISSUES
SUMMARY
This Unresolved Safety Issue (USI) was resolved in March 1984, with the publication of NUREG-0927, "Evaluation of Mater 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-1 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:
NMP-2 was built/licensed with design features consistent with requirements of NUREG-0927 to minimize the potential of water hammer, as referenced in USAR Section 1. 13, system sections (5.4.6, 5.4.7, 6.3) of USAR and NRC SER (NUREG 1047). Training per TMI Action I.A.2.3 is incorporated per USAR Section 1.10.
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REFERENCES:
NMP-2 A-1
- 1. REQUIREMENT 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-1, 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, "Resolution 03/15/84 of Unresolved Safety A-l, Water Hammer"
- 2. IMPLEMENTATION DOCUMENTS:
TITLE NUDOCS NO. DATE NUREG-1047 (SER) 02/85 Updated Final Safety Analysis Report Letter C. D. Terry, NMPC, 8912130189 11/28/89 to NRC
- 3. VERIFICATION DOCUMENTS:
TITLE NUDOCS NO. DATE
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PLANT NMP-2 DOCKET NO(S) . 50-410 PROJECT MANAGER Robert E. Martin TECHNICAL CONTACT J. Kudrick OSI NO. A-8 TITLE Mark II Containment Pool 0 namic Loads MPA NO. TAC NOS.
ISSUES
SUMMARY
- 6. USI NO. A-8 TITLE: Mark II Containment Pool D namic Loads This USI was resolved in August 1981 with the publication of NUREG-0808, "Mark II Containment Program Load Evaluation and Acceptance Criteria," and Standard Review Plan (SRP) Section 6.2.1. 1C. The requirement is that the 11 BWRs having the Mark II containment shall meet the requirements of GDC 16.
As stated in NUREG-0808, the original design of the Mark II containment system considered only those loads normally associated with design-basis accidents that were known at the time. These included pressure and temperature loads associated with a LOCA, seismic loads, dead loads, jet impingement loads, hydrostatic loads due to water in the suppression chamber, overload pressure test loads, and construction loads. However, since the establishment of the original design criteria, additional loading conditions were identified that must be considered for the pressure-suppression containment-system design.
In the course of performing large-scale testing of an advanced design pressure-suppression containment (Ma} k III), and during inplant testing of Mark I containments, new suppression-pool hydrodynamic loads were identified that had not been included explicitly in the original Mark II containment-design basis.
These additional loads result from dynamic effects of drywell air and steam being rapidly forced into the suppression pool during a postulated LOCA and from suppression'-pool response to safety/relief valve (SRV) operation; these are generally associated with plant transient operating conditions. Because these new hydrodynamic loads had not been considered, the NRC staff determined that a detailed reevaluation of the Mark II containment system was required.
The issuance of NUREG-0808, NUREG-0802, "Safety Relief Valve Quencher Loads:
Evaluation for BWR Mark II and III Containments," and NUREG-0487, "Mark II Containment Lead Plant Program Load Evaluation and Acceptance Criteria,"
documented acceptable methods for calculating the hydrodymanic loads associated with plant transient conditions. Specifically, the loads referenced in these NRC staff reports, as modified by the acceptance critet ia, constituted the
'esolution of USI A-8. SRP Section 6.2. I has been modified to reflect the applicability of these reports to Mark II containment evaluations.
Implementation is believed to be complete for all Mark II BWRS. As part of the licensing process, the staff required that the applicants utilize the new calculation methodology defined in the reference documents before a full power license was issued.
IMPLEMENTATION AND STATUS
SUMMARY
PLANT SPECIFIC:
NMP-2 meets the requirements of NUREG 0808, 0487, Suppl. 1/2 and 0802 per USAR Section 1.13, Appendix 6A and NRC SER, SSER3, (NUREG 1047) Sections 3.9.3.1 and 6.2.1.7.
