ML20033E061
| ML20033E061 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 02/21/1990 |
| From: | Siegel B 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-46, REF-GTECI-A-47, REF-GTECI-EL, REF-GTECI-SC, REF-GTECI-SY, TASK-A-09, TASK-A-44, TASK-A-46, TASK-A-47, TASK-A-9, TASK-OR GL-89-19, NUDOCS 9003060283 | |
| Download: ML20033E061 (3) | |
Text
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UNITED STATES
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g NUCLEAR REGULATORY COMMISSION g
l W ASHINGTON, D. C. 20666
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February 21, 1990 Docket Nos. 50-237 and 50-249 MEMORANDUM FOR:
File FROM:
Byron L. Siegel, Project Manager Project Directorate 111-2 Division of Rebctor Projects - 111, IV, V and Special Projects
SUBJECT:
STATUS OF IMPLEMENTATION OF UNRESOLVED SAFETY ISSUES AT DRESDEN NUCLEAR POWER STATION UNITS 2 AND 3 The current implementation status of unresolved safety issues (USIs) at the Dresden Nuclear Power Station is set forth in the enclosures to this memorandum. contains a copy of the information provided by the licensee in its response to Generic letter 89-19. contains a status summary for each USl applicable to Dresden.
This status summary is based upon the licensee's response to the Generic Letter, discussions with the licensee, and my review of available NRC records and information. is a cooy of the staff's data base printout for Dresden.
It reflects the staff's assessment of USI implementation for all 27 USIs.
It is basea on review of the licensee's response to Generic Letter 89-21,, and evaluation by project managers, the USl team, Gnd NRR technical staff.
Four of the USI's discussed below are incomplete:
Anticipated Transients Without Scram (A-09); Station Blackout (A-44); Seismic Qualification of Equipment in Operating Plants (A-46); and Safety implications of Control Systems in LWR Nuclear Power Plants (A-47).
Anticipated Transients Without Scram (A-09):
Licensee has installed all modifications to meet the rule. Technical Specifications associated with ATWS modifications were submitted September 29, 1989 ard are under staff review.
A diversity issue associated with ARI and RPT analog trip units is under appeal to the NRC by the BWROG.
Station Blackout (A-44): The licensee's submittal has been reviewed by the staff there is a difference of opinion regarding the plant's acceptable blackout duration capability.
In a December 20, 1989 meeting the licensee proposed changes to its submittal to address staff concerns.
The licensee's proposed changes for meeting the rule will be formally
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1eM v 293-v4 (3sjg) m
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i File 2-February 21, 1990 submitted by the end of February 1990 andmill 'be reviewed by.the staff.
The staff's SE should be issued by June 1990. -The licensee's.implemen-tation date as expected to be not later June 1992 based on the requirement
'in the rule.
Seismic Qualifications of Equipment in Operating Plants (A-46):- The staff has prepared a supplemental SE for the Generic Implementation Procedure, Revision 1 which is under review by CRGR. An additional supplement is scheduled for June 1990 and overall closeout of implementation-is projected for 1993.. Dresden plant walkdowns are scheduled to be completed by the i
middle of 1991.
Safety Implications of Control Systems in LWR Nuclear Power Plants t
(A-47): The response to this USI is not due until March 20, 1990.
The licensee's schedule for response to issues A-09, A-44, A-46 and A-47 is either dependent upon staff actions or regulations and is therefore acceptable.
Appropriate Technical Review Branches have reviewed the USI summary and memorandum.
/s/
Af
.i w Byron L.-Siegel, Project Manager Project Directorate III-2 Division of Reactor Projects - III, IV, V and Special Projects
Enclosure:
As stated-
=cc w/ enclosures:
K. Ecceleston.
Enclosure:
As' stated cc w/ enclosures:
K. Ecceleston L
DISTRIBUTION Docket File PDIII-2 r/f
'J. Craig L. Luther B. Siegel p
R. Wessman PD13 (J4 P II I:LA PDIII-2:Pt - PDIII-2:PD PD :
'llu er BSiegel:t1 JCraig RWe man 1/
0 1/2(/90 1/w /90 1/
0 DOCUMENT NAME:
USI DRESDEN I
4
4 1
'l File 2
Februar.v 21, 1990 i
submitted by the end of February 1990 and will be reviewed by the staff.
The staff's SE should be issued by June 1990.
The licensee's implemen-
_tation date as expected to be not later June 1992 based on the requirement i
in the rule.
