ML20033F189
| ML20033F189 | |
| Person / Time | |
|---|---|
| Site: | Dresden, Braidwood  |
| Issue date: | 03/02/1990 |
| From: | Sands S Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML17202J231 | List: |
| References | |
| REF-GTECI-A-09, REF-GTECI-A-44, REF-GTECI-A-47, REF-GTECI-A-49, REF-GTECI-EL, REF-GTECI-RV, REF-GTECI-SY, TASK-A-09, TASK-A-44, TASK-A-47, TASK-A-49, TASK-A-9, TASK-OR GL-89-12, NUDOCS 9003160376 | |
| Download: ML20033F189 (41) | |
Text
,
.L - ** =egk UNITED STATas f
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Nuct. EAR RESULATORY COMMISSION af memwoTow. o. c. rosss Docket flos. 50-456 and 50-457 MEtiORAllDUM FOR:
File FPOM:
Stephen P. Sands, Projects 11anager Project Directorate III-2 Division of Reactor Projects - 111, IV, Y and Special Projects
SUBJECT:
STATUSOFIMPLE!4ENTATIONOFUNRESOLVEDSAFETYISSUES(USI)
AT BRAIDWOOD STATION, UllITS 1 AND 2 l
The current implementation status of unresolved safety issues (USIs) at the Braidwood Station, Units 1 and 2, is set forth in the enclosures to this l
mer.orandum.
l l contains a copy of the information provided by the licensee, ComonwealthEdisonCompany(Ceco),intheirresponsetoGenericLetter89-12 (GL ft9-12).
Additionally, Enclosure 2 contains a status summary for each USI applicable to this facility. This status summary is based upon the licensee's response to the Generic Letter, discussions with the licensee, and my review of available llRC records and information. Appropriate flRR technical branches have aisc reviewed the-USI sumary and this inemo.
For those items that are incomplete, y assessment of safety significance is as follows:
USI NUMBER TITLE STATUS A-9 K per 10 CFR 50.62 To be impler.iented during next refueling outages A-44 Station Blackout Licensee's 4/17/89 response is under staff review A-47 Safety Implications of Generic Letter 89-19 was recently Control Systems in LWR sent to licensee; response due 3/90 Nuclear Power Plants l
L A-49 Pressurized Thermal Shock Licensee's 1/17/86 response is under staff review.
File 2-i It is my conclusion that there is no urgent safety concerns for these USIs for i
Braidwood Station while awaiting completion.
Step n P. Sands, Projects Manager i
Project Directorate III-2 i
Division of Reactor Projects - III, IV, V and Special Projects i
Enclosures:
As stated cc w'/ enclosures:
'R. Wessman i
K. Eccleston J
.R.
Herman l
R. Sebc11 l
W. Paulson i
i 1
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e
Tile 2
It is my conclusion that there is no urgent safety concerns for these USIs for Braidwood Station while awaiting completion.
Stephen P. Sands, Projects Manager Project Directorate 111-2 Division of Reactor Projects - III, IV, Y and Special Projects
Enclosures:
As stated cc w/ enclosures:
R. Wessmen.
K. Eccleston l
R. Herman i
R. Scholl W. Paulson 1
1 DISTRIBUTION II6cTet File -
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DOCultENT tlAf tE:
[BRAIDWOODUSI]
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PLANT Braidwood 1/2 DOCKETN0($).
50-456/457 PROJECT MANAGER S. P. Sands TECHNICAL CONTACT A. Serkir USI NO. A-1 TITLE Water Hamer MPA NO. N/A TAC NOS. None ISSUES
SUMMARY
This Unresolved Safety Issue (USI) was resolved in March 1984, with the publication of NUREG-0927. " Evaluation of Water Hamer in Nuclear Power Plants
- Technical Findings Relevant to Unresolved Safety Issue A-1.'
Also on March 15,1984, the EDO sent the Comissioners SECY 84-119 titled, " Resolution of Unresolved Safety issue A-1, Water Hamer."
In SECY 84-119, the staff concluded that the frequency and severity of water hamer 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 _ventino procedures; (b) proper design of feed-water valves and control systemst 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.
JtdidinvolveStandardReviewPlan(SRP) changes (forward fits) and a cumprehensive set of guidelines and criteria to evaluate and upgrade utility training programs (per TMl 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 hamer that opened a direct pathway to the environment.
. IMPLEMENTATION AND STATUS
SUMMARY
(PLANT SPECIFIC):
Byron SER (NUREG-0876) dated February 1982, pages 13-24 staff concluded licensee had satisfied the requirements of I.A.2.3.
Licensee also comitted to I.A.2.3 in a letter dated 10/61.
Section 10.4.7 of Cyron SER (NUREG-0876) discusses Water Hammer.
Implementation considered date of low power license:
Braidwood 1: 05/21/87 Braidwood 2:
12/18/87
i
REFERENCES:
Braidwood 1/2 A-1 i
1.
REOUIREMENT DOCUMENTS' TITLE NUDOCS NO.
DATE Letter from Denton to Utilities.
8403150310 03/05/84
" Notice of Issuance and i
Availability NUREG-0927 Rev. 1,
)
Safety Issue A-1" j
2.
IMPLEMENTATION DOCUMENTS:
)
i TITLE NUDOCS NO.
DATE NUREG-0927 " Evaluation of Water 8306060413 05/31/83 Hamer 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, i
10.3, and 10.4.7 j
- Resolution 03/15/84 of Unresolved Safety A-1,
)
Water Hammer" Letter from licensee 8110130351 10/5/81 ByronSER(NUREG-0876) 2/82
]
3.
VERIFICATION DOCUMENTS:
TITLE NUDOCS NO.
DATE
)
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l PLANT Braidwood 1/2 DOCKETN0(S). 50-456/457 PROJECT PMAGER S. P. Sands TECHNICAL CONTACT Jai Rajan USI NO. A-2 TITLE Asymmetric Blowdown Leeds in RCS MPA NO. D-10 TAC N05.
1SSUES
SUMMARY
This USI was resolved in January 1981 with the publication of NUREG-0609,
" Asymmetric Blowdown Loads on PWR Primary Systems."
In October 1975, the NRC notified each operating PWR licensee of a potential safety problem concerning the fact that asymmetric LOCA loads had not been considered in the design of any PWR 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.
i NUREG-0609 provided guidance for these analyses. For operating PVPs, Multi-Plant Action (MPA) Item D-10 was established by NRC's Division of Licensing for implementation purposes.
During the course of the wntk 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 Register in final form on April 11, 1986, and in a subsequent revision to GDC'1 published in the Federal Register 1
l on July 23, 1986.
In addition, it has also been satisfactorily demon'strated in
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the course of the A-2 effort that there is a very low likelihood of simultaneous pipe leading with both LOCA and safety shutdown earthquake (SSE) loads.
Therefore, the lest revision of GDC-4 represented the final technical action of NRC regarding the issue of asymmetric blowdown loads issue in PWRs primary coolant main loop piping.
IMPLEMENTATION AND STATUS
SUMMARY
(PLANT SPECIFIC):
This was covered in the Byron SER (NUREG-0876, dated 2/82) Section 3.9.2.4 -
Exemption from GDC-4 with respect to the need to analyze large primary loop i
pipe ruptures as structural design basis was issued on 10/28/85. However, the i
restraints and deflectors were removed after the rule was changed, and the issue was closed prior to licensing.
i 1
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REFERENCES:
Braidwood 1/2 A-2 1.
REQUIREMENT DOCUMENTS:
TITLE NUDOCS NO.
DATE Generic Letter " Evaluation of Primary Systems for Asymetric 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 l
r
" Asymmetric Blowdown Loads on PWR
~
Primary Systems," NUREG-0609 US NRC HRR 01/81 GDC-4, " Environmental and Dynamic Effects Design Basis" GL 84-04, " Safety Evaluation of i
Westinghouse Topical Reports Dealing t
WithEliminationofPostulatedPige Breaks in PWR Primary Main Loops.
2.
IMPLEMENTATION DOCUMENTS:
i TITLE NUDOCS NO.
DATE GDC-4 exemption, Byron 2 8511040476 10/28/85 4
3.
VERIFICATION DOCUMENTS:
TITLE NUDOCS NO.
DATE e.
t W
f e
1 PLANT Braidwood 1/2 DOCKET N0($) 50;456/457 l
PROJECT MANAGER S.P. Sands TECHNICAL CONTACT E. Murphy
{
t USI NO. A-3, A-4, and A-5 TITLE Steam Generator Tube Integrity MPA NO.