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REFERENCES NMP-2 A-8 I. REQUIREMENT DOCUMENTS:
TITLE NUDOCS NO. DATE GDC-16, Containment Design NUREG-0808 "Mark II Containment Program Load Evaluation and Acceptance Criteria" August 1981 Standard Review Plant 6.2.1.1.c, "Pressure Suppression Type BWR Containments" Revision 1-4 NUREG-0487, "Mar k II Containment November 1978 Lead Plant Program Load Evaluation and Acceptance Criteria"
- a. Supplement 1 September 1980
- b. Supplement 2 February 1981 NURG-0802, "Safety Relief Valve October 1982 Quencher Loads: Evaluation for 8WR Mark II and III Containments"
- 2. IMPLEMENTATION DOCUMENTS:
TITLE NUDOCS NO. DATE NUREG-1047 (SER) 2/85 Updated Final Safety Analysis Report Letter C. D. Terry, NMPC, 8912130189 11/28/89 to NRC
- 3. VERIFICATION DOCUMENTS TITLE NUDOCS NO. DATE
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PLANT NMP-2 DOCKET NO(S). 50-410 PROJECT MANAGER Robert E. Martin TECHNICAL CONTACT J. Mauck OSI NO. A-9 TITLE ATWS er 10 CFR 50.62 HPA NO. TAC NOS. 66573 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), diverse scram systems for CE and BSW 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)(5), 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 :
The licensee provided information on compliance with the ATWS rule for NMP-,2 by letters dated May 2, 1986 and April 3, 1987. As stated in Section 15.8 of NUREG-1047 Supplement 2 (NMP-2 SER), the staff approved the modification made to-the plant to comply with the ATWS rule with the exception of the ATWS mitigation systems instrumentation logic.
In Supplement 4, Section 15.8 of NUREG-1047 the staff approved the licensee's proposed plan to increase the SLCS sodium pentaborate solution.
In a safety evaluatio'n transmitted to the licensee by letter dated March 24, 1988, the staff evaluated the instrumentation logic and found it acceptable for NMP-2. This safety evaluation completed the staff review of the design modifi-cations to comply with the ATWS rule.
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REFERENCES:
NMP-2 A-9
- 1. REQUIREMENT 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 Letter J. Zwolinski (NRC) 8612230219 12/11/86 to C. V. Mangan (NMPC)
- 2. IMPLEMENTATION DOCUMENTS:
TITLE NUDOCS NO. DATE NUREG-1047 (SER), and 2/85 Supplements 2 and 4 11/85, 9/86 Letter C. V. Mangan, NMPC, 8605070139 05/02/86 to NRC Letter C. V. Mangan, NMPC, 8704070251 04/03/87 to NRC Letter M. F. Haughey, NRC, 8803300163 03/24/88 to C. V. Mangan, NMPC Letter C. D. Terry, NMPC, 8912130189 11/28/89 to NRC
- 3. VERIFICATION DOCUMENTS:
TITLE NUDOCS NO. DATE
~g PLANT NMP-2 DOCKET NO(S). 50-410 PROJECT MANAGER Robert E. Martin TECHNICAL CONTACT K. Wichman USI NO. A-10 TITLE BWR Feedwater Nozz1e Crackin NPA NO. B-25 TAC NOS. 00499 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 letter stated that NUREG-0619 provided the resolu-tion of the staff's generic technical activity USI A-IO, which resulted from the inservice discovery of cracking in feedwater nozzles and control rod drive return line nozzles. NUREG-0619 descr ibes 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 :
NMP-2 is in compliance with NUREG-0619 as identified in USAR Sections 1.12 (issues 2 and 36) and 1.13. The staff accepted the licensee's response to the issues in NUREG-0619 as discussed in NUREG-1047 (SER), Section 3.9.3. 1.
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REFERENCES:
NMP-2 A-10
- 1. REQUIREMENT DOCUMENTS:
TITLE NUDOCS NO. DATE Letter from D. Eisenhut transmitting NUREG-0619, "BNR Feedwater Nozzle and Control'Rod Drive Return Line Nozzle Cracking,"
resolution of A-10 to licensees 11/13/80 Generic Letter 81-11, "BMR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking (NUREG-0619)" 02/20/81
- 2. IMPLEMENTATION DOCUMENTS:
TITLE NUDOCS NO. DATE NUREG-1047 (SER) 2/85 Letter C. D. Terry, NMPC, 8912130189 11/28/89 to NRC
- 3. VERIFICATION DOCUMENTS:
TITLE NUDOCS NO. DATE
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PLANT NMP-2 DOCKET NO(S). 50-410 PROJECT MANAGER Robert E. Martin TECHNICAL CONTACT B. Elliott USI NO. A-Il TITLE Reactor Vessei Nateriais Too hness NPA NO. TAO NOS.
