ML17347B437
| ML17347B437 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 11/27/1989 |
| From: | Harris K FLORIDA POWER & LIGHT CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| REF-GTECI-MI, REF-GTECI-SC, TASK-***, TASK-OR GL-89-21, L-89-415, NUDOCS 8912050117 | |
| Download: ML17347B437 (32) | |
Text
P.O. Box14000, Juno Beach, FL 33408-0420 NOV R7 1989 L-89-415 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D. C. 20555 Gentlemen:
RE: Turkey Point Units 3 & 4 Docket Nos. 50-250 and 50-251 Res onse to Generic Letter 89-21 NRC Generic Letter 89-21, "Request for Information Concerning Status of Implementation of Unresolved Safety Issue (USI)
Requirements", issued October 19, 1989, requested the status of implementation of USIs for which a final technical resolution has been achieved and which are applicable to our facility. A table of specific USIs was included, with a guide for updating the status of those USIs applicable to Florida Power & Light (FPL). A response was requested within 30 days of receipt of the generic letter, which was received by FPL on October 27, 1989.
FPL has completed a document search of our files for each applicable USI and the results are provided in the attached which was provided with Generic Letter 89-21. If any additional information is required on this matter, please contact us ~
Yours very truly, K. Harris Vic resident Turk y Point Plant Nuclear Attachments KNH/TCG/rh cc: Stewart D. Ebneter, Regional Administrator, Region II, USNRC Senior Resident Inspector, USNRC, Turkey Point Plant 8912050ii7 85'ii27 PDR ADOCK 05000250 p PDC an FPL Group company
ENCLOSURE 1 UNRESOLVED SAFETY ISSUES FOR WHICH A FINAL TECHNICAL RESOLUTION HAS BEEN ACHIEVED USI/MPA NUMBER TITLE REF. DOCUMENT APPLICABILITY STATUS/DATE* REMARKS A-1 Mater Hammer SECY 84-ll9 All NC NUREG-0927, Rev. 1 NUREG-0993, Rev. 1 NUREG-0737 Item I.A.2.3 SRP revisions A-2/ =
Asymmetri c Rl owdown NUREG-0609 PWR c (~>/3s} See Nuke 1 (AktadIed MPA D-10 Loads on Reactor Primary GL 84-04, GDC-4 Coolant Systems A-3 Westinghouse Steam NUREG-0844 M-PMR C (10/35) See Nofe 2 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-PMR NA Integrity SECY 88-272 GL 85-02
{No requirements)
RSM Steam Generator NUREG-0844) SECY 86-9? MW-PMR Tube Integrity SECY 88-272 GL 85-02
{No Requirements)
E A-6 Mark I Containment NUREG-0408 Mark I-BWR NA Short-Term Program
~
- C - COMPLETE NC - NO CHANGES NECESSARY NA - NOT APPLICABLE I - INCOMPLETE E - FVALUATING ACTIONS REQUIRED
US I/NPA MUMBER TITLE REF. DOCUNENT APPL ICAB IL ITY STATUS/DATE* REMARKS A-7/ . Nark I Long-Term NUREG-0661 Nark I-BMR MA D-Ol Program NUREG-0661 Suppl. 1 GL 79-57 A-8 Nark II Containment NURECi-0808 Hark I I-BMR MA Poo1 Dynamfc Loads NURE6-0487, Suppl. 1/2 NUREG-0802 SRP 6.2.l.lC GDC 16 A-9 Anticipated Transients NUREG-0460, Vol. 4 Al 1 l (6/91) Ho/e 3 Hfthout Scram 10 CFR 50.62 A-.10/ BMR Feedwater Nozz1e NUREG-0619 MA NPA 8-25 Cracking Letter from DG Efsenhut dated 11/13/80 GL 81-11 A-11 Reactor Vessel Naterfal NUREG-0744, Rev. 1 A11 I (Aeu'~ng Tmk Mole 4 Touohness 10 CFR 50.60/ Rmuhts) 82-26 A-12 Fracture Toughness of NURE6-0577, Rev. I MC Steam Generator and SRP Revision Reactor Coolant Puitp 5.3.4 Supports A-17-- Systems Interactfons Ltr: DeYoung to A41 C (9/89) Mote 5 1icensees - 9/72 NUREI -l174, NUREG-1229, NUREG/CR-3922, NUREG/CR-4261, NUREG/
CR-4470, GL 89-18 (No requirements)
A-24/ Oua1iffcatfon of C1ass NUREG-0588, Rev. 1 Al 1 C (1/89) Moke 6 NPA B-60 1E Safety-Related SRP 3.11 Equipment 10 CFR 50.49 GL 8 -09, GL 84-24 GL 85-15
<<0&oben 1989 In earcaCed ScheE(use
US I/MPA NUMBER TITLE REF. DOCUMENT APPLICABILITY STATUS/DATE+ RFMARKS A-26/ Reactor Vessel Pressure DOR Letters to PWR C (j2/82) Moke, 7 ~
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 Shutdown Requirements RG 1.113, 01/79.