TAC NOS.
ISSUES
SUMMARY
USIs A-3, 4, and 5, were resolved in September 1988 with the publication of NUREG-0844 "NRC Integrated Program for the Resolution of Unresolved Safety Issues A-3, A-4, and A-5 Regarding Steam Generator Tube Integrity." USIs A-3, L
L A-4, and A-5 did not result in new generic requirements for industry in view of the small potential for reducing risk.
h Steam generator tube integrity was designated an unresolved safety issue in widespread degradation, frequent leaks, and occasional ruptures (ject to 1978 after it became apparent that steam generator tubes were sub i.e., gross failures). USI-Task Action Plans A-3, A-4, and A-5 were established to
)
l evaluate the safety significance of these problems for Westinghouse, Combustion i
i Engineering, and Babcock & Wilcox steam generators, respectively. These studies were later combined into a single effort because PWR vendors were all
]
experiencing many of the same problems.
l 1
l NUREG-0844 provides a generic risk assessment that indicates that risk from I
steam generator tube rupture (SGTR) events is not a significant contributor to j
the total risk at a given site, nor to the total risk to which the general public is-routinely exposed. This finding is considered indicative of the effectiveness of licensee programs and regulatory requirements for ensuring steam generator tube integrity in accordance with 10 CFR Part 50, Appendices A and B.
j NUREG 0844 also identifies a number of staff-recomended actions that can further improve the effectiveness of licensee programs in ensuring the integrity of steam generator tubes and in mitigating the consequences of a SGTR. As part of the integrated program, the staff issued Generic Letter 85-02 encouraging licensees of PWRs to upgrade their programs, as necessary, to meet the intent of the staff-recomended actions; however, such recomended actions do not constitute NRC requirements. The staff's assessment of licensee responses to Generic Letter 85-02 was provided to the Comissir.,n in SECY 86-97.
IMPLEMENTATION AND STATUS
SUMMARY
(PLANT SPECIFIC):
The Byron SER (NUREG-0876, dated 2/82) on pages 5-21 and C-9 and supplement No. 5 to the SER, issued 10/84, page 5-2, discusses the integrity of the steam generator tubes. Licensee response - letters dated 6/17/85 and 8/22/85.
Response was found acceptable - No SER sent to licensee.
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REFERENCES:
Braidwood 1/2-A-3, 4, 5 1.-
REQUIREMENT DOCUMENTS:
TITLE NUDOCS NO.
DATE NUREG-0844, "NRC Integrated September 1988 Program for the Resolution of Unresolved Safety issues A-3, A-4, and A-5 Regarding Steam Generator Tube Integrity" Generic Letter 85-02 04/17/85 SECY-86-97, Steam Generator USI Program - Utility Responses to Staff Recommendations in Generic
- Letter 85-02 03/04/86 2.
IMPLEMENTATION DOCUMENTS:
TITLE NUDOCS NO.
DATE Letter from licensee 8506240527 06/17/85 Letter from licensee 8508260099 08/22/85 Byron SER (NUREG-0876).
02/82 Byron SSER 5 10/84 3.
VERIFICATION DOCUMENTS:
TITLE NUDOCS NO.
DATE 4
f
l
. PLANT Braidwood 1/2 DOCKETN0(S).
50-456/457 PROJECT ltANAGER S. P. Sands TECHNICAL CONTACT J. Mauck i
USI fl0. A-9 TITLE ATWS per 10 CFR 50.62 MPA NO.
TAC NOS. 64035. 64062 ISSUES
SUMMARY
j This USI was resolved in June 1984 with the publication of a final rule (10 CFR 50.62)torequireimprovementsinplantstoreducethelikelihoodoffailureof thereactorprotectionsystem(RPS)toshutdownthereactorfollowing anticipated transients and to mitigate the consequences of an anticipated I
transient without scram (ATWS) event.
I The rule includes the following design-related requirements:
50.62(C)(1),
diverse and ' independent auxiliary feedwater initiation and turbine trip for all PWRs;50.02(C)(2),diversescramsystemsforCEandBAWreactors;50.62(C)(3) alternaterodinjection(ARI)forBWRs;50.62(C)(4);standbyliquidcontrol system (SLCS) for BWRs; and 50.6?(C)(5), automatic trip of recirculation purps under conditions indicative of an ATWS for BWRs.
Information requirements and i
an implementation schedule are also specified.
1 l
IMPLFFINTATION AND STATUS SUMt!ARY (PLANT SPECIFIC):
Staff SER issue 6/13/89. By letter dated 8/16/89 licensee j
proposed 4 changes to ATWS design bases on it human factors engineering review. Staff. approved changes in letter dated 9/12/89. Schedule for implementation submitted 2/15/89.
Unit 1 - 03/90 Unit 2 - 10/90
.)
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REFERENCES:
Braidwood 1/2 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
^49FR26045(10CFR50.62) 06/26/84-1,.
lt 2.
IMPLEMENTATION DOCUMENTS:
TITLE NUDOCS NO.
DATE SER 8906190272 6/13/89 Letter on changes 8909200051 9/12/89 3..
VERIFICATION DOCUMENTS:
TITLE NUDOCS NO.
DATE m
t
PLANT Braidwood 1/2 DOCKETN0(S). 50-456/457 PROJECT MANAGER S P. Sands' TECHNICAL CONTACT B. Elliott 2
USl NO. A-11 TITLE Reactor Vessel Materials Toughness MPA NO.
TAC N05.
ISSUES
SUMMARY
This USI was resolved in October 1982 with the publication of NUREG-0744,
" Pressure Vessel Paterial 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.
l 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.
i l
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 i
have beltline materials with Charpy upper shelf energy less than 50 ft-lb.
l 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 be'ow 50 ft-lb.
l IMPLEMENTATION AND STATUS
SUMMARY
(PLANT SPECIFIC):
BraidwoodSER(NUREG-1002)page5-4.
Staff approved revised heatup and cooldown curves. These curves were issued with the licensee as part of the Technical Specifications.
Unit 1 --5/21/87 Unit'2 - 12/18/87 b
I.
is
REFERENCES:
Braidwood 1/2 A-11
').
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 BraidwoodSER(NUREG-1002) 11/83 Done in conjunction with Braidwood 1/2 low power license issue dates:
Unit 1 05/21/87 Unit 2 12/18/87 3.
VERIFICATION DOCUMENTS:
TITLE NUDOCS NO.
DATE f
N I
PLANT Braidwood 1/2 DOCKETN0(S).
50 456/457 PROJECT MANAGER S. P. Sands TECHNICAL CONTACT R. Johnson (RES)
US1 NO. A-12 TITLE Potential of Low Fracture Toughness and Lamellar Tearing in PWR SG and RCP Supports MPA NO.
TAC NOS.
k ISSUES
SUMMARY
t This USI was resolved in October 1983 with the publicativn of NUREG-0577,
" Potential of Low Fracture Toughness and Lamellar Tearing in PWR Steam s
Generator and Reactor Coolant Pump Supports." The resolution contained no backfit requirements; it only applied to plants with a new construction permit issued after October 1983.
Standard Review Plan Section 5.3.4 was issued at the same time this USI was resolved.
The concern in this'US1, as the title indicates, was the potential of low fracture toughness of some materials selected for fabrication of steam penerator(SG)andreactorcoolantpump(RCP)supportsinoperatingPWRs.
Lamellar tearing was also of concern.
Fracture toughness is a measure of a material's resistance to fracture in the presence of a previously existing crack.
Generally, a material is considered to have adequate fracture toughness if it can withstand loading to its design limit in the presence of detectable j'
flaws under stated conditions of stress and temperature.
1he modifications to address this 051 could involve maintaining minimum t aperature around the supports above its fracture transition temperature, or total replacement of existing SG and RCP supports with supports fabricated of material grade which has a higher Charpy upper shelf energy and a lower transition temperature. Analysis performed for the resolution of this USI determined that, even with the failure of the SG and RCP supports, the amount of incremental release of radioactivity would not be sufficiently high enough to justify any modification in terms of increasing the toughness of these supports.- This conclusion is based on a value-impact analysis docurented in Appendix C of HUREG-0577.
lHPLEMENTATION AND STATUS
SUMMARY
(PLANT SPECIFIC):
Byron SER (HUREG-0876) dated February 1982, page C-13, and Braidwood SER (NUREG-1002) dated 11/83, page C-1.