Seismic Qualifications of Equipment in Operating Plants (A-46): The staff has prepared a supplemental SE for the Generic Implementation Procedure, Revision 1 which is under_ review by CRGR. An additional supplement is I
j scheduled.for June 1990 and.overall closeout of implementation is-projected for 1993. 'Dresden plant walkdowns are scheduled to be_ completed by the middle of 1991.
Safety Implications of Control Systems in. LWR Nuclear Power Plants (A-47): The response to this USI is not due until March 20, 1990.
The licensee's_ schedule for response to issues A-09, A-44, A-46 and A-47 is either dependent upon staff actions or regulations and is therefore acceptable.
, Appropriate Technical Review Branches have reviewed the USI summary and memorandum.
/tt B ron L. Sie el, Project Manager Project Directorate III-2 Division of Reactor Projects - III, IV, V and Special Projects
Enclosure:
'As stated 1
- cc w/ enclosures:
K. Ecceleston L
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'[f, __ )!Comm:nrealth Edison 72 West Acams Street. Chroo. Ithnois -
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' A00'e55 Repy to Post OfhCe 50s 767
<y Chicago. Ithno4 60690 0767 November 29,.1989 Dr. Thomas E, Murley, Director.
Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, CC' 20555 L
Subject:
Dresden Nuclear Power Station Units 2 and 3 Response to Generic Letter 89-21 Implementation Status of USI Requirement NRC Docket Nos. 50-237 and 50-249
Reference:
Letter from JG Partlow to Licensees dated October 19, 1989
.(received October 30, 1989)
Dear Dr. Hurley:
Enclosed is the information requested by the subject Generic Letter (which was transmitted by the referenced letter) concerning the implementation status of Unresol'ved Safety Issue (USI) requirements relevant to Dresden Units 2 and 3.
5 The status for each relevant USI has been determined to the best of.
our ability in the allotted 30 day period since receipt of the referenced letter. Commonwealth Edison believes the enclosed information accurately reflects the Dresden USI=1mplementation status at this time.
However, should additional information become available which differs from that enclosed, the revisions will be promptly communicated to the NRR Project Manager for Dresden, Please contact his. office further information be required.
Very truly yours, A
k ohn A. S11ady-Nuclear Licensing Administrator
(
0418T Attachments cc:
A.B. Davis - Regional Administrator B.L. Siegel - Project Manager, NRR S.G. DuPont - Senior Resident Inspector, Dresden Aol A i i
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-QNRESOLVED'SAFETYISSUESFORHHICHAFINALTECNhlCALRESOLUTIONHASBEENACHIEVED-'
- QRESDEM_ UNITS 2 AND 3; L -
USI/MPA-NDMBER TITLE-STATUS /DATE.
~ REMARKS Highreactorwaterhevelfeedpump.tripwas
- A-1.
- Hater Hammer
- C ~5/85:
installed in the early 1970's. The~ training upgrade ~ requirements of NUREG 0737,1 Item I.A.2.3 were incorporated into the training-program prior.to accreditation.,Dresden was-accredited in May 1985.
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Asymmetric Blowdown M/A HPA 0-10 Loads on Reactor Primary Coolant Systems A-3 He.tinghouse Steam N/A Generator Tube Integrity A-4 CE Steam Generator-Tube N/A l
Integrity A-5 B&W Steam Generator N/A
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Tube Integrity C - COMPLETE NC - NO CHANGES NECESSARY NA - NOT APPLICABLE I - INCOMPLETE E - EVALUATING ACTIONS' REQUIRED w
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USI/MPA LNUMBER TITLE
$[AIUS/DATE REMARKS E
A-6 Mark I Containment =
C Superceded by USI A-7 Short-Term Program A-7/
Mark I Long-Tern I
-3/90 NRC Letter Zwolinksi to Farrar dated 9/18/85 D-01 Program
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completed the' staff review of the issue.
Implementation was reviewed by the NRC.-
Inspection Report 237/87019; 249/87018 - No deficiencies were noted.
The majority of the modifications for long-term program were completed between 9/77 and 8/84.
Two remaining partial mods M12-2(3)-86-41C and M12-2(3) -
03C and D deal with small bore piping and'are scheduled for D3Ril.
A-8 Mark II Containment N/A Pool Dynamic Loads A-9 Anticipated Transients C
2/87 Modifications for ARI, RPT, and SLCS were Without Scram completed 2/87. SER issued on 11/8/88.
Tech Spec change'for SLCS and RPT was submitted 9/29/89. -Per 9/29/89 submittal no Tech. Spec.
change ' f<or ARI is required.
A-10/
11/83 Modifications were completed U-2, S/81 and MPA B-25 Cracking U-3, 5/82. SER issued 11/1/83.