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 Vessel 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 service 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 it is demonstrated that lower values will provide margins of safety against failure equivalent to those provided by Appendix G of the ASME code. USI A-11 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 :
Nine Mile Point 2 Safety Evaluation Report NUREG-1047 Section 5.3 addresses USI A-ll. The staff concluded in Section 5.3 that the reactor vessel materials are acceptable and fulfill the requirements of GDC I, 4, 14, 30, 31 and 32; the materials testing and monitoring requirements of Appendices B, G and H of 10 CFR Part 50; and the requirements of 10 CFR 50.55a.
I
REFERENCES:
NMP-2 A-11
- 1. REQUIREMENT 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 NUREG-1047 (SER) 2/85 Letter C. D. Terry, NMPC, 8912130189 ,11/28/89 to NRC
- 3. VERIFICATION DOCUMENTS:
TITLE NUDOCS NO. DATE
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PLANT NMP-2 DOCKET NO(S) . 50-410 P
PROJECT MANAGER Robert E. Martin TECHNICAL CONTACT 0. Thatcher USI 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 Vulnerabilities. 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, which addressed the status of USI's, no licensee actions were required. The response to the September 26, 1972 letter to all licensees does not apply to NMP-2 licensed in 1987.
REFERENCES:
NMP-2 A-17 1 ~ REQUIREMENT DOCUMENTS:
TITLE NUDOCS WO. 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 Eva luat i on of Sys tern Interact Events and Sources" i on 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 I
Letter C. D. Terry, NMPC, 8912130189 11/28/89 to NRC
. 3. VERIFICATION DOCUMENTS:
TITLE NUDOC NO. DATE
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PLANT NMP-2 DOCKET NO(S). 50-410 PROJECT MANAGER Robert E. Martin TECHNICAL CONTACT P. Shemanski USI NO. A-24 TITLE ualification of Class IE E ui ment NPA NO. TAC NOS. 42476 ISSUES
SUMMARY
This USI was resolved in July 1981 with the publication of NUREG-0588, Revision I, "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 sugary, 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 3.1l. This was further addressed by the first-round environmental qualification inspections conducted by the NRC.
IMPLEMENTATION AND STATUS
SUMMARY
PLANT SPECIFIC:
In section 3.11 of Supplement No. 4 to the NUREG 1047 (NMP-2 SER), the staff concluded that the licensee's environmental qualification program was acceptable and that the licensee has demonstrated conformance with the require-ments of 10 CFR 50.49. By letters dated August 18 and August 21, 1986 the licensee informed the staff that all equipment within the scope of 10 CFR 50.49 was fully qualified,
Further information provided in the licensee's letter of 11/28/89 indicates that NMP-2 meets or exceeds the requirements of NUREG-0588 as identified in USAR Section 1. 13. The EQ Program is described in the NMPC EQD (Equipment Qualification Document) dated April, 1985, and is in compliance with 10 CFR 50.49 as identified in letters NMP-2L 0822 and 0833 dated 8/18/86 and 8/21/86 respectively and as discussed in NRC SSER4 (NUREG-1047).
I'
REFERENCES:
NMP-2 A-24 I. REQUIREMENT 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 Qualification 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 and 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, 8608200097 08/18/86 to E. G. Adensam, NRC Letter C. V. Hangan, NMPC, 8608250252 08/21/86 to E. G . Adensam, NRC NUREG-1047 (SER) 9/86 Supplement 4 Letter C. D. Terry, NMPC, 8912130189 11/28/89 to NRC
- 3. VERIFICATION DOCUMENTS:
TITLE NUDOCS NO. DATE
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PLANT NMP-2 DOCKET NO(S). 50-410 PROJECT MANAGER Robert E. Martin TECHNICAL CONTACT R. Jones USI NO. A-31 TITLE RHR Shutdown Re uirements MPA NO. TAC NOS.
ISSUES
SUMMARY
This USI was resolved in May 1978 with the publication of Standard Review Plan (SRP) Section 5.4.7. Only those plants expected to receive an operating license after January 1, 1979 were affected by this resolution. The USI involved establishment of criter ia for the design and operation of systems necessary to take a power reactor from normal operating conditions to cold shutdown.