The NRC staff concluded plant could be operated safely before the ultimate resolution of USl A-12.
Final resolution incerporated fix on plants requesting CP after October 1983 - No fix required for Braidwood.
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REFERENCES:
Braidwood 1/2 p"t A-12 1
-1.
REQUIREMENT DOCUMENTS:
c l-TITLE NUDOCS NO.
DATE NUREG-0577 Rev. 1, "Futential 10/83
]
of Low Fracture Toughness and Lamellar Tearing in PWR Steam j
t Generator and Reactor Coolant 1
Pump Supports" 2.-
IMPLEMENTATION DOCUMENTS:
)
i F
TITLE NUDOCS NO.
DATE j
Byron SER (NUREG-0876) 02/82 i
E BraidwoodSER(NUREG-1002) 11/83 1
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3.
VERIFICA110W DOCUMENTS:
TITLE NUDOCS NO.
DATE N/A-e
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PLANT Braidwood 1/2 DOCKETN0(S).
_50 456/457 PRNECT PANAGER S. P. Sands TECHNICAL CONTACT D. Thatcher USI NO. A-17 TITLE Systems Interactions in Nuclear Power Plants MPA NO.
TAC N05.
ISSUES
SUMMARY
f GenericLetter(GL)89-18,datedSeptember6,1989 was sent to all power I
reactorlicenseesandconstitutestheresolutionoYUSIA-17. The generic letter did not require any licensee actions.
GL89-18hadtwoenclosureswhich(a)outlinedthebasesfortheresolutionof 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 l
review this information in other programs, such as the Individual Plant Examination (IPE)forSevereAccidentVulnerabilities.
Specifically, the staff expected that insights concerning water intrusion and flooding from internal i
sources, as described in the appendix to NUREG-1174, would be considered in the i
IPE program.- Also considered in the resolution of this USI was the expectation j
trat licensees would continue to review information on events at operating nuclear power plants in accordance with the requirements of TM1 Task Action Planitem1.C.5(NUREG-0737).
l 1MPLEMENTAT10N AND STATUS
SUMMARY
(PLANT SPECIFIC):
l Braidwood applied for CP in September 1973; therefore did not receive the 9/72 letter regarding flooding.
Licensee responded to GL 88-20 IPE Response submitted 10/27/89 Estimated implementation-10/93 l'
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-R ERENCES:
Braidwood 1/2 A-17 j
1.
REQUIREMENT DOCUMENTS:
TITLE NUDOCS N0.
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 i
for Resolution of US! 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" j
NUREG/CR-4470 " Survey and August 1986
)
Evaluation of. Vital Instrumentation and Control Power Supply Events" HRC Letters to Licensees 9/72 1
Informing Licensees of Staff Concerns Regarding Potential t
Failure of Non-Category I L
Equipment 4
2.
IMPLEMENTATION DOCUMENTS:
TITLE NUDOCS NO.
DATE l
IPE Response 8911080207 10/27/89 l
l 3.
VERIFICATION DOCUMENTS:
l l.
TITLE NUDOC NO.
DATE 1
e 8
1 a
]
1 PLANT. Braiwood 1/2 DOCKET N0(S), 50-456/457 PROJECT MANAGER S. P. Sands TECHNICAL CONTACT P. Shemanski USl NO. A-24 TITLE Qualification of Class IE Equipment
)
HPA NO.
TAC NOS.
ISSUES
SUMMARY
This USI was resolved in July 1981 with the publication of NUREG-0588, Revision i
1 " Interim Staff Position on Environmental-Qualification of Safety-Related l
ElectricalEquipment." Part I of the report is the original NUREG-0588 that I
was issued for comment; that report, in conjunction with the Division of 0 erating Reactor (DOR) Guidelines, was endorsed by a Comission Memorandum anJ i
0 der as the interim position on tnis 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 00R 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 i
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 cualification efforts i
completed as a result of Conmission Memorandum anc Order CL1-80-21 and also reflects different versions IEEE 323, dependent on the date of the construction j
permit Safety Evaluation Report (SER). Therefore, plant-specific requirements may vary in accordance with the rule, 1
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 j
found in the determination of compliance with 10 CFR 50.49. This was addressed i
by 72 SERs for operating plants issued shortly after publication of the rule 1
and subsequently in operating license reviews pursuant to Standard Review Plan Section 3.11. This was further addressed by the.first-round environmental j
qualification inspections conducted by the NRC.
IMPLEMENTATIONANDSTATUS
SUMMARY
(PLANTSPECIFICJ:
SER on IE Equip. Qualification (EQ) published in Byron SSERs #5, 6, and 7.
NUREG-0876 dated 10/84, 2/85 and 11/86 respectively and Braidwood SER NUREG-1002 supp. #2, dated 10/86, pages 3-16 and 17.
Implementation completed for Braidwood same as Byron. Licensee was in full compliance with 50.49 prior to license of both units on 11/86.
Unit 1 - 10/86 Unit 2 - 10/86
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REFERENCES:
Braidwood 1/2 r
A-24 i.. -RE001REMENT DOCUMENTS:
l
-TITLE NUDOCS NO.
DATE
. DOR " Guidelines for Evaluating Environmental Qualification of Class IE Electrical Equipment in Operating Reactors" NUREG-0588,." Interim Staff-Position on Environmental Qualification of Safety Related Electrical Equipment" 12/79 Commission Memorandum and Order, CL1-80-21, on DDR Guidelines and NUREG-0588 05/23/80
+
NUREG-0588, Revition 1-07/81 10CFR50.49(48FR-2730-2733) 01/21/83 Standard and Review Plan 3.11, Environmental Qualification of
-4 Mechanict.1 and Electrical
- Equipment 07/81 2.
IMPLEMENTATION DOCUMENTS:
TITLE NUDOCS N0.
DATE ByronSER(NUREG-0876)-
Supplement 5 10/84 Supplement 6 02/85 Supplement 7 31/86 Braidwood SER.(NUREG-1002) 10/86 Supplement 2 3.
VERIFICATION 00CUMENT3:
TITLE NUDOCS NO.
DATE B
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r, PLANT Braidwood'1/2 DOCKETN0(S). 50-456/457 PROJECT MANAGER- [. P. Sands TECHNICAL CONTACT Chu Liang i.
l
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USI NO. A-26' TITLE Reactor Vessel Pressure Transient Protection i
MPA NO.
TAC NOS.-
i
'I$$UES
SUMMARY
/
This USI was-resolved in September 1978 with the publication of NUREG-0224, i
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 g.
s plants were in startup or shutdown conditions. The issue affected all operating 1
and future plants,-and the staff established MPA B-04 for implementing the
-solution at operating PWRs.
J
- 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 f
vessel temperatures.
Since the reactor vessels have less toughness at lower L
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 vesse1~ damage, the NRC staff concluded that measures should be taken to minimize the number of future L
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..Thisgeneric.letterprovidesguidanceregardIngreviewofpressure-
. temperature limits and indicates that licensees may have to revise low-
-temperature-overpressure protection setpoints.
IMPLEMENTATION AMD STATUS
SUMMARY
(PLANT SPECIFIC):
Staff approved both Byron and Braidwood overpressure protection in the Byron
'SER (HUREG-0876) on 2/82,-pages 5-2 through 5-5 in section 5.2.2.2, and L
BraidwoodSER.(NUREG-1002)pageC-1.
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REFERENCES:
Braidwood 1/2 A-26 i
l 1.
REQUIREMENT. DOCUMENTS:
= TITLE NUDOCS NO.
DATE 1:
l:
" Reactor Vessel l
Pressure Transient Protection for PWRs."
9/78 m.
NRC Letters to Licensees
-Informing Licensees of Staff Concerns Regarding l
Overpressure Low-Temperature'
. Conditions in PWRs
-August 1976 i
1 I
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 i--
-Section 5.2 l
2.
IMPLEMENTATION' DOCUMENTS:
' TITLE NUDOCS NO.
DATE 1
.ByronSER(NUREG-0876) 02/82 BraidwoodSER(NUREG-1002) 11/83; 3.
VERIFICATION DOCUMENTS:
1 TITLE NUDOCS NO.
DATE l:
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= PLANT ~ Braidwood 1/2-DOCKETN0(S).
50-456/457 PROJECT MANAGER' $. P. Sands TECHNICAL CONTACT R. Jones
.USI.NO. Ai31 TITLE RHR Shutdown Requirements MPA NO.