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USI/MPA NUMBEB TITLE STATUS /DATE REMARKS A-11 Reactor Vessel Material I 6/90 Tech. Spec. aumendment to incorporate radiation Toughness effects into PT curves'were' submitted percR.G.
1.99.
Based upon'the relatively low end-of-11fe fluence and the small~espected drop in end-of-life charpy upper-shelf energy from RG 1.99, General Electric and CECO believe that there are no technical' concerns with-older BHR
~ vessels dropping below the 50 ft.-Ibn limit.
However, CECO plans to approach the BNROG to establish a generic program to develop' baseline data for initial unirradiated charpy upper shelf values for older BWR vessels and provide the associated documentation. Based upon recent discussions with GE, CECO expects a report to be issued in approximately 6 months.
l A-12 Fracture Toughness of N/A l
Steam Generator and Reactor Coolant Pump Supports
+
A-17 Systems Interactions C
10/89 Event reviews are conducted in accordance with
-NUREG 0737 Ites I.C.5 and are controlled by DAP's 2-11 and 2-12.
CECO's response to GL 88-20 committed to an IPE program. Water intrusion and flooding from internal sources are part of the IPE.
USI/MPA NUMBER TITLE STATUS /DATE REMARKS A-24/
. Qualification of Class C
8/86 SER issued 2/11/86. SER granted an extension HPA B-60
,1E Safety-Related for U-3.
Mods completed for U-2 4/85 and U-3 Equipment 8/86.
Followup inspections 237/86013; 249/86015 and 237/89010; 249/89009. During the 1989 Inspection, the inspector had a concern with EQ electrical enclosures that contain taped spilces or terminal blocks may not have weep holes. Consequently,'Dresden is currently performing walkdowns of all EQ circuits enclosures to address this concern.
A-26/
Reactor Vessel Pressure N/A HPA B-04 Transient Protection A-31 Residual Heat Removal N/A Shutdown Requirements A-36/
7/83 TER issued 7/11/83 found phase I C-10, Near Spent Fuel acceptable. Superceded by Phase II.
C-15 Phase I 4
Phase II C
6/85 Draft TER issued 6/84. As documented in GL 85-11, Dresden 2 and 3 were part of MRC pliot review program of Phase II. GL 85-11 concluded that based on the results of the Phase II pliot program there are not residual concerns of sufficient significance to demand further generic action.
4-39 Determination of SRV N/A-See A-7
' Pool Dynamic Loads and Pressure Transients
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'USI/MPA
. HUBEER TITLE SIAIDS/DATE REMARKS A-40 Seismic Design I
Halting final NRR acceptance'of EPRI -
Criteria Selsmicity Owner's Group position on Items 1, 2, and 3 of the proposed SRP. Item 4 dealing with the design of flexible vertical tanks will be addressed as part of USI A-46.
A-42/
Pipe Cracks In Bolling I
12/90 Leakag? monitoring Tech. Spec. revision is in HPA B-05 Water Reactors Off-site Review.
Expected approval date.
11/30/89, and project submittal to NRC in early 1990.
Tech. Spec.~ requirements for ISI program will be removed from the Tech. Specs.'and included in the ISI. program following NRR review of GL 88-01 response and SER issuance.
Revised ISI program is expected to be submitted to NRR for approval by the end of 1990.
A-43 Containment Emergency N/A Sump Performance A-44 Station Blackout I
The 4/17/89 CECO response. proposed the coping approach in combination with some hardware changes. During a working meeting held on 10/4/89, there appears to be a diference of opinion between the staff and CECO regarding plant acceptable blackout duration capability.
Discussions are on-going.
A-45 Shutdown Decay Heat C
USI A-45 will be addressed by IPE as-implemented-Removal Requirements by GL 88-20.
Dresden IPE is expected to be completed by 4/92.-
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NUMBEB TITLE LSIAIUS/DATE REMARKS A-46 Seismic Qualification I
6/91*
Waiting-staff approval'of SQUG generic of Equipment in inspection plan and issuance of the final Operating Plants SER. -SQUG walkdowns will be conducted following staff approval of GIP. Based upon an' early 1990. approval, the walkdowns have been tentatively scheduled for completion in old 1991..
A-47 Safety Impilcation E
Response to GL 89-19 is due 3/90. May require of Control Systems-a Tech. Spec. revision.
A-48 Hydrogen Control I
Haiting for NRR to issue the staff position.
Measures.and Effects Discussions are ongoing.
of Hydrogen Burns on Safety Equipment A-49 Pressurized Thermal N/A Shock
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Enclosure:
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PLANT-- Dresden Unit Nos.-2 and-3:
DOCKETN0(S). 50-237 and 50-249 PROJECT MANAGER _ Byron Siegel TECHNICAL CONTACT A. Serkir i
q.