RG 1.139 SRP 5.4.7 A-36/ Control of Heavy Loads NUREG-0612 All I (1/9t) 4 bloke 3 C-IO, Near Spent Fuel SRP 9.1.5 C-15 GL 81-07, GL 83-42, GL 85-11 Letter from DG Eisenhut dated 12/22/80 A-39 Determination of SRV NUREG-0802 BWR NA Pool Dynamic Loads NUREGs-0763,0783,0802 and Pressure Transients NUREG-0661 SRP 6.2. 1. 1. C A-40 Seismic Design SRP Revisions, NUREG/ All NC Criteria CR-4776, NUREG/CR-0054, 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 BWR NA MPA B-05 Water Reactors NUREG-0313, Rev. 2 GL 81-03, GL 88-01
- Oekobu. 1939 Ti<eg~ed S eh.educe
USI/MPA MUMBER TITLE REF. DOCUMENT APPLICABILITY STATUS/DATE* REMARKS A-43 Con~ainment Emergency NUREG-0510, 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 A11 1 (6/9l) No~e 9 NUREG-1032 NUREG-1109 10 CFR 50.63 A-45 Shutdown Decay Heat SECY 88-260 All I (6/9l) No~e 10 Removal Requirements NUREG-1289 ~
NUREG/CR-5230 SECY 88-260 (No requirements)
I (Av~~g Note l l A-46 Seismic gual if i cation NUREG-1030 Al 1 Appaovak)
NRC of Equipment in NUREG-1211/
Operating Plants GL 8?-02, GL 87-03 A-47 Safety Implication NUREG-1217, NUREG- Al 1 E (3/20/90) o+ Control Svstems 1218 GL 89-19 A-48 Hydrogen Control 10 CFR 50.44 All, except Measures and Effects SECY 89-122 PWRs with o< Hydrogen Burns large dry on Safety Equipment containments Pressurized Thermal (3/87) No~e, 12 A-49 RGs 1.154, 1.99 PWR C Shock SECY 82-465 SECY 83-288 SECY 81-687 10 CFR 50.61/
~ '
t '
1
~
l NOTES Ref. NRC Letter from Gordon E. Edison to W. F. Conway, dated November 28, 1988.
2 ~ Ref. NUREG 0933, Rev. 1 and FPL Letter L-85-398 to the NRC dated October 16, 1985.
- 3. Ref. NRC Letter from Gordon E. Edison to W. F. Conway, dated May 19, 1988. (Installation of ATWS mitigating system actuation circuitry will be performed per the Integrated Schedule).
4 ~ Ref. FPL Letter L-89-190 to the NRC dated June 16, 1989.
(Incorporates recent industry developments including new analytical techniques, the progress made by the Working Groups and the additional material data to be provided by the Babcock 6 Wilcox Owner's Group into the characterization of the Turkey Point vessel fracture toughness. Schedule dependent on receipt of test results from B & W Owner's Group).
- 5. Ref. Generic Letter 89-18, issued September 6, 1989.
(Remaining issues to be addressed under other programs, e.g.,
GL 88-20 on Individual Plant Examinations).
- 6. Ref. NRC letter from Alan R. Herdt to W. F. Conway, dated January 6, 1989, which accepted the FPL "EQ" program.
7 ~ Ref. NRC letter from Daniel G. McDonald to Dr. Robert E.
Uhrig, dated December 23, 1982.
- 8. Ref. NRC letter from Daniel G. McDonald to Dr. Robert E. Uhrig dated November 1, 1983. Also, FPL letter L-83-146 to the NRC dated March 15, 1983. (Installation of a separate load cell on the reactor head Integrated Schedule).
lift rig will be performed per the
(FPL has scheduled implementation of all Station blackout modifications including development of necessary procedures to coincide with the EPS Enhancement Project per the Integrated Schedule).
- 10. Ref. FPL Letter 89-389 to the NRC dated October 31, 1989 (Individual Plant Examination Required by GL 88-20 will resolve issue).
Ref. FPL Letter 89-352 to the NRC dated October 2, 1989.
(Implementation program still under development).
- 12. Ref. NRC Letter from Daniel G. McDonald to C. O. Woody dated March ll, 1987.
Enc1osure 2 COMPLETED DATA SHEETS FOR IMPLEMENTATION OF UNRESOLVED SAFETY ISSUES TURKEY POINT
PLANT Turke Point 3 DOCKET NO(S). 50-250 PROJECT MANAGER Gordon Edison N<<NNT T USI NO. A-2 TITLE As rrmetric Blowdown Loads in RCS MPA NO. D-IO TAC NOS. None ISSUES
SUMMARY
This USI was resolved in January 1981 with the publication of NUREG-0609, "Asyrrmetric Blowdown Loads on PWR Primary Systems."