TAC NOS.
ISSUES
SUMMARY
This US1 was resolved in May 197B with the publication of Standard Review Plan (SRP)Section5.4.7.
Only those plants expected to receive an operating
' license af ter January 1,1979 were affected by this resolution. The USI involved establishment of criteria 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 I would require full compliance for-Construction Permit (CP) or Preliminary Design Approval (PDA) applications
-which were docketed on or af ter January 1,1978. Class 2 required a partial
' implementation for all plants for which CP or PDA applir.ations 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 e
ard 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 effected 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):
Pages 5-22 through 5-26 of Byron SER (NUREG-0876, February 1982) and Braidwood SER (NUREG-1002, dated, 11/83) discuss Byron /Braidwood conformance to BTP RSB 5-3. ' The SER-stated that Byron /Braidwood met BTP RSB 5-1 with the exception of conducting a Natural Circulation test or demonstrating that the results of the Diablo Canyon test apply to Byron /Braidwood. Closecut of Natural Circulation issue was done by letter dated November 4, 1988 which stated that the results of the Diablo Canyon tests were applicable to Byron /Braidwood thus, no testing on Byron /Braidwood was required.
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IREFERENCES:
Braidwood 1/2 A-31=
t1.
' REQUIREMENT DOCUMENTS:
TITLE.
NUDOCS NO.
DATE-
^
4 NUREG-0800 " Standard Review Plan,"
5/78 l;
- SRP Section 5.4.7 NUREG-0606 " Unresolved Safety
- Issues Summary" y
Regulatory Guide 1.139, " Guidance for-Residual Heat Removal"
L J2.
IMPLEMENTATION 00CtMENTS:
l'
. TITLE NUDOCS NO.
DATE Byron SER (NUREG-0876) 02/82 L
BraidwoodSER.(NUREG-1002) 11/83 1:
Letter. National. Circ. Test 8811090159 11/4/88
'3..
VERIFICATION' DOCUMENTS:
TITLE NUDOCS NO.
DATE s
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- PLANT Braidwood 1/2-DOCKETN0(5).
50-456/457
-PROJECT MANAGER: S. P. Sands TECHNICAL CONTACT J. Wermiel USI NO, 'A-36 TITLE-Control of Heavy Loads, Phases I & II
TISSUES
SUMMARY
ThisUSI.wasresolvedinJuly1980withthepublicationofNUREG-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.
If these loads were to drop in certain locations in the plant, they may impact spent fuel, fuel in the core, or equipment that may be required to achieve safe L
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 recomend 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 f
. 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, andFebruary3,1981(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 4
Qualifications, training and specified conduct of operators 5.
Special lifting devices should satisfy the guidelines of ANSI N14.6.6.
6Property "ANSI code" (as page type) with input value "ANSI N14.6.6.</br></br>6" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process..
LiftinD 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 CHAA-70 Phase 11 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 Lalternatively, single-failure-proof cranes or load drop analyses in the spent c
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 11 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 11 review of heavy loads was not necessary and Phase II was considered completed.
l
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-While not a. requirement, NRC encouraged the implementation of any actions-j_
identified in Phase II regarding the handling of heavy _ loads that were considered appropriate.
I
- IMPLEMENTATION AND STATUS
SUMMARY
(PLANT SPECIFIC):
Phase I was found acceptable in Byron SER (NUREG-0876, SSER= #5) dated 10/84, page 9-1.-
~
Phase 2 covered in SSER #6 of Byron SER (NUREG-0876), dated 2/85, andLin BraidwoodSER-(NUREG-1002)supp.#1, dated 9/86,page9-1. The staff stated
-that based on-its review no further action by the licensee was required.
Initia1'responsetoGL-(12/22/80) was submitted 4/7/82. This issue was resolved prior to licensing.
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REFERENCES:
Braidwood 1/2 A-36
'1.
REOUIREMENT DOCUMENTS:
TITLE NUDOCS NO.
DATE Letter, Darrell G. Eisenhut, NRC, to all licensees, applicants for Ols and holders of cps transmitting T:
HUREG-0612 and staff positions
'12/22/80 Generic Letter 85-11 Hugh L.
Thompson, NRC,-to all licensees for Operating Reactors, " Completion of Phase 11 of ' Control of Heavy Loads at Nuclear Power Plants' NUREG-0612" 06/28/85 2.
IMPLEMENTATION-DOCUMENTS:
TITLE-NUDOCS NO.
DATE
-Byron SER.(NUREG 0876)
SSER #5 10/84 SSER #6 02/85 BraidwoodSER(NUREG-1002)
SSER #1 09/86 3.
VERIFICATION DOCUMENTS:
TITLE-NUDOCS NO.
DATE t
!z PLANT Braidwood 1/2 DOCKETN0(S), 50-456/457 PROJECT MANAGER S. P. Sands TECHNICAL CONTACT H. Ashar
.USI NO. A-40 TITLE Seismic Design criteria y
I MPA-NO.
TAC N05.
q-
- ISSUES
SUMMARY
l The staff has resolved USI A-40 as documented in NUREG/CR-5347, "Recommenda-tions for Resolution of Public Comments on USI A-40," issued in June 1989, and NUREG-1233, " Regulatory Analysis for USI A-40," issued in September 1989.
For plants not covered under the scope of USI A-46, " Seismic Qualification of' Equipment in Operating Plants," the staff concluded that tanks in plants that were subject to licensing review by the staff after 1984 had been reviewed.to L
current requirements and found acceptable.
For tanks in plants reviewed during L
1980-1984, the staff identified four plant sites (six units) that were not l
explicitly reviewed to current requirements. The four plants (Callaway 1/2, 3
Wolf Creek, Shearon Harris 1, and Watts Bar 1/2) are being handled on a plant-i specific basis.
USI A-40 originated in 1977. -The. basic objectives were (a) to study the Lseismic design criteria, (b) to quantify ~the conservatism associated with the criteria,and(c)torecommendmodificationstotheStandardReviewPlan(SRP)-if changes are justified. Lawrence Livermore National Laboratory (LLNL) completed l'
the study and published its findings in NUREG/CR-1161, " Recommended Revisions to USNRC - Seismic Design Criteria," dated May 1980. The report recomended specific changes to the Standard Review Plan (SRP).
HRC. staff reviewed the report and developed some other changes that would reflect the present state of
'l seismic design practices. The resulting SRP changes were issued for public comment in June 1988, and the final SRP dhanges are to be publishr.d in October 1989.
The major SRP changes consist of (a) clarification of development of site
- )
o L
specific spectra, (b) justification for use of single synthetic time-history by
)
pcwer spectral-density function, (c) location and reductions of input ground l
cmotionforsoilstructureinteraction,pnd(d)designofabove-groundvertical l
tanks. Except for item (d), these it m do not constitute any additional requirements for current licenses and applications, and thus, no backfitting is 1
0 being required for these items.
However, the revised provisions could be used h
for margin studies and reevaluations or individual plant examination for K
externalevents(IPEEE).
K The participant utilities in the Seismic Qualification Utility Group (SQUG)
. agreed to implement the changed criteria for flexible vertical tanks for their plants.
For the four: plants where this issue has to be resolved on an indi-vidual basis a 10 CFR 50.54(f) request-for-information letter has been sent to l
the affected utilities.
If the information received indicates that large above-ground vertical tanks do not meet the new criteria, plant-specific backfits will be considered.
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= IMPLEMENTATION AND-STATUS
SUMMARY
(PLANT, SPECIFIC):
I
, Byron SER (NUREG-0876) dated 2/82 on pages C-14-and 15, the' staff' concluded the seismic design basis and the design of Byron /Braidwood were acceptable.
Resolution of,A-40 should not effect the staff decision. The techniques being considered are essentially the same as those utilized in the Byron review.
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REFERENCES:
Braidwood-1/2 A-40
~
- 1.
REQUIREMENT DOCUMENTS:.