SRP Section 5.4.7 stated that, for purposes of implementation, plants would be divided into three classes: Class 1 would require full compliance for Construction Permit (CP) or Preliminary Design Approval (PDA) applications which were docketed on or after January 1, 1978. Class 2 required a partial implementation for all plants for which CP or PDA applications were docketed before January 1, 1978, and for which an Operating License (OL) issuance was expected on or after January 1, 1979. Class 3 affected all operating reactors and all other plants for which issuance of the OL was expected before January 1, 1979. The extent to which Class 3 plants would require implementation was based on the combined staff review of related plant features. In general, the outcome of these evaluations were that only plants receiving an OL after January 1, 1979 were affected by this USI resolution, and there were no backfits to operating plants that had received an operating license before January 1, 1979.
IMPLEMENTATION AND STATUS
SUMMARY
PLANT SPECIFIC :
NMP-2 is in compliance with requirements as identified in USAR Section 1. 13, Section 5.4.7 and as discussed in the NRC SER (NUREG-1047).
4
REFERENCES:
NMP-2 A-31
- 1. REQUIREMENT DOCUMENTS:
TITLE NUDOCS NO. DATE NUREG-0800 "Standard Review Plan," 5/78 SRP Section 5.4.7 NUREG-0606 "Unresolved Safety Issues Summary" Regulatory Guide 1.139, "Guidance for Residual Heat Removal" Regulatory Guide 1.113
- 2. IMPLEMENTATION DOCUMENTS:
TITLE NUDOCS NO. DATE NUREG-1047 (SER) 2/85 Letter C. D. Terry, NMPC, 8912130189 11/28/89 to NRC
- 3. VERIFICATION DOCUMENTS:
TITLE NUDOCS NO. DATE
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PLANT NMP-2 DOCKET NO(S). 50-410 PROJECT MANAGER Robert E. Martin TECHNICAL CONTACT J. Wermiel USI NO. A-36 TITLE Control of Neav Loads Phases I 6 II M A 00. C-10 -1 TAC 00 . 00 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 shutdow'n and continue decay heat removal. USI A-36 was established to systematically examine staff licensing criteria and the adequacy of measures in effect at oper ating 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 f'r 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 submittals in two parts: Phase I (six month response) and Phase II (nine month response). Phase I responses were to address Section 5. 1. 1 of NUREG-0612 which covered the following areas:
0
- 1. Definition of safe load paths
- 2. Development of load handling procedures
- 3. Periodic inspection and testing of cranes
- 4. gualifications, 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 r'equirement 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.
lo 1 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:
As stated in Section 9.1.5 of the SER the licensee responded to the NRC December 22, 1980 generic letter. The staff reviewed the response and concluded that the licensee's program for resolving Phase I concerns of NUREG-0612 was acceptable and that no further action was required concerning Phase II of NUREG-0612.
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REFERENCES:
HMP-2 A-36
- 1. REQUIREMENT DOCUMENTS:
TITLE NUDOCS NO. DATE Letter, Darrell G. Eisenhut, NRC, 8103190732 12/22/80 to all licensees, applicants for OLs and holders of CPs transmitting NUREG-0612 and staff positions Generic Letter 85-11, Hugh L. 8506270216 06/28/85 Thompson, NRC, to all licensees for Operating Reactors, "Completion of Phase II of 'Control of Heavy Loads at Nuclear Power Plants'UREG-0612" 2.'MPLEMENTATION DOCUMENTS:
TITLE HUDOCS NO. DATE Letter C. 'V. Mangan, NMPC, 8407130180 07/10/84 to A. Schwencer, HRC Letter C. V. Mangan, NMPC, 8412050352 11/30/84 to A. Schwencer, HRC NUREG 1047 (SER) 02/85 Letter C. D. Terry, NMPC, 8912130189 11/28/89 to HRC
- 3. VERIFICATION DOCUMENTS:
TITLE HUDOCS NO. DATE
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PLANT NMP-2 DOCKET NO(S). 50-410 PROJECT MANAGER Robert E. Martin TECHNICAL CONTACT J. Kudrick US I NO. A-39 TITLE Determination of SRV Pool Dynamic Loads and Tem erature Limits flPA 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.'n addition, NUREGs 0763, 0783 and 0802 were issued for Mark I, Mark II, and Hark III containments, respectively.