In October 1975, the NRC notified each operating PMR licensee of a potential safety problem concerning the fact that asymmetric LOCA loads had not been considered in the design of any PMR piping system. In June 1976 the NRC informed each PWR licensee that it was required to reassess the reactor vessel support design of its facility. The staff expanded the scope of the problem in January 1978 with a request for additional information to all PWR licensees.
NUREG-0609 provided guidance for these analyses. For operating PMRs, Multi-Plant Action (MPA) Item D-10 was established by NRC's Division of Licensing for implementation purposes.
During the course of the work on USI A-2, it was demonstrated that there were only a very limited number of break locations which could give rise to signifi-cant loads. Subsequently, after substantial new technical work, it was demon-strated that pipes would leak before break and that new fracture mechanics techniques for the analyzing of piping failures assured adequate protection against failures in primary system piping in PWRs (Generic Letter 84-04). This was reflected in a revision of General Design Criteria (GDC)-4 (Appendix A to 10 CFR Part 50) published in the Federal Re ister in final form on April 11, 1986, and in a subsequent revision 7o KC- pu >shed in the Federal Re ister on July 23, 1986. In addition, it has also been satisfactoriTyyemons ra e in the course of the A-2 effort that there is a very low likelihood of simultaneous pipe loading with both LOCA and safety shutdown earthquake (SSE) loads.
Therefore, the last revision of GDC-4 represented the final technical action of NRC regarding the issue of asymmetric blowdown loads issue in PMRs primary coolant main loop piping.
IMPLEMENTATION AND STATUS
SUMMARY
(PLANT SPECIFIC):
In a letter dated 11-1-88 FPL documented the applicability of "leak-before-break" to Turkey Point. In GL 84-04 NRC provided an SER for Westinghouse plants, provided certain conditions were met, one of which is applicable to Turkey Point. That condition requires an adequate leak detection system. The leakage detection requirement is implemented as Tech Spec 3.1.3. The NRC reviewed the licensee's response to GL 88-05 (Boric Acid Corrosion), which required procedures for locating small leaks. The NRC also reviewed the capabi lity following the conoseal leak. Since Turkey Point is bounded by the Westinghouse analysis and has adequate leak detection, dynamic effects of RCS pipe break need not be designed for. Based on an evaluation received from NRR's materials branch dated 11-10-88, Memo from C. Y. Cheng to Gordon E.
Edison, a letter dated 11-12-88 was sent to the licensee formally closing MPA D-10.
REFERENCES:
P lant Name A-2
- 1. RE UIREMENT DOCUMENTS:
TITLE NUDOCS NO. DATE Generic Letter "Evaluation of Primary Systems for Asymmetric LOCA Loads" 01/20/78 Task Action Plan A-2, "Asymmetric Blowdown Loads on Reactor Primary Coolant System," NUREG-0371 Task Action Plans for Generic Activities 11/78 "Asyrmetric Blowdown Loads on PWR Primary Systems," NUREG-0609 US NRC NRR 01/81 I
GDC-4, "Environmental and Dynamic Effects Design Basis" GL 84-04, "Safety Evaluation of Westinghouse Topical Reports Dealing With Elimination of Postulated Pipe Breaks in PWR Primary Main Loops."
- 2. IMPLEMENTATION DOCUMENTS:
TITLE NUDOCS NO. DATE Letter from Conway (FPL) to 8811080057 11-1-88 U.S. NRC Letter from Edison (NRC) to 8901090146 11-28-88 Conway (FPL)
Letter from Edison (NRC) to 8903210051 3-10-89 Conway (FPL)
- 3. VERIF ICATION DOCUMENTS:
TITLE NUDOCS NO. DATE Insp. Report 8 50-251/87-16 8706169909 5-15-87
PLANT Turke Point 3 DOCKET NO(S). 50-250 PROJECT tkANAGER Gordon Edison TECHNICAL CONTACT J. Mauck USI NO. A-9 TITLE ATMS er 10 CFR 50.62 NPA NO. A-020 TAG NOS. N59153 ISSUES SS1tlARY:
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)(l),
diverse and independent auxiliary feedwater initiation and turbine trip for all PWRs; 50.62(C)(2), diverse scram systems for CE and B8W reactors; 50.62(C)(3) alternate rod in~ection (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 SUGARY (PLANT SPECIFIC):
In a letter to NRC dated 7-15-87 and supplemented 11-19-87, FPL provided a detai led response to the rule. This response relied on WOG report WCAP-10858. The NRC staff had reviewed and approved WCAP-10858-P-A on 7-7-86. In a letter dated 5-. 19-88, Edison (NRC) to Conway (FPL), NRC approved Rev. 1 of WCAP-10858-P-A, and also approved the FPL design subject to: ( 1) certain human factor reviews, (2) satisfactory completion of isolation device qualification testing, and (3) possible Tech Specs (NRC staff still considering the need for TS). The design approval was based on an SER from Newberry (NRR, ICSB) to Berkow (NRR) dated 3-30-88. Installation of the NSAC equipment is scheduled to be completed during the dual unit outage (which is scheduled to begin early in 1991) for both Turkey Point Units and is being tracked in the FPL Integrated Schedule, which is a part of the license.