-USI NO. A-1 TITLE. Water Hamer MPA NO.
N/A TAC NOS. N/A
-ISSUES
SUMMARY
This Unresolved Safety Is'"* (USI) was resolved in March 1984, with the publication of NUREG-092'
- Evaluation of Water Hamer in Nuclear Power Plants
- Technical Findings Relevant to Unresolved Safety Issue A-1."
Also on March i
15, 1984, the EDO sent the Commissioners SECY 84-119 titled, " Resolution of Unresolved Safety Issue A-1, Water Hammer."
-In SECY 84-119, the' staff concluded that the frequency and severity of water hammer occurrences had been.significantly reduced through (a) incorporation of design features such as keep-full systems, vacuum breakers, J-tubes, void detection systems, and improved venting procedures; (b) proper design of feed-water valves and contro1' systems; and (c) increased operator awareness and training; Therefore, the resolution of USI A-1 did not involve-any hardware or design changes on existing plants.
It did involve Standard Review Plan (SRP) changes _(forward fits) and a comprehensive set of guidelines and criteria to evaluate and upgrade utility training programs (per TMI Task Action Plan Item I.A.2.3).
In addition, the assumption was made that for BWRs with isolation
[
condensers (ICs) a reactor-vessel high water-level feedwater pump trip was in place or being installed. This was necessary because calculated values had postulated an IC failure by water hammer that opened a direct pathway to the environment.
IMPLEMENTATION AND STATUS
SUMMARY
DRESDEN:
Dresden has an IC. A high reactor water level feedwater pump trip was l-
' installed in the early 1970's.
TMI Item I.A.2.3 was incorporated into the j
training programs prior to accreditation. Dresden was accredited.in May 1985.
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REFERENCES:
Dresden A-1
' i 1.-
REQUIREMENT-DOCUMENTS:
TITLE-
~ NUDOCS NO.
DATE Letter from Denton to Utilities, 8403150310 03/05/84~
" Notice of Issuance and Availability HUREG-0927 Rev. 1, Safety Issue-A-1" 2 '.
IMPLEMEllTAT10tl DOCUMENTS:
l 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 Hamer" SRP Sections:
3.9.3, 3.9.4,=
5.4.6, 5.4.7, 6.3,.9.2.1, 9.2.2, 10.3, and 10.4.7 SECY-84-119, " Resolution 03/15/84 of Unresolved Safety A-1, llater ' Harmer" 3.
VERIFICATI0tl DOCUMENTS:
TITLE NUDOCS N0.
DATE O
m
p f
4 q
e
- PLANT _Dresden.
DOCKETN0(S).50-237/249
- PROJECT MANAGER-Byron-L. S_ieael TECHNICAL CONTACT J. Kudrick
- USI NO.,A-6
, TITLE Mark I containment Short Term Program-g'
_ MPA NO.
TAC NOS.
1SSUESSUMMAR6 This-USI was. resolved in December 1977 with the publication of.NUREG-0408,
" Mark I' Containment Short-Term Program Safety Evaluation Report,"
The objectives of the Mark I short-term program werer (a) to examine the containment systen of.each BWR facility with a Mark I containment design to:
verify _ that it would maintain its integrity and functional capability when subjected to the'most probable hydrodynamic loads induced by a postulated design-basis LOCA, and (b)..to verify that licensed Mark I BWR facilities could continue to operate safely, without undue risk to the public health and safety until such time-as a methodical, comprehensive long-term program is conducted.
The NRC staff used a safety factor of at leas't two to failure for the weakest structural or mechanical component in the Mark I containment system in judging that containment integrity and' functions would be assured under most probable
. design-basis LOCA-induced hydrodynamic loads.
- As indicated in NUREG-0408, the staff required full implementation of the calculation of the hydrodynamic loads and structural analysis as an interim measure until_ complete implementation of the long-term program had been achieved.
In NUREG-0408 the staff concluded that the cbjectives of the Short-s
~ Term Program had been satisfied, thus documenting the basis for resolving this safety issue. This issue is considered complete for all affected BWRs.
- IMPLEMENTATION AND STATUS
SUMMARY
DRESDEN:
In conjunction with the Mark I Short Term Program the licensee performed a plant unique analysis for Dresden which demonstrated a factor of safety of at least two for the weakest element in the suppression chamber support system and attached piping. The NRC staff's evaluation of this analysis is contained in NUREG-0408, which recommended implementation of a drywell to suppression i
chamber differential pressure.