BHR plants are equipped with safety/relief valves (SRVs) to protect the reactor from overpressurization. Plant operational transients, such as turbine trips, wi 11 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. llith 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 criteria, 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):
The staff's evaluation of NMP-2 compliance with NUREG-0661, NUREG-0763, NUREG-0783 and NUREG-0802 is provided in Sections 3.9.3.1 and 6.2 of NUREG-1047 (NMP-2 SER) and in Section 6.2 of Supplements 1 and 3 of NUREG-1047.
In the SER, Confirmatory Issue 6 required the applicant to reevaluate the capability of the containment internal structures to resist the newly identified safety/relief valve (SRV) loads and to submit the results of the reevaluation to the staff for review and evaluation. The applicant has completed the reevaluation, and the results are contained in tables attached to the applicant's letter dated Har ch 18, 1986 (C. V. Hangan to E. G. Adensam).
In addition, the licensee states in its 11/28/89 letter that NMP-2 is ih compliance with the requirements of NUREG 0783 as identified in USAR Section 1. 13 and Appendix 6A and as discussed in the NRC SER, SSER 3, 4, 5 and 6 (NUREG 1047), Sections 3.9.3.1 and 6.2.1.7.
REFERENCES HMP-2 A-39
- 1. REQUIREMENT 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 BLAIR 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. V. Mangan, NMPC, 8603200278 03/18/86 to E. G. Adensam, NRC NUREG-1047 (SER), and 2/85, 6/85 Supplements 1 and 3 and 7/86 Letter C. D. Terry, NMPC, 8912130189 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.
~ .J PLANT NMP-2 DOCKET NO(S). 50-410 PROJECT MANAGER Robert E. Martin TECHNICAL CONTACT W. Koo USI NO. A-42 TITLE Pi e Cracks in Boilin Water Reactors MPA NO. 8-06 TAC NOS. 69148 ISSUES
SUMMARY
This USI was resolved in February 1981 with the publication of NUREG-0313, Revision I, "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 primary 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 fabr ication 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 comunications (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, Revision 2 which was transmitted to all holders of BWR operating licenses via Generic Letter 88-01.
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sa IMPLEMENTATION AND STATUS
SUMMARY
(PLANT SPECIFIC):
The licensee states that NMP-2 is in compliance with the requirements of GL 81-04 (NUREG - 0313, Rev. 1) as incorporated into USAR Section 5.2.3 and discussed in NRC SER (HUREG 1047 Section 5.2.3). For BWRs in licensing review at the time GL 81-04 was issued, such as NMP-2, the implementation document is the SER that addressed conformance to NUREG-0313, Rev. 1, as discussed above.
In addition to the scope of USI A-42 NMP-2 has responded to GL 88-01 (NUREG 0313, Rev. 2) by letters dated July 28, 1988, November 1, November 28, and December 14, 1989.
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REFERENCES:
NMP-2 A-42
- 1. REQUIREMENT DOCUMENTS:
TITLE NUDOCS NO. DATE NUREG-0313, Revision 1, "Technical Report on tlaterial Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping," 07/80 Generic Letter 81-03, "Implemen-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 NUREG-1047(SER) 2/85 Letter C. D. Terry, NMPC, 8808050150 07/28/88 to NRC Letter C. D. Terry, NMPC, 8911150118 11/01/89 to NRC Letter C. D. Terry, NMPC, 8912130189 11/28/89 to NRC Letter C. D. Terry, NMPC, to NRC 8912200079 12/14/89
- 3. VERI FI CATION DOCUMENTS:
TITLE, NUDOCS NO. DATE
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PLANT NMP-2 DOCKET NO(S) . 50>>410 PROJECT MANAGER Robert E. Martin TECHNICAL CONTACT A. Serkiz USI NO. A-43 TITLE Containment Emer enc Som Performance NPA NO. TAC NOS.
ISSUES
SUMMARY
- 19. USI NO. A-43 TITLE: Containment Emer enc Sum Performance The resolution of this USI was presented to the Coomission in October 1985 in SECY-85-349. NUREG-0897, Revision 1, "Containment Emergency Sump Performance,"
presents the results of the staff's technical findings. These findings estab-lished a need to revise current licensing guidance on these matters. RG 1.82 Revision 0 and Standard Review Plan Section 6.2.2, "Containment Heat Removal Systems" were revised to reflect this new guidance. No licensee actions were required.