REFERENCES:
Plant Name 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. IMPLEMIENTATION DOCUMENTS:
TITLE NUDOCS NO. DATE Letter, FPL to NRC 8702330020 7-15-87 Letter, FPL to NRC 8711240199 11<<19-87 Letter, NRC (Edison) to 8806030191 5-19-88 FPL ( Conway)
- 3. VERIFICATION DOCUMENTS:
TITLE NUDOCS NO. DATE
PLANT Turke Point 3 DOCKET NO(S). 50-250 PROJECT MANAGER Gordon Edison TECHNICAL CONTACT B. Elliott USI RO. A-11 TITLE Reactor Vessel Materials Too hoess MPA NO. A-007 TAC NOS. 68249 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 possibi lity 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-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):
In an internal memo form Shao to Murley, dated 9-24-87, it was documented that Turkey Point Units 3 and 4 were believed to have a USE less than 50 ft-lb. This had been recognized several years earlier by FPL in a letter dated 2-3-83 (Uhrig to Eisenhut). FPL submitted fracture toughness analyses of beltline welds, as required by 10 CFR 50, Appendix G, in letters dated 5-3-84 (Williams to Eisenhut, Pro rietar ), and 3-25-86 (Woody to Thompson, Pro rietar ). NRC transmitted sa e y eva uations dated 10-30-87 and 5-31-88 in ica ing additional information was needed. FPL responded in letters dated 5-4-88 and 1-31-89 (Conway to NRC) and 6-16-89 (Woody to NRC) which indicated adequate margin of safety for continued operation and described plans for further tests and analyses. In an internal memo dated 5-4-89 (Shao to Varga), the NRC Division of Engineering and Systems Technology provided a safety evaluation justifying continued operation with USE less than 50 ft-lb.
REFERENCES:
Plant Name 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 DOCUt IENTS:
TITLE NUDOCS NO. DATE Uhrig (FPL) to Eisenhut (NRC) 8302070305 2-3-83 Williams (FPL) to Eisenhut 8405090058 5-3-84 to Thompson (NRC)
Woody (FPL) to Thompson (NRC) 8603310205 3-25-86 McDonald (NRC) to Woody (FPL) 8711050220 10-30-87 Edison (NRC) to Conway (FPL) 8806070211 5-31-88 Conway (FPL) to NRC 8805160014 5-4-89 Conway (FPL) to NRC 8902070160 1-31-89 Moody to NRC 8906210029 6-16-89 Shao to Murley 9-24-87 Shao to Varga 5-4-89
- 3. VERIFICATION DCCUMENTS:
TITLE NUDOCS NO. DATE
PLANT Turke Point 3 DOCKET NO(S) . 50<<250 PROJECT MANAGER Gordon Edison TECHNICAI CONTACT D. Thatcher USI NO. A-17 TITLE S stems Interactions in Nuclear Power Plants MPA NO. NOS. None ISSUES
SUMMARY
'AC 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 di d not require any licensee acti ons.
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 P l an Item I.C. 5 (NUREG-0737) .
IMPLEMENTATION AND STATUS
SUMMARY
(PLANT SPECIFIC:
In letters dated 11-6-72, 11-17-72, 1-5-73, 1-7-75, 8-18-75, 3-6-79, the licensee provided information regarding systems interactions and particularly water intrusion and internal flooding. These letters were in response to NRC concerns and evaluations documented in letters dated 9-26-72 and 12-5-74. In a letter dated 9-4-79 the NRC staff issued a safety evaluation which concluded that a sufficient level of protection is provided from flooding of equipment important to safety-related systems, and that no further action was required by the licensee.
REFERENCES:
P lant name A-17
- 1. RE UIREMENT DOCUMENTS:
TITLE NUDOCS NO. DATE Generic Letter 89-18 09/06/89 HUREG-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" Letter, DeYoung (NRC) to 9-26-72 Coughlin (FPL)
- 2. IMPLEMENTATION DOCUMENTS:
TITLE NUDOCS NO. DATE Letter, Coughlin (FPL) to 11-6-72 DeYoung (NRC)
Letter, Schmidt (FPL) to 11-17-72 O'eary (NRC)
Letter, Coughlin (FPL) to 1-5-73 DeYoung (NRC)
Letter, Uhrig (FPL) to 1-7-75 Lear (NRC)
Letter, Uhrig (FPL) to 8-18-75 Lear (NRC)
Telecopy, Whittier (FPL) to 3-6-79 Grotenhuis (NRC)
Letter Schwencer (RRC) to 9-4-79 Uhrig FPL)
- 3. VERIFICATION DOCUflEfITS:
TI TLE NUDOC NO. DATE
PLANT Turke Point 3 DOCKET NO(S). 50-250 PROJECT MANAGER Gordon Edison TECHNICAL CONTACT P. Shemanski USI NO. A-24 TITLE ualification of Class 1E Equi ment MPA NO. 8-60 TAC NOS. None 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; it does supplement, in selected areas, the provisions of the 1971 and 1974 versions of 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 (SERI: 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):
On 5-21-81 the NRC staff issued an SER based on FPL submittals dated 6-2-80, 7-3-80, 10-31-80, and 2-27-81. On 12-13-82 another NRC staff SER was issued based on licensee submittals dated 8-27-81, 9-14-81, 2-9-82, and 3-24-82. FPL provided additional information on 9-24-82, 2-1-83 and 7-12-84. In the 7-12-84 submittal the licensee indicated that they believed they had resolved all issues. The NRC staff SER dated 10-25-84 documented compliance with 10 CFR 50.49 and acceptable proposed resolution of deficiencies. The staff inspected implementation in three IR's (87-08, 88-27, and 88-38). FPL responded to a Notice of Violation in a letter from Conway (FPL) to NRC dated 3-9-88. The latest verification was provided in IR 88-38 dated 1-6-89.