- TITLE NUDOCS NO.-
DATE Regulatory Analysis for NUREG-1233 Sept. 1989 USI A-40 i
~
Recommendations for Resolution NUREG/CR-5347-June 1989' of Public Comments on US1 A-40 Standard Review Plan NUREG-0800
.To be issued o
Sections 2.5.2, 3.7.1, 3.7.2,3.7.3(Revision 2)
Response of Seismic NUREG/CR-4776-Feb. 1987 Category I Tanks to Earthquake Excitation-
- Engineering Characteri-NUREG/CR-3805 Feb.-Aug. 1986 zation of Ground Motion, Vols. 3 4,5 Proceedings:of'the '
NUREG/CR-0054 June 1986 Workshop on. Soil-Structure Interaction Value Impact Assessment NUREG/CR-3480 Aug. 1984 forSeismicDeQgnCriteria 1
Seismic Hazard Analysis NUREG/CR-1582 Oct.1981 Application of Methodology, 1
Results.and. Sensitivity 1.:
Studies, Vol. 4 Recommended Revision to NUREG/CR-1161 May 1980 fluclear Regulatory Connission Seismic Design Criteria i
Power Spectral. Density Functions NUREG/CR-3509 June 1988 Compatible with NRC R.G. 1.60
-Response Spectra 1.
l 2.
IMPLEMENTATION DOCUMENTS:
y TITLE-NUDOCS NO.
DATE q
Request for Information Letters Docket Nos.
May 1989 L
to Owner's of Callaway 182, Wolf 483, 486, 482, Creek 1, Shearon Harris 1, Watts 400, 390, 391
- Bar 1&2 3.
VERIFICATION DOCUMENTS:
TITLE NUDOCS N0.
DATE
PLANTlBraidwood1/2-DOCKET N0(S). ~50-456/457 PROJECT MANAGER S. P. Sands TECHNICAL CONTACT A. Serkir-Lj USI NO. A-43 TITLE Containment Emergency Sump Performance l-MPA NO.
TAC N05.
- ISSUES
SUMMARY
- 19. USI NO. A-43 TITLE:
Containment Emergency Sump Performance
- The resolution of-this USI was presented to the Commission in October 1985 in l'
SECY-85-349. -NUREG-0897, Revision 1, " Containment Emergency Sump Performance,"
l l
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-1 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 residual 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-LOCAadverse i
l conditions resulting from potential vortex formation.and air ingestion and l
subsequent pump failure, (3) blockage of sump screens with LOCA generated
- insulation debris causing inadequate net positive suction head (NPSH) on pumps,.
l and (c) RHR and containment spray pumps inoperability due to possible air, debris, or particulate ingestion on pump seal and bearing systems.
l l
.This revised guidance applies only to future construction permits, preliminary l
design approvals,-final design-approvals, standardized designs, and applica-l tions for licenses to manufacture. The staff performed a regulatory analysis L
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 ll
- regulatory analysis does not support any new generic requirements for present licensees to perform debris assessments.
- IMPLEMENTATION AND STATUS
SUMMARY
(PLANT SPECIFIC):
Byron SER (NUREG-0876) pages C-15 and 16, dated 2/82 and Braidwood SER (NUREG-1002) dated 11/83, page C-1, addresses USI A-43.
Commitment was made by licensee to perform inplace tests to verify sump
. rec rcu al tion capability.
Item was closed on page 6-4 of Byron SSER #5 dated i
10/84 Resolved prior to licensing.
9 4
f f
REFERENCES:
Braidwood 1/2 A-43 1.
REQUIREMENT DOCUMENTS TITLE NUDOCS NO.
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 NO.
DATE ByronSER(NUREG-0876) 02/82 Byron SSER #5 10/84 BraidwoodSER(NUREG-1002) 11/83
' 3.
VERIFICATION DOCUMENTS:
TITLE NUDOCS NO.
DATE I
PLANT Braidwood:1/2 DOCKETN0(S)'.
50-456/457 PROJECT MANAGER S. P. Sands TECHNICAL CONTACT P. Gill l'SI NO.'
A-44 TITLE Station Blackout p
MPA NO.
TAC NOS. 68515, 68516 I
ISSUES
SUMMARY
ThisUSIwas'resolvedinJune1988withthepublicationof-anewrule(10 CFR 50.63).and Regulatory Guide 1.155.
f 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 comment'on 11 arch 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, 1
NUMARC-8700.. In order to comply with the A-44 resolution, licensees will be recuired to:.
.l maintain onsite emergency ac power supply reliability above a minimum' level L
. develop procedures and training for recovery from a station blackout determine the duration of a station blackout that the plant should be able l
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 l-Section 50.63(c)(1) of the rule required each licensee to submit a response l-
. 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.
IltPLEMENTATION AND STATUS
SUMMARY
(PLANT SPECIFIC):
Licensee responded on 4/17/89. Review is scheduled to be completed in flarch 31, 1990. Commitment was made to change procedures 1 year after NRC review is completed. Final implementation to be spring 1991.
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REFERENCES:
~
Braidwood 1/2 x
A-44 1.'
REQUIREMENT 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-from licensee.
8904240429 04/17/89 1
3.
VERIFICATION' DOCUMENTS:
TITLE NUDOCS NO.
DATE i
u
.i 4
I l
L L
- PLANT < Braidwood 1/2' DOCKETN0(S)'.
50-456/457 PROJECT MANAGER S. P.-Sands TECHNICAL CONTACT R. Jones USI NO. A-45 TITLE Shutdown Decay Heat Removal Requirements
- MPA NO.
TAC N05.
ISSUES
SUMMARY
4 US1 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 c
. 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 Connission 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 E
!~
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 i
L of achieving resolution of specific plant concerns associated with A-45.-
L L
IMPLEMENTATION AND STATUS
SUMMARY
(PLANT l SPECIFIC):
IPE-response submitted 10/27/89 IPE due 9/92 l-l
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REFERENCES:
'Braidwood 1/2-A-45 1.
REQUIREMENT' DOCUMENTS TITLE NUDOCS NO.
DATE Federal Register Notice "10 CFR a:
LPart-50, Shutdown Decay Heat Removal Requirements" NUREG/CR-5230 " Shutdown Decay Heat April 1989 Removal Analysis:
Plant Case Studies and Special Issues Summary Report" HUREG-1289 " Regulatory and Backfit 11/30/88-Analysis for the Resolution of.
USI A-45" SECY-88-260 " Shutdown Decay Hest 09/13/88 Removal Requirements 2.
IMPLEMENTATION DOCUMENTS:
TITLE NUDOCS NO.
DATE t
2 3.
VERIFICATION DOCUMENTS:
TITLE NUDOCS NO.
DATE-
.b f
1' PLANTc Briidwood 1/2-DOCKETN0(S).- 50-456/457
-PROJECT MANAGER: S. P. Sands TECHNICAL CONTACT J. Mauck I
- 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 d
LLetter(GL)88-19.
1
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'The generic letter states:
"The staff has concluded that all PWR plants should provide automatic steam generator overfill protection, all BWR plants p
should provide automatic reactor vessel overfill protection, and l-that plant procedures and technical specifications for all plants should. include provisions to verify periodically.the q
operability of the overfil1~ protection and to assure that automatic overfill protection is available to mitigate main feedwater overfeed events-during reactor power operation. Also, L
the. system design and setpoints should be selected with the-H objective of minimizing inadvertent trips of the main feedwater system during plant startup, normal' operation, and protection system surveillance. The Technical Specifications recomenda-1 tions are consistent with the criteria and the risk considera-tions of the Comission. Interim Policy Statement on Technical Specification Improvement. 'In addition,'the staff recommends e
H 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."
i Also, page 2 of the generic letter provides for additional actions'for CE and B&W plants.' The generic letter provides arpifying guidance for licensees.
~The generic letter requires that licensees provide NRC with their schedule and comitments within 180 days of'the letter's date. The implementation schedule
- for' actions on which comitments are made should be prior to startup af ter 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):
1 Response due 3/90 e
t
REFERENCES:
lBraidwood:1/2 A-47 1.
REQUIREMENT DOCUMENTS u-
~
TITLE NUDOCS NO.
DATE 4
Generic Letter 89-19 09/20/89
" Request for Action Related to Resolution of USI-A-47" NUREG-1217 " Evaluation of Safety June 1989 Implications of Control Systems in LWR Nuclear Power Plants" NUREG-1218 " Regulatory Analysis July 1989 for Resolution of USI A-47" 2i IMPLEMENTATION DOCUMENTS:
TITLE NUDOCS NO.
DATE 3.
VERIFICATION DOCUMENTS:
TITLE NUDOCS NO.