Initially, an issue existed concerning the availability of adequate recircula-tion cooling water following a loss-of-coolant accident (LOCA) when long-term recirculation of cooling water from the PWR containment sump, or the BWR resjdual heat removal system .(RHR) suction intake, must be initiated and maintained to prevent core melt.
The technical concerns evaluated under USI A-43 were: (a) post-LOCA adverse conditions resulting from potential vortex formation and air ingestion and subsequent pump failure, (b) blockage of sump screens with LOCA generated insulation debris causing inadequate net positive suction head (NPSH) on pumps, and (c) RHR and containment spray pumps inoperability due to possible air, debris, or particulate ingestion on pump seal and bearing systems.
This revised guidance applies only to future construction permits, preliminary design approvals, final design approvals, standardized designs, and applica-tions for licenses to manufacture. The staff performed a regulatory analysis to determine if this new guidance should be applied to operating plants. The results of this analysis were reported in NUREG-0869 Revision 1, "USI A-43 Regulatory Analysis," issued in October 1985. The staff concluded that the regulatory analysis does not support any new generic requirements for present licensees to perform debris assessments.
IMPLEMENTATION AND STATUS
SUMMARY
PLANT SPECIFIC :
NMP-2 is in compliance with requirements as discussed in NRC SER, Section 6.2.2 (NUREG-1047). Licensee's design reviews are also performed against GL 85-22 as part of the 10 CFR 50.59 safety evaluation process.
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REFERENCES:
NMP-2 A-43
- 1. REQUIREMENT DOCUMENTS TITLE NUDOCS HO. DATE NUREG-0869, Rev. 1, "USI 10/85 A-43 Regulatory Analysis" NUREG-0897, Rev. 1, "Containment 10/85 Emergency Sump Performance" GL 85-22, "Potential for Loss 12/03/85 of Post-LOCA Recirculation Capability Due to Insulation Debris Blockage"
- 2. IMPLEMENTATION DOCUMENTS: .
TITLE NUDOCS HO. DATE NUREG-1047 (SER) 02/85 Letter C. D. Terry, NMPC, 8912130189 11/28/89 to NRC
- 3. VERIFICATION DOCUMENTS:
TITLE NUDOCS NO. DATE
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PLANT NMP-2 DOCKET NO(S). 50-410 PROJECT MANAGER Robert E. Martin TECHNICAL CONTACT P. Gi 11 US I NO. A-44 TITLE Station Blackout MPA NO. TAC NOS. 68571 ISSUES
SUMMARY
'his 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 concurr ent with turbine trip and the unavailability of the redundant onsite emergency ac power systems. WASH-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 ru1e was published for comment on March 21, 1986. A final rule, 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 ac power source, if available, to cope with a station blackout 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 responded to the requirements of 10 CFR 50.63 as discussed above. Operator action is required to shed non-essential loads from non-class IE batteries to cope with a station blackout for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Procedure changes are required as identified and are projected to be completed within 12 months following NRC notification per 10 CFR 50.63(c)(3). The licensee's response is under review. The implementation date on A-44 is expected to be not later than 12/31/91, based on the estimate for staff completion of the plant-specific SER by 12/20/90 and the ]icensee's stated lead time to complete procedural modifications after rece>pt of the SER.
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REFERENCES:
NMP-2 A-44 I. RE UIREMENT DOCUMENTS:
TITLE NUDOCS NO. DATE 10 CFR 50.63, "Loss of All Alternating Current Power" 06/21/88 Regulatory Guide 1. 155, "Station Blackout" 08/88
- 2. IMPLEMENTATION DOCUMENTS:
TITLE NUDOCS NO. DATE Letter C. D. Terry, NMPC, 8904240050 04/13/89 to NRC Letter C. D. Terry, NMPC, 8912130189 11/28/89 to NRC
- 3. VERIFICATION DOCUMENTS:
TITLE NUDOCS NO. DATE
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PLANT NMP-2 DOCKET NO(S). 50-410 PROJECT MANAGER Robert E. Martin TECHNICAL CONTACT R. Jones USI ND. A-45 TITLE Shutdown Deca Heat Removal Re uirements HPA ND. TAG NDS.