REFERENCES:
Plant Name A-24
- 1. RE UIREMENT DOCUMENTS:
TITLE NUDOCS NO. DATE DOR "Guidelines for Evaluating Environmental gualification 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 and Review Plan 3.11, Environmental qualification of Mechanical and Electrical Equipment 07/81
- 2. IMPLEMENTATION DOCUMENTS:
TITLE NUDOCS NO. DATE Letter, Williams (FPL) to 8407170439 7-12-84 to Denton (NRC)
Letter, Varga (NRC) to Uhrig 8304060619 3-29-83 (FPL)
Letter, Varga (NRC) to Uhrig 8212290534 12-13-82 (FPL)
Letter, Varga (NRC) to Uhrig 8106010421 5-21-81 (FPL)
Orders for Modification, 8011110522 10-24-80 Varga (NRC) to Uhrig (FPL)
Conway (FPL) to NRC 8803140225 3-9-88 Letter, Varga (NRC) to FPL 8411090058 10-25-88
- 3. VERIFICATION DOCUMENTS:
TITLE NUDOCS NO. DATE IR 8 50-250,251/87-08 8707240211 7-21-87 IR 8 50-250,251/88-27 8811030126 10-26-88 IR 8 50-250,251/88-38 8901130088 1-6-89
PLANT Turke Point 3 OOCKET NO(S). 50-250 PROJECT MANAGER Gordon Edison T<< IH L ONTA T ~h USI NO. A-26 TITLE Reactor Vessel Pressure Transient Protection MPA NO. B-04 TAC NOS.
ISSUES
SUMMARY
This USI was resolved in September 1978 with the publication of NUREG-0224, "Reactor Vessel Pressure Transient Protection for PWRs," and Standard Review Plan Section 5.2. The licensees of all operating PWRs were requested to provide an overpressure prevention system that could be used whenever the plants were in startup or shutdown conditions. The issue affected all operating and future plants, and the staff established MPA B-04 for implementing the solution at operating PWRs.
Since 1972, there have been numerous reported incidents of pressure transients in PWRs where technical specification pressure and temperature limits have been exceeded. The majority of these events occurred while the reactors were in a solid-water condition during startup or shutdown and at relatively low reactor vessel temperatures. Since the reactor vessels have less toughness at lower temperatures, they are more susceptible to brittle fracture under these condi-tions than at normal operating temperatures. In light of the frequency of the reported transients and the associated potential for vessel damage, the NRC staff concluded that measures should be taken to minimize the number of future transients and reduce their severity.
Generic Letter 88-11, "NRC Position on Radiation Embrittlement of Reactor Vessel Materials and its Impact on Plant Operations," was published July 12, 1988. This generic letter provides guidance regarding review of pressure-temperature limits and indicates that licensees may have to revise low-temperature-over pressure protection setpoints.
IMPLEMENTATION AND STATUS
SUMMARY
PLANT SPECIFIC):
On 3-14-80 the NRC staff issued Amendments No. 55 and 47 to Units 3 and 4.
The SER in those amendments found that the proposed Turkey Point Overpressure Mitigation System (OMS) and the proposed Tech Specs were acceptable. These design mods had been proposed by FPL in letters dated 12-11-78 and 10-18-77, as well as a series of submittals referenced in the NRC SER in Amendments 55 5 47. In Amendments 134 and 128, dated 1-10-89, the staff reviewed the affect of new P/T limits on the OMS and found it acceptable.