DATE f
1 PLANT Braidwood-1/2 DOCKETN0($)'. 50-456/457 PROJECT MANAGER S. P. Sands-TECHNICAL COMTACT B. Elliott-USI NO.
A. TITLE ~ Pressurized Thermal Shock MPA NO.
TAC NOS. 64030, 64057 ISSUES
SUMMARY
Thefinal-rule (10CFR50.61)'onpressurizedthermalshock(PTS)wasapproved-by the Connission in July 1985. Regulatory Guide 1.154, " Format and Content of Plant-Specific' Pressurized Thermal Shock Safety Analysis Reports for PWRs,"
was later published'in February 1987. Thus, this issue was resolved and new requirements were established, applicable to PWRs 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 facility.
Neutron irradiation of reactor pressure vessel weld and plate materials decreases the fracture toughness of the materials. The fracture toughness sensitivity to radiation-intoced 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 thfeaten vessel integrity.-
Severe pressurized' overcooling events are' improbable since they require multiple failures and improper operator performance. However, certain J
-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.
It4PLEMENTATION AND STATUS
SUMMARY
(PLANT SPECIFIC):
Licensee' responded to'10 CFR 50.61 on 1/17/86. SER still under review.
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REFERENCES:
Braidwood 1/2 A,
1.-
P.E0llIREMENT DOCUMENTS:
TITLE NUDOCS 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-?88, " Proposed Pressurized Thermal Shock Pule" 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 and Its Impact on Plant Operations" 7/12/88 2.
'1HPLEMEPTATION DOCUMENTS:
TITLE NUDOCS NO.
DATE Letter from licensee 8601230286 1/17/86
-3.
VERIFICATION DOCUMENTS:
TITLE NUDOCS NO.
DATE I
-:n cf ENCLOSURE 2 6'
- Page No._.
02/02/90
. LISilN6 0F INCOMPLETE USI DATA
'FOR IW UT FROM PROJECT MNA6ERS ISSUE ' ISSUE DESCRIPT!VE MME
.!MPLEMENT 1MPLEMENT LICENSEE ComENT
~ STAFF COMENT.
NUMBER DATE STATUS-88 PLANT EMET BRAIDN000 1 i
A 01 WATERNAMER
//
NC A ASYMMETRIC BLONDON LOADS ON
//
NC 3
REACTORPRIMARYCn0LANTSYSTEMS A-03 NESilN6 HOUSE STEAM BENERATOR TUBE //
NC 1W 0 DNLY INTE6 Riff A-64 CE STEM GENERATOR TU8E INTEGRITY //
N/A-CE PLANTS ONLY A.B4W STEAM SENERATOR TUBE _
/ -/. N/A B&W PLANTS ONLY-
'INTEGRiiY l
- A-06 ' MARK I SNORT-TERM PRO 6 RAM
/. /
N/A
- 1 BNR Olli A-07 MRK I LON6-TERM PROGRM
//
N/A MK 1 BUR ONLY A-08 MARK 11 CONTA! MENT POOL DYNAMIC
//
N/A
- !! BNR ONLY LOADS - LONG-TERM PROGRAM L
!A 09' ATNS ~ ~
03/31/91 1 A-10 ~BNR FEEDWATER N0ZILE CRACKIN6
/ /- N/A BUR ONLY g
Al!. REACTOR VESSEL MATERIALS-
//
NC l'
1006HNESS l.
A-12 FRACTURE TOU6HNESS OF STEAM
//
N/A
- CP AFTER 83 ONLY '
6ENERAIDR AND REACTOR COOLANT PUMP SUPPORTS :
-l l,
LA-17 SYSTEMSINTERACTION
/- /
NC NO REQUIREMENTS t
E' A-24 -QUALIFICATION OF CLASS IE
//
NC
- SAFETY-RELATEDEOUIPMENT l
A-26 REAC10R VESSEL PRESSURE TRANSIENT //
NC LTOPS 0 OL
('
PROTECil0N A-31 RHR SHUTDOWN REQUIREMENTS
//
NC LICENSIN6 SER-A-36 CONTROL OF HEAVY LOADS NEAR SPENT //
-NC SL-85-11 ENDED FUEL A 39 DETERMINTION OF SAFETY RELIEF
//
N/A BNR ONLY VALVE POOL DYNAMIC LOADS MO TEMPERATURE LIMITS A-40 SE!SMIC DESIGN CRITERIA -
//
HC SHORT-TERM PROGRAM A-42 P!PE CRACKS IN B0!LIN6 NATER
//
N/A BNR DNLY REACTORS.
l A-43 CONTAINMENT EMER6ENCY SUMP
//
NC INFO ONLY PERFORMNCE A-44 STAi!0N BLACK 0UT 03/31/91 1 PROCEDURES SER 3/31/90 A-45 SHUTDOWN DECAY HEAT REMOVAL
//
NC IPE SUBSUMED BY SEVERE ACC REDVIREMENTS
~A-46 SEISMIC 00ALIFICAT!DN OF
//
N/A OLD PLANTS ONLY EQUIPMENT !N OPERATING PLANTS A-47' SAFETY IMFLlCATIONS OF CONTROL 03/31/90 E NEW REQUIREMENTS SYSTEMS
' A 48 HYDR 06EN CONTROL MEASURES AND
//
N/A N/A DRY CONTAIN EFFECTS OF HYDROGEN BURNS ON
' SAFETY EDU1PMENT A-49 PRE!SURl!ED THERMAL SHOCK 04/30/90 1 NA111NS NRC REVIEN RESPONSED 1/17/06 1
@p V
1 J
ENCLOSURE 2:
1
,y PageNo.1 2'
+ ' 02/02/90?
LISilN6 0F INCOMPLETE U$l DATA-
. FOR INPUT FROM PROJECT MANAGERS -
,e
~
-ISSUE ISSUE DESCRIPTIVE NAME
. IMPLEMENT IMPLEMENT LICENSEE COMENT STAFF Co MENT NUMER1 DATE
- STATUS
- 1
)
, $$ PLMi NAME: BRAIDWOOD 2 4-01 = WATER HA MER.~
//
NC
--A-02' ASYMETRIC BLOWDON LDADS ON
./ /
E
- REAC10R PRIMARY COOLANT SYSTEMS A-03 NESilN6 HOUSE STEAM SENERATOR TUBE / -/
E INFODNLY.
'INTE6RITY.
A 04'-
CE STEAM 6ENERATOR TUDE INTE6R11Y //
N/A CE PLMTS ONLY
'A-05 B6W STEAM SENERATOR TUDE
/=/
N/A B6W PLANTS ONLY INTE6R!iY L
A-06 MARK 1SHORT-TERMPR06AAM:
/ /
N/A E ! BWR DEY.
l
~ MARK 11 CONTAl MENT POOL DYNAMIC'
'l /
-A 07-MARK 1 LON6-TERM PROGRAM
'/ /
N/A MK1BWRONLY L
.A-08 D
LOADS - LON6 TERM PR06 RAM l1
~
!A 09 ATWS 03/31/90 1-
.A-10 BNRFEEDWATERN0ZZLECRACKIN6
//
N/A IWR DNLY I
A-11 ~ REACTOR VESSEL MATERIALS
//
E l:
TOU6HNESS l
. A-12 FRACTURE TOU6HNESS OF STEAM
//
N/A CP AFTElt 83 DNLY lf
.6ENERATOR AND REACTOR COOLANT 1
l PUMPSUPPORTS-
'A-17 SYSTEMS INTERACT 10N
//
E NOREQUIREMENTS A-24' QUAllFICAi!ONOFCLASS1E
/ /
NC SAFETY-RELATEDEQUIPMENT A-26 REACIDR VESSEL PRES 8URE TRANS!ENT //
NC LTOPS I DL PROTECTION
.A-31' RHR SHUTDO N REQUIREMENTS
//
NC LICENSIN6 SER A-36. CONTROL OF HEAVY LOADS NEAR SPENT: //
E 6L 85-11 ENDED
. FUEL l <.