ISSUES
SUMMARY
USI A-45 was resolved by SECY 88-260, "Shutdown Decay Heat Removal Requirements (USI-A-45)," issued September 13, 1988, without imposing any new licensing requirements other than the Individual Plant Examination (IPE), as described below. At the same time the staff issued NUREG-1289, "Regulatory and Backfit Analysis: USI A-45." Since all of the significant USI A-45 results have been found to be highly plant specific, the Commission decided it was not appropriate to propose a single generic corrective action to be applied uniformly to all plants.
The Commission is currently implementing the Severe Accident Policy (50 FR 32138) and will require all plants presently operating or under construction to undergo a systematic examination termed the IPE. The reason for this examina-tion is to identify any plant-specific vulnerabilities to severe accidents.
The IPE analysis intends to examine and understand the plant emergency pro-cedures, design, operations, maintenance, and surveillance, in order to identify vulnerabilities. The analysis will examine both the decay heat removal systems and those systems used for other related functions. This includes CE plants without power-operated relief valves.
NRC has decided to subsume A-45 into the IPE program as the most effective way of achieving resolution of specific plant concerns associated with A-45.
IMPLEMENTATION AND STATUS
SUMMARY
PLANT SPECIFIC :
Per the guidance of GL 89-21 this issue will be evaluated under the IPE program. This program is expected to be completed by July 31, 1992 (NMP-2L 1212, dated October 31, l989).
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REFERENCES NMP-2 A-45
- 1. REQUIREMENT DOCUMENTS TITLE NUDOCS NO. DATE Federal Register Notice "10 CFR Part 50, Shutdown Decay Heat Removal Requirements" NUREG/CR-5230 "Shutdown Decay Heat April 1989 Removal Analysis: Plant Case Studies and Special Issues Summary Report" NUREG-1289 "Regulatory and Backfit 11/30/88 Analysis for the Resolution of USI A-45" SECY-88-260 "Shutdown Decay Heat 09/13/88 Removal Requirements
- 2. IMPLEMENTATION DOCUMENTS:
TITLE NUDOCS NO. DATE Letter C. D. Terry, NMPC, , 8911070304 10/31/89 to NRC Letter C. D. Terry, NMPC, 8912130189 11/28/89 to NRC
- 3. VERIFICATION DOCUMENTS:
TITLE NUDOCS NO. DATE
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PLANT NMP-2 DOCKET NO(S). SO-410 PROJECT MANAGER Robert E. Martin TECHNICAL CONTACT J. Mauck USI NO. A-47 TITLE Safety Implication of Control Systems in LWR Nuclear Power Plants MPA NO. TAC NOS.
ISSUES
SUMMARY
USI A-47 was resolved September 20, 1989, with the publication of Generic Letter (GL) 88-19.
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 assure 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 oper ation, and protection system surveillance. The Technical Specifications recommenda-tions are consistent with the criteria and the risk considera-tions of the Coomission 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 BSW 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-2 is currently evaluating the requirements of GL 89-19 and is expected to provide its response by March 19, 1990.
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REFERENCES:
NMP-2 A-47
- 1. REQUIREMENT 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 Imp)ications 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, 8912130189 11/28/89 to NRC
- 3. VERIFICATION DOCUMENTS:
TITLE NUDOCS NO. DATE
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PLANT NMP-2 DOCKET NO(S). 50-410 PROJECT MANAGER Robert E. Martin TECHNICAL CONTACT J. Kudrick USI HO. A-48 TITLE Hydrogen Contro1 Heasures and Effects of Hydrogen Burns on Safety E ui ment HPA NO. TAC NOS.
ISSUES
SUMMARY
USI A-48 was resolved on fully implemented April 19, 1989, as stated in SECY 89-122. It is considered at BWR Nark I and Nark II facilities, as these facilities use inerting as a method of hydrogen control.
USI A-48 was initiated as 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 Hark 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 Hark I and II not only resolved but understood to be fully implemented in the affected containments is plants.