REFERENCES:
Plant Name A-26
- l. RE UIREMENT DOCUMENTS:
TITLE RUOOCS RO. DATE NUREG-0224 - "Reactor Vessel Pressure Transient Protection for PWRs." 9/78 NRC Letters to Licensees Informing Licensees of Staff Concerns Regarding Overpressure Low-Temperature Conditions in PWRs August 1976 Generic Letter 88-11, "NRC 7/12/88 Position on Radiation Embrittlement of Reactor Vessel Materials and Its Impact on Plant Operations" Standard Review Plan Section 5.2
- 2. IMPL EMENTATION DOCUMENTS:
TITLE NUDOCS NO. DATE Letter, Uhrig (FPL) to 10-18-77 Lear (NRC)
Letter, Uhrig (FPL) to 12-11-78 Lear (NRC)
Letter, Schwencer (NRC) 8004240251 3-14-80 to Uhrig (FPL) transmitting Amendments 55 and 47.
Letter McDonald (NRC) to 8301060527 12<<23-82 Uhrig (PPL) (Corrected SER)
Letter, Edison (NRC) to 8901190137 1-10-89 Conway (FPL)
- 3. VERIF ICATION DOCUMENTS:
TITLE NUDOCS NO. DATE
PLANT Turke Point 3 DOCKET NO(S). 50-250 PROJECT MANAGER Gordon Edison TECHNICAL CONTACT J. Wermiel USI NO. A-36 TITLE Control of Heav Loads Phases I 8 II NPA NO. C-10 C-15 TAC NOS. Naae 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 todrop in certain locations in the p'tant, 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 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:
- 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 r educe 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.
IMPLEMENTATION AND STATUS SUGARY (PLANT SPECIFIC'he NRC staff issued an SER dated 11-1-83 which concluded that Control of Heavy Loads, Phase I, was acceptable. This was based on submittals from FPL dated 3-15-83, 8-15-83, and 10-7-83 which described details of load handling equipment and operations. The licensee has implemented all modifications agreed to with the staff, except the installation of a separate load cell for the head lift rig. That action is included in the Integrated Schedule and is currently scheduled to be implemented in an outage which is scheduled to start 1/31/91.
REFERENCES:
P lant Name A-36
- 1. RE UIREMENT DOCUMENTS:
TITLE NUDOCS IID. DATE Letter, Darrell G. Eisenhut, NRC, to a11 licensees, app licants 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 06/28/85 Plants'UREG-0612"
- 2. IMPLEMENTATION DOCUMENTS:
TITLE NUDOCS NO. DATE NRC to FPL 8303040267 2/22/83 FPL (Uhrig) to NRC 8303170504 3/15/83 FPL (Uhrig) to NRC 8308190381 8/15/83 FPL (Uhrig) to NRC 8310180381 10/7/83 McDonald (NRC) to Uhrig (FPL) 8311290358 11/1/83
- 3. VER IF ICATION DOCUMENTS:
TITLE NUDOCS NO. DATE
PLANT Tur ke Point 3 DOCKET HO (S) . 50-250 PROJECT MANAGER Gordon Edison TECHNICAL CONTACT P. Gi 1 1 US I NG. A-44 TITLE Station Blackout MPA HO. A-022 TAC NOS. 68618 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. 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 rule was published for cogent 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 docurrent, 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 station blackout 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)(l) 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):
On April 17, 1989 in a letter from Conway to NRC, FPL responded to the requirements of the new rule (10 CFR 50.63). In October, 1989 the HRC staff audited FPL's analytical basis for their planned modifications. The NRC had not issued an SER as of 12-15-89, but anticipates issuing an SER by 3/31/90.
The impleoantation date on A-44 is expected to be not later than 3/31/92, based on the estimate for staff completion of the plant-specific SER and allowing up to 2 years for the licensee to complete iayleoantation.
REF ERENCES: Plant Name A-44
- 1. RE UIREMENT DOCUMENTS:
TITLE NUDOCS NO. DATE 10 CFR 50.63, "Loss of All Alternating Current Power" 06/21/88 Regulatory Guide.l.155, "Station Blackout" 08/88
- 2. IMPLEMENTATION DOCUMENTS:
TITLE NUDOCS HO. DATE Letter, Conway (FPL) to NRC 8904250001 4-17-89
- 3. V ER IF ICATION DOCUMENTS:
TITLE NUDOCS NO. DATE
PLANT Turke Point 3 DOCKET NO(S). 50-250 PROJECT NNAGER Gordon Edison TECHNICAL CONTACT P. Y. Chen USI NO. A-46 TITLE Seismic gualification of Equipment in Operating Plants l'IPA NO. 8-105 TAC NOS. 68303 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 (S(UG) 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 SHRUG and the staff on a generic basis before individual member utilities proceed with p lant-speci f i c imp 1 ementati on.
IMPLEMENTATION AND STATUS
SUMMARY
(PLANT SPECIFIC):
The Turkey Point licensee (FPL) is not a member of SHRUG. A separate approach to GL 87-02 has been taken by FPL. In letters dated 5/15/87, 10/1/87, and 2/1/88 FPL has responded to GL 87-02. The licensee and the NRC staff appear to have a fundamental disagreement over the importance of the probability and intensity of seismicity in the Turkey Point area. An appeal meeting was held on 6/2/88, and the staff agreed to consider a scaled-back program of equipment walkdowns and reviews. Such a program was submitted by FPL on 8/4/88. In a letter dated 8/4/89 the NRC staff agreed to the FPL program subject to a number of conditions. FPL responded to the staff on 10/2/89 and committed to a supplemental response by 12/15/89. Walkdowns would begin during an outage no earlier than mid '90.