=A DETERMINAT10N OF SAFETY RELIEF
//
N/A DNR DNLY VALVE POOL DYNAMIC LOADS AND 1
TEMPERATURELIMITS L
A-40 'SE!SMIC DES!6N CRITERIA -
//
E SHORT-TERM PROGRAM A-42 PIPE CRACKS IN B0! LING WATER -
//
N/A INR ONLY REACTORS
.A-43 CONTAl MENT EMER6ENCY SUMP
//
E INFO DNLY
~
PERFORMANCE
-A STATION BLACK 0UT 03/31/91 I PROCEDURES SER 3/31/90 A-45' SHUTD0 H DECAY HEAT REMOVAL
//
NC SUBSUMED BY SEVERE ACC E
REQUIREMENTS fA-46 SE!SMIC OVALIFICATION OF
/ /
N/A OLD PLANTS ONLY EQUIPMENTINDPERATINGPLANTS y"
.A-47 SAFETY IMPLICATIONS OF CONTROL 03/31/90 E NEW REQUIREMENTS SYSTEMS A-48 HYDR 06EN CONTROL MEASURES AND
//
N/A N/A DRY CONTA!N EFFEdTSOFHYDR06ENBURNSON
'SAFETYEDUlFMENT A-49 PRESSURl!ED THERMAL SHOCK 04/30/90 I WAITING NRC REY!EW RESPONDED 1/17/06
y
+
EllCLOSURE 3-i
'i
_ C;mm:nwealth Edison 72 West Adams Street Chicago HMo's-Aco<ess Rep y to Pes: Ohce Boa 77 I
s Chcago Heos 60690 0767 November 28, 1989 f
Dr. Thomas E. Hurley, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555
Subject:
Byron. Station Units 1 and 2 i
Braidwood Station Units 1 and 2 Response to Generic Letter 89-21 EC Docket Nos. 50-454/455 and 50-456/457
Reference:
Generic letter 89 Request for Information Concerning Status o_f Implementation of a
Unresolved Safety Issue (USI) Requirements
_1
Dear Dr. Hurley:
Generic Letter 89-21 was issued as a part of the NRC's continuing-4 effort to validate staff ~ understanding regarding implementation of unresolved safety issues (USIs). An-important aspect of this effort is to ensure that the licensee and the NRC agree on the status'of USI resolution implementation:
.]
at each facility, This letter presents Commonwealth Edison (Edison's)-
J response to Generic Letter 89-21, for the Byron and Braidwood Stations.
Byron and Braidwood stations have reviewed-the guidance provided in-the Generic Letter.
The attached of:USIs tables (Enclosure I from the generic letter)-have been updated to reflect the. current status of USI implementation.
The generic letter's gu_ide for-' updating USI statut was utilized for this process. Accordingly, Attachments "A" and "B" of this 1 9tter provide the status of the efforts at Byron and Braidwood Stations, respectively.
-Pleae direct any questions regarding this response to this office.
Very truly yours, (Y
k S.C. Hunsader Nuclear Licensing Administrator A8+9270024 s911Ps PDR ADOCK 05000454 qf[
P puu L
'/ sci:0351T:28 1.
Attachments: A-Byron Station B-Braidwood Station p
l lr
( (I '
cc:
A.B. Davis-RIII Resident Inspectors-B(/BW l -
L. 01shan-NRR I
iv.-
~
A_TTACM M T "A_"
nVRM mTIM EPCLOSURE 3-UIWtE50t.VED SAFETY ISSUES FOR l#ffCN A FIP.AL TECHNICAL RE50tWTi If5I/MPA MtiMBER TITLE
' REF. DOCl#IENT APPtICA8It_fTY STAT 1f5/DATE*
REfWWIRS.
^!
A-1 Water Hanner SECY 84-119 Al1-C It'/81 RYRom sER' MREG-09?7. Rev. I pg-13-24 M REC-0993. Rev. 1.
MREG-0131 f tem 1
I.A.7.3 SRP revisions A-2/
Asyuusetric RIcenfoun IIllREG-0609 PWR c et 3/87 avaan sEn MPA D-10 toads on Reactor Priquery CL 814-04 GDC-4 U2 11/86
'sECTIon 3.9.2.4' Coolant Systems A-3 Westinghouse. Steam NtREG-0444 W-PMt C "3 Generator Tube Integrfty.
SECY M-97
'5ECV 88-772 s
GL 85-02 (flo requirements)
A-4 CE Steam Generator Tube MREG-0844. SECY 86-97 CE-PWR Integrfty SECY 88-272 I
GL 85-02 (IIs requirements).
A-5 RAW Steam Generator HUREC-0844. SECY M -97 flatf-PWP Tube Integrity 5fCY R8-272 GL 85-02 (flo 8tequirements) 1-6 Mark I Containment IIUREG-040ft Part I-BMt Short-Terie Program
- C - COMPLETE
.e-NC - NO CIVINGFS NECESSARY NA - NOT APPLICA8tE I - INCOMPLETE F - FVAt 1947fuc ArTrnee ormemen -
~~
~
N 7-ENCLOSURE 31 b
q USI/MPA MemfR TITLE -
REF. DOCtIMFNT APPLICA8ttITV-STATIPS/94TE*
REftRRRS A-7/
Mark ;I Long-Term mmEG-0661 Nrk I-rim
'na' I)-01 Program
' MmEG-0661 Suppl. I
-GL 79-57 A-8 Mark II Contefament.
NUREG-0808 Mert II-8Mt nA Pool Dynamic, loads NUREG-0487. Suppl. 1/2 MfREG-0902 SPP 6.2.1.1C GOC 16 A-9 Anticipated Transfents MREG-0460. Vol. 4 All I al-3/9e scutset.E WIthout Scram 10 CFR 50.62 n2-10/90 svantTTre 2-15-89 A-10/
8WR Feedwater Herzle NUREG-OSl9 BWp nA HPA B-?S Cracking Letter from DG Eisenhut dated 11/13/80 GL Al-El i
A-Il Reactor Vessel Material
'NUREG-0744. Rev. I All c st-10/s4 armen ssan M Toughness 10 CFR 50.60/-
.U2-11/86 rc. 5-3 82-26 g
A-12
. Fracture Toughness of'
_NUREG-0577. Rev. I PNP c 2/s2 avaan sEn-Steam Generator and' SRP Revision c-13 i
Reactor Coolant Pump
. 5.3.4 Supperts i
A-17 Systems Interactions Ltr: DeYoun All n 10/92 ct. ss-2e licensees g to 9/72 Irr ansponsa i
NL9Ec-3174. NIREG-svantTTEo 10/27/s96 1229. NtmEG/CR-3922 MREG/CR-4761. NUPEG/
CR-4470. Gt. 89-18 ~
(No requ9rements) 8 A-74/
Ovalf fication of Class -
MIREG-0588. Rev. I All c at lo/s4 synon sea secrson MPA B-60 IE Safety-Related SRP 3.11.
Equipment
-le CFR 50.49 U2 11/86 3.11-synoussra #5 Gl. 82-09. Gl. 84-24 ssaa #6 St. 85-15 ssen er
~,
___.C_..
id _ _....-
~3~
. ENCLOSURE 3
.a USI/MPA MWMER TITEE REF. IIOCIMERT '
APPtIC48tlITY-STATUS / BATE
- RF9Wptf(S i
A-76/
Reactor Vessel Pressure
-IIN! Letters to~
PINT MPA R-04 Transfent Protectfon-
.Lfcensees 9/76 C UI-lO/s4 avnow SER n2-11/86 sEcTron 5.2.2.2 t
MmEG-0224 Marn-0371 SRP 5.2 GL 88,
A-3)
Residual Heat Removal MIREG-0606 All Ots After C 2/82 minou seR secTven Shutdown Respufrements RG 1.113 01/79.
1I/ss 5.4.3.
RG l.139 SRP 5.4.7 (NATURAL CinCUIATIoW) i A-36/
Contre) of Heavy loads IIUREG-0612 All C-PnAsE I
]
C-10 Near Spent Feel SRP 9.I.5 Ut-10/s4 synon sta
~;
~
C-15 GL 81-07. GL 83-42 52-11/86 sEctron 9.1.5 GL 85-11
-s.vnen ssen 45 l'
E
't t
c sE H
.st a ssna M 12/22/80
- I I
A-39 Determination of SWV MmEG-0902 OMt WA Pool Dynamic loads
- MmEGs-0763.0783.0802 and Pressure Transtents HUREG-0661
~
SPP 6.2.I.I.C A-40 Sefsafc Design SRP Revisfons, ifMREG/
All C 2/s2 synes sen
'Criterla CR-4776. RUREG/CR-9054 C-14 4 15 flUREG/CR-3480. IIURES/
CR-1582; IIURES/CR-Il61 letMfG-1233. ' flullEG-4716 MORES /CR-3005 MCEG/CR-5347 -
A-47/
Pfpe Cracks in Sofffag MMEG-0313. Rev.1 ONR
. NA MPA B-05 Water Reactors MOREG-03I3. Rev. 2 l
GL SI-03. GL 88-01
-+
4 EftCLOSURE 3 ~
IISI/ PPA M MIER TITLE REF. HOCtMENT; APPLICA811.ITY ^
STATUS /DATE*
RffWtRKS A-43 Contatament Emergency M *EG-0510, All
'c et lo/s4 "synon sEm'c-ts Senp Performance m MEG-0869. Rev.~1 e2 11/s6i A c-16s arnon MMEG-0897. R.G.I.87 ssna #5, rg.'6-4' (Rev. 6). SRP 6.2.2l GL 85,
-No Requfrements A-44 5tatten Blackout RG 1.155 All I (Implement 1' mespense
~
MREG-1932 4
year' af ter nec review) b=It ted NUREG-Il09 4/17/s9 10 CFR 50.63
/
A-45 Shutdown Decay. Heat SECY DR-260 All.