The rule for BWRs with Hark III containments and PWRs with ice condenser containments was published on January 25, 1985. ,he ru'le 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 Mat k 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 Hark 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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REFERENCES:
NMP-2 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):
NMP-2 is in compliance with the requirements of 10 CFR 50.44 as identified in USAR Section 1.13, 6.2.5 ard as discussed in the NRC SER (NUREG 1047).
REFERENCES'.
REQUIREMENT DOCUMENTS:
TITLE WUDOCS NO. DATE 10 CFR 50.44, Standards for 12/81 Combustible Gas System in Light-'I<ater-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 HUDOCS NO. DATE HUREG-1047 (SER) 2/85 Letter C. D. Terry, NMPC, 8912130189 11/28/89 to NRC
- 3. VER IF I CATION DOCUMENTS:
TITLE NUDOCS NO. DATE
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h/C/ ssuRE 3 Page Ho. 2 02/06/90 LISTIH6 OF INCOtlPLETE USI DATA FOR IHPUT FRON PROJECT NANAGERS ISSUE ISSUE DESCRIPTIVE MANE INPLENENT INPLENENT LICENSEE COHNENT STAFF CONHENT NUtlBER DATE STATUS 55 PLA NT HAt!E: NINE tllLE POIHT 2 A-01 MATER HANNER A-02 ASYNHETRIC BLOMDOWH LOADS OH PMR ONLY REACTOR PRltlARY COOLANT SYSTEtlS A-03 WESTINGHOUSE STEAN GENERATOR TUBE / / H/A WESTINGHOUSE OHLY INTEGRITY A-04 CE STEAN 6EHERATOR TUBE IHTEGRITY I / N/A CE PLANTS ONLY
. A-05 BfcM STEAN GENERATOR TUBE / / H/A BC(M PLANTS ONLY INTEGRITY A-06 NARK I SHORT-7ERN PROGRAN / I N/A tlK I BMR OHLY A-07 NARK I LONG-TERN PROGRAM / / N/A NK I BMR ONLY A-08 HARK ll CONTAIHtlENT POOL DYHANIC I / HC LOADS - LOHG-TERN PROGRAN A-09 AIMS I I HC A-10 BMR FEEDMATER NOZZLE CRACKING I I XC A-11 REACTOR VESSEL NATERIALS / / NC A-12 TOU6HMESS FRACTURE 70U6HNESS OF STEAN I / H/A Pk8 GF-ARBOR~
~~y LY GENERATOR AHD REACTOR COOLANT PUHP SUPPORTS A-17 SYSTENS INTERACTION I / HC HO REQUIREMENTS A-24 QUALIFICATION OF CLASS 1E / / HC SAFE7Y-RELATED EQUIPtlEHT A-26 REACTOR VESSEL PRESSURE TRANSIENT / / H/A PMR ONLY PROTECI'IOH A-31 RHR SHUTDOWN REQUIRENEHTS / / t(C LICENSIHG SER A-36 CONTROL OF HEAVY LOADS I / HC 6L-85-11 ENDED
~Mr A-39 DETERNIHA710N OF SAFETY RELIEF / / NC VALVE POOL DYNAHIC LOADS AND TEHPERATURE LINITS A-40 SEISNIC DESIGN CRITERIA- / / NC SHORT-TERN PR06RAN A-42 PIPE CRACKS IN BOILIN6 MATER I / HC GL-88-01 N/A REACTORS A-43 COHTAIHNEHT ENER6EHCY SUtlP / I NC INFO ONLY PERFORMANCE A-44 STATION BLACKOUT 12/31/91 I PROCEDURES SER 12/31/90 A-45 SHUTDOMH DECAY HEAT RENOVAL / I NC IPE SUBSUHED ~ERE-AGC / lY/ B / tc H REQUIRENEHTS A-46 SEI SNI C QUAL IF ICATION OF / / N/A Wry /=p/4'. f'Ri.lpga EQUIPNENT IH OPERATING PLANTS e 6 A-47 SAFETY INPL I CAT IONS OF CONTROL 03/19/90 E NEW REQUIRENENTS PL///~5'HERTED SYSTENS A-48 HYDROGEN COHTROL tlEASURES AND / / NC EFFECTS OF HYDR06EN BURHS OH SAFETY EQUIPHEHT A-49 PRESSURIZED 7HERNAL SHOCK / / . H/A PMR OHLY
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