REFERENCES'. Plant Name A-46
~RE UIREHENT 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
- 2. IMPLEMENTATION DOCUMENTS:
TITLE NUDOCS NO. DATE Letter, Woody (FPL) to NRC 8705210604 5-15-87 Letter, Woody (FPL) to NRC 8710070004 10-1-87 Letter, Woody (FPL) to NRC 8802050097 2-1-88 Summary of Meeting 8807050455 6-16-88 Letter, Edison and Norris (NRC) 8908140423 8-4-89 to Woody (FPL)
- 3. VERIFICATION DOCUMENTS:
TITLE NUDOCS NO. DATE
- 1. ~0
REFERENCES:
TITLE NUDOCS NO.
P lant A-46 Name DATE Generi c 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 Equ ipment in Operati ng Plants, Unresolved Safety Issue A-46" 02/87
- 2. IMPLEMENTATION DOCUMENTS:
TITLE NUDOCS NO. DATE Letter, Woody (FPL) to NRC 8705210604 5-15-87 Letter, Woody (FPL) to NRC 8710070004 10-1-87 Letter, Woody (FPL) to NRC 8802050097 2-1-88 Summary of Meeting 8807050455 6-16-88 Letter, Edison and Norris (NRC) 8908140423 8-4-89 to Woody (FPL)
- 3. VERIFICATION DOCUMENTS:
TITLE NUDQCS NO. DATE
PLANT Turke Point 3 DOC KET NO(S) . 50-250 PROJECT MANAGER Gordon Edison TECHNICAL CONTACT J. Mauck USI NO. A-47 TITLE Safety Implication of Control Systems in LWR Nuclear Power Plants MPA NO. TAO NOS. None ISSUES
SUMMARY
USI A-47 was resolved September 20, 1989, with the publication of Generic Letter (GL) 88-19.
The generic let'ter 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 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 BOW plants. The generic letter provides amplifying guidance for licensees.
The generic letter requires that licensees provide NRC with their schedule ard 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):
The licensee plans to respond to GL 88-19 by March, 1990 as required.
REFERENCES:
Plant name A-47
- 1. REIEUIREMENT 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 LMR Nuclear Power Plants" NUREG-1218 "Regulatory Analysis July 1989 for Re so 1u ti on of US I A-47"
- 2. IMPL EMENTATI ON DOCUMENTS:
TITLE NUDOCS NO. DATE
- 3. VER IF I CAT ION DOCUMENTS:
TITLE NUDOCS NO. DATE
m ss PLANT Turke Point 3 DOCKET NO(S). 50-250 PROJECT MANAGER Gordon Edison TECHNICAL CONTACT B. Elliott USI IIO. 4-49 TITLE Pressurized Thermal Shock MiPA NO. A-021 TAC NOS. 59992 ISSUES
SUMMARY
The final rul~ (10 CFR 50.61) on pressurized thermal shock (PTS) was approved by the Commission in July 1985. Regulatory Guide le154, "Format and Content of Plant-Specific Pressurized Thermal Shock Safety Analysis Reports for Pk!Rs,"
was later published in February 1987. Thus, this issue was resolved and new requirements were established, applicable to PHRs only. The rule required that each operating reactor meet the screening criteria provided in the rule or provide supplemental analysis to demonstrate that PTS is not a concern for the faci 1 i ty.
Neutron irradiation of reactor pressure vessel weld and plate materials decreases the fracture toughness of the materials. The fracture toughness sensitivity to radiation-induced change is increased by the presence of certain materials such as copper. Decreased fracture toughness makes it more likely that, if a severe overcooling event occurs followed by or concurrent with high vessel pressure, and if a small crack is present on the vessel's inner surface, that crack could grow to a size that might threaten vessel integrity.
Severe pressurized overcooling events are improbable since they require multiple failures and improper operator performance. However, certain precursor events have happened that could have potentially threatened vessel integrity if additional failures had occurred and/or if the vessel had been more highly irradiated. Therefore, the possibility of vessel failure due to a severe pressurized overcooling event cannot be ruled out.
IMPLEMENTATION AND STATUS
SUMMARY
(PLANT SPECIFIC):
In letters dated 1-23-86, 6-5-86, and 7-22-86, the licensee responded to the new PTS rule (10 CFR 50.61) for Units 3 5 4, and stated that their analyses ana measurements showed they met requirements. The NRC staff issued an SER dated 3-11-87 which described NRC's independent calculations which confirmed that the licensee met the requirements.