E 10/92 Analysis remaine -
Removal Requirements MREG-1289
' to be.ccepteted.
NUREG/C#-5230
-ct,as-20/tre~
SECY 88-260 me.p e
- 6. tete.
-(No requirements) 10/27/s9.
A-46 Setssic Qualiffcation NUREG-1930 All I
sqec provided of Equipment in NUREG-1211/
generic taptement Operating Plants GL 87-02. GL 87-03
~
atica. Rev. C on 4 6/03/ss; saa re 1"~
A-47 Safety Impitcatfon
.NUREG-1217. NUREG-All E 3/90 ceived 7/29/ss. s' of Control Systems 121R' new, 1 isheitted
'GL 29-19 12/s8; Rev. I isoder NRC A-48 Hydrogen Control.
.10 CFR 50.44 All. except NA review..
Measures ae.d Effects SECY 89-l??-
PWRs with of NyA ;;;.. Burns large dry on Safety Equfpuent
. containments A-49
~Pressurired Thermal l PGs 1.154. 1.99-PNR c w1 1/s6 s1r e a sE n I a/ ss Shock SFCY 82-465 I U2
'For U2 InIC review SECT R3-288 SECY 81-687
'in progrese of~
-I/17/s6' 6 1etel.
10 CF#'50.61/
GL 88-11-'
j.
~,
~ -
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- ii
_7 ENCLOSURE 3 USI/MPA IIl8MRfB -
TIY1E Rtr. fiOCUPFWT APPt ICASit ITY STATW5/94TE*
REpingR$
A-7/
Merk I long-Ters MtfREG-0661 b rk I-WWR i
nA.
D-01 Program l
N18 REG-0661 Suppl. I Gl. 79-57 A-8 h rk II Confafonent IIUREG-ORGR Mark IT-SuR nA Pool Dynamic Leeds NUREG-0487, Suppl. 1/2 NUREG-0902 SPP 6.2.I.1C GDC 16 A-9 Anticipated Transients NUREG-0460. Vol. 4 All st-3/91 schedule
- 6. steed i
Wfthout Scram 10 CFR 50.62 v2-3/90 2-15-89 m
A-10/
3NR Feedweter IIorrie NUREG-0619 HNW
!*A Braidwood sta pg.
MPA P-75 Cracking Letter from DG Eisenhet c-I dated 11/13/80 GL Al-Il i
t A-Il Reacter Vessel Meterfel NUREG-9744, Sev. I All c 9/96 staid d saa T
,M ;s In CFR 50.60/
s=pp. I pg. 5-4 a?-26 I
A-12 Tractere Toughness of IIUREG-0577. Rev. I PNP na t
Steam Generator and 3RP newe ten s
Neactor Coolant PInip 5.3.4 SePyer's i
A-17 Systems Interactfees Ltr: BeVeeng to All E 3e/93 CL 88-2e -
Ifceasees - 9/72 ra me. m IIUREC-1174, sflWWEG-sebettted 10/27/s9
{
1229. IINRE#fCR-3922, IIUREC/CR-4Mi, IIUREG/
l CR-4470, Gl. 89-13 l
(8le regeframents) 1 A-74/
Cualf fication of Class Hl811EG-9530. Rev. I AII c se/s6 staid d saa PPA 8-60 IE Safety-Belated SIIP 3.11 j-Egefreent IS CFR 56.49 s.pp. 2 pg. 3-16,17 R 82-89. EI. M -N EL 35-15 i
I..
3-EMCLOSURE 3 UST/MPA MIMRER TITLE REF. DottmENT APPtfC40ftiTY STATU5/SRTE*
RremRet5 A-76/
Reetter Vessel Pressere DMt Letters to PtIR c II/s3 syron sta pg. 3-4 F#A R-04 Transtent Protectfen Licensees 9/76 staid d saa M MEG-0224 Pn C-3 msRFC-n371 SPP 5.7 Gl. 88-11 A-38 Residual fleet Removst IllmEG-0606 All Ots After c 11/83 syren sem pg. 3-2 Sluttelsesa Requfrements RC 1.113 01/79.
II/88 (**t#f*I 23.24 8t*Id***d RG j.!39 circulatlon) sea pg. c-I I
SNP 5.4.7 A-36/
Control of fleevy Leeds smREG.0612 All c 9/s6 sesidweed sea f
C-10 Itear Spent Feel SNP 9.I.5 s.pp.I pg. 9-t C-15 GL 81-07. GL 83-47 GL H5-11 Letter,from OG EIsenIwt deted 32/22/80 l
A-39 Deterefnetten of SRT MWEG-0907 Otm mA Peel Dynomfc Leeds IIllREGs-0763.0783.pW82
{
and Pressure Transfeats NUREG-0661 SPP 6.2.1.1.C A-40 Sefsofc Design SRP Revistems. HUREG/"
Agy c 11/s3 syre. sen pg.
Criteria ER-4776. IlWREG/CR-0954 C-34.35 ImREG/CR-348n. pgNES/
sesid d sta CR-1587. Nt!!fG/CR-II69 PR C I M8RFG-1233. IIUREG-4716 M8 REG /CR-3005 M8RE4/CR-5347 IIURES/CR-3509 I
A-47/
Pipe Cracks In Belling lumEG-03I3. Rev. I SIR NA MPA 8-05 lister lleacters HUREC-03I3. Ilev. 2 1
GI. 81-03. R 88-01 1-F'-
a e 2 e_,._-
m
+-.u--u. -
u,,
2,___._,_
ws ae__,,a-v-
__m,,
_ms_m,aa-_ mas 3
.m.
we
+- -
,1 o
-4 EMCLOSURE 3 175I/ PPA ORIMBER TITLE
_REF. f8nctBEIIY APPtICASILITY STATW5/DATE*
- EIMRR$
A-43 Con'afament Emergency MWEG-0510 All O
Suny Performance 8888 REG-0969. Rev.1 C H /83 syren su c-Is.
MmEG-0897. R.G. I.87 C-I' *
"7'*" su (Rev. 8). SRP 6.2.2
- 5 PK 6-'
"'*8d"'*8 E 85-22 sea c-s No Neguirements A-44 Statten Blackout RG 1.155 All I (1mplement 1 ye r me.p
.e e.seted MIREG-1932 etter unc reetew) 4/I7/89 MfREG-IIO9 10 CFR 50.63 A-45 Shefdown Decay fleet SECY Mi-268 All E le/93 A lys t e reen te.
Itemeval Requirements IIUREG-1289 to be completed.
IIUREG/CR-5230 cl. ss-re/ter me.p e
SECT 88-250
- h. steed se/27/s9 (No requiremed s) e A-46 Sefsmic Ovalificatten seusTG-Ic30 Aii a
sque prosaea me.e,se of Egulpment la sgUREG-1211/
8=ple=eatetson aav o Operating Plants GL 37 32, et gy-g3 6/3/ssa su reces,ed 7/29/es. sque A-47 Safety Implicatten NUREG-IZlT. HUREG-All E 3/M Be I beltted 12/88 of Centrol Systems 1218
""d"
""C '"
GL 39-19 A-48 Hydrogen Control IS CFR 50.44 All. except
~~
Measures and Effects SECY 89-122 PWRs wfth of ",1 ;g, turns large dry on Safety Egufpment contafanents A-49 Presserfred Thermal RGs 1.154, 1.99 pgM Awastseg nec reese.
Shock SFCY 82-465 of **h=8tted SECT R3-298 d*t'd /'i/86 SECT 81-687' l
le CFR 50.61/
El. as-Il
.