REFERENCES:
Plant Name A-49
- 1. RE(EUIREMENT DOCUMENTS:
TITLE NDDGCS NO. DATE 10 CFR 50.61, "Fracture Toughness 7/85 Requirements for Protection Against Pressurized Thermal Shock Requirements" Reg. Guide 1.154, "Format and Content 1/89 of Plant-Specific Pressurized Thermal Shock Safety Analysis Reports for PWRs" SECY 82-465, "Pressurized Thermal Shock" 11/23/82 SECY 83-288, "Proposed Pressurized Thermal Shock Rule" 07/15/83 Regulatory Guide 1.154 "Format and Content of Plant-Specific Pressurized Thermal Shock Safety Analysis Reports for Pressurized Water Reactors" 02/87 Generic Letter 88-11, "NRC Position on Radiation Embrittlement of Reactor Vessel Materials ana Its Impact on Plant Operations" 7/12/88
- 2. IMPLEMENTATION DOCUMENTS:
TITLE NUDOCS NO. DATE Letter, Woody (FPL) to 8601290065 1-23-86 Thompson (NRC)
Letter, Woo+ (FPL) to 8606100358 6-5-86 Thompson (NRC)
Letter, Woo+ ,(FPL) to 8607280057 7-22-86 McDonald (NRC)
Letter HcDooald (NRC) to 8703190152 3-11-87 Needy FPL)
- 3. V ER IF ICATION DOCUMENTS:
TITLE NUDOCS NO. DATE
ENCLOSURE 3 Pa9e No.
02/07/90 I.IST?NS OF INCOMPLETE USI DATA FOR INPUT FROtt PROJECT MANAGERS ISSUE ISSUE DESCRIPTIVE HAHE INPLENEttT INPLEHEHT LICENSEE COMMENT STAFF CONNEHT HUNGER DATE STATUS tf PLANT ttAHi: TURKEY POINT 3 A"01 MATER HAMMER / / HC A-02 ASYMMETRIC BLQMDGMH LOADS QN / I NC LEAK BEFORE BREAK REACTOR PRIMARY COOLANT SYS'TENS A-03 MESTIHSHOUSE STEAN GENERATOR TUBE / / NC IHFO ONLY INTEGRITY A"04 CE STEAN SEHERATOR TUBE INTEGRITY H/A CE PLANTS ONLY A-05 BN S'TEAN GENERATOR TUBE H/A BIM PLANTS ONLY lttTEGRITY A-06 NARK I SHORT-TERN PRQ6RAN I / H/A NK I BMR ONLY A-07 NARK I LONG-TERN PROGRAN / I N/A HK I BMR ONLY A-08 NARK I I CONTAINMENT POOL DYNANIC / / H/A A(K II BMR ONLY LOADS - LONG-TERN PROGRAN A-09 AIMS 12/31/91 I A-10 BMR FEEDMATER NOZZLE CRACKIN6 / / H/A BMR ONLY A-II REACTOR VESSEL MATERIALS 12/31/92 I ( 50 FT-LB HO CRITERIA TOUGHNESS A-12 FRACTURE TQU6HHESS OF STEAN / / H/A CP AFTER 83 OttLY GEHERATOR AND REACTOR CQQLAN'T PUMP SUPPORTS A-17 SYSTEMS IHTERACTIOH / / NC HO REQUIREttENTS A-24 QUALIFICATION GF CLASS IE 07/12/84 C SAFETY-RELATED EQUIPtlEHT A-26 REACTOR VESSEL PRESSURE TRANSIENT 03/14/80 C LTOPS PROTECTION A-31 RHR SHUTDQMH REQUIREMENTS / I NIA NEM PLANTS ONLY. SRP.
A-36 CONTROL OF HEAVY LOADS HEAR SPENT 12/31/'91 I LOAD CELL FUEL A-39 DETERNIHATIOH OF SAFETY RELIEF / / H/A BMR ONLY VALVE POOL DYHANIC LOADS AND TEMPERATURE LINITS A-40 SEISMIC DES16H CRITERIA- / / HC NOT SQUB SHQRT-1ERH PROGRAM A-42 PIPE CRACKS IH BOILIN6 MATER / / N/A BMR ONLY REACTORS A-43 CONTAINMENT ENER6EHCY SUMP / I NC IHFO ONLY PERFGRttANCE A-44 STATION BLACKOUT 03/31/92 I D6 ADDITIOH SER 3/31/90 A-45 SHUTDOWN DECAY HEAT REttOVAL / / NC SUBSUMED BY SEVERE ACC REQUIRENEtlTS A-46 SEISNIC QUALIFICATIOH OF 12/31/91 I REQ UNDER DEVEL EQUIPHEHT IH QPERATINB PLANTS A-47 SAFETY INPLICATIONS OF CONTROL 03/20/90 E NEM REQUIREttEHTS SYSTENS A-48 HYDROGEN CONTROL ttEASURES AHD / / H/A N/A DRY CONTAIN EFFECTS OF HYDROGEN BURNS ON SAFETY EQUIPttENT A-49 PRESSURIZED THERNAL SHOCK 07/22/86 C