The licensee submitted a license amendment
- application that was approved by the staff (Amendment 37 for Dresden's 2 and Amendment 35 for Dresden 3) that requires the maintenance of a 1.00 psid drywell - suppression chamber differential pressure. This assures the
- integrity of the suppression chamber when subjected to post - LOCA suppression
-pool hydrodynamic forces. The staff anticipated that the corrective actions taken by the licensee in response to the Mark I Short Term Program would be implemented as an interim measure. However, the licensee, as part of-the corrective actions for the Mark I Long Term Programs, retained this pressure differential permanently.
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REFERENCES:
Dresden 1 -
A-6 J
i li REOUIREMENT DOCUMENTS:
TITLE NUDOCS NO.
DATE HUREG-0408, " Mark ~I Containment 12/77 l
Short Term Program Safety'
'l
- Evaluation Report"^(See Table I-?-
foreletters to BWR licensees requesting. action) 1 n
. 2 '.
IMPLEMENTATION DOCUMENTS:
TITLE NUDOCS NO.
DATE "Dresden Nuclear Generating 08/76 Plant Units'2 & 3 Short i
Terms Program Plant Unique Torus Support and Attached.
Piping Analysis" HUTECH Report COM-01-040, t
3..
- VERIFICATION DOCUMENTS.
' TITLE-NUDOCS N0.
DATE i
+
9
w We PLANT Dresden Unit Nos. 2 and'3 DOCKET 110(S).
50-237 and 50-249 s
' PROJECT MANAGER Pyron Siegel TECHNICAL CONTACT
-J. Kudrick USI;NO. -A-7'
. TITLE M_ ark 1 Long Term Program
- HPA NO.
TAC NOS. 07935/07936
-ISSUES
SUMMARY
i This USI'was resolved in August 1982 with the publication of Supplemed 1 to
- NUREG-0661, ? Safety Evaluation Report, Mark 1 Containment Long-Term Program" and Standard Review Plan Section 6.2.1.1.C.
For operating BWRs, MPA 0 01 was established for implementation purposes.
The focus of this USI was the suppression pool hydrodynamic loads, associated with a postulated LOCA, which had not explicitly been included in the original Mark I, containment design. The issue was identified during large-scale testing of a Mark III containment design. The staff addressed this issue.in NUREG-0661, g
l published in July 1980, and in Supplement 1 to ilVREG-0661, published in August 1982.
The objective of the long-term program (LTP) was to establish the design-basis loads that'are. appropriate for the anticipated life of each Mark I BWR facility and to restore the originally intended design-safety margins for each Mark.I 4
containment system. The principal thrust of'the LTP was the development of generic methods for defining suppression pool-hydrodynamic loadings and the associated. structural assessment techniques for the Mark I configuration. On the basis of experimental and analytical programs conducted by the Mark-I Owners Group, it was determined that the hydrodynamic load definition pro-cedures, with some modifications defined in NUREG-0661, provided a conservative estimate of these loading conditions. Thus, the requirements associated with this-USI'were concerned with the structural assessment of Mark I containments and related structures to the hydrodynamic loads defined by the staff in the LTP.
In January 1981, the staff issued " Orders For Modification of License and Grant of Extension of Exemptions" to each licensee of a Mark I plant. The orders required the licensees to assess the suppressinn-pool hydrodynamic loads in accordance with General Electric documents and NUREG-0661 on a defined L
schedule.
For some plants, the implementation schedule was extended by a L
subsequent order.
IMPLEMENTATION AND STATUS
SUMMARY
DRESDEN:
Staff's SER-dated September 18, 1985 to licensee concluded that either the modifications made were in accordance with NUREG-0061 criteria or that the L
identified deviations were acceptable.
Licensee completed implementation of the modifications associated with staff's SER between September 1977 and August 1984. However, after the program was completed the licensee identified:
1) areas where the support configurations used in the analysis were not consistent with the as built configurations; 2) areas where the long term modifications were not consistent with the design drawings; and 3) errors in the analyses.
The licensee evaluated the as built configuration for operability and determined i.
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REFERENCES:
Dresden iJ A-7 o
it to be functional and has completed corrective actions except for some repairs related to Dresden 2 which are scheduled for completion by the er:I of o
December 1990 and two modifications related to small bore piping for Dresden 3 1
which are scheduled for completion by March 1990.. The licensee was cited for violation of Appendix B, Criteria 3 (Design Control)'and Criteria 5 failure to--
follow procedures and drawings. These are addressed in Inspection Reports:
237/87006-010 and 02B and 249/87011-01B and-.02B.
1.
REQUIREMENT DOCUMENTS:
q TITLE NUDOCS NO.
DATE NUREG-0661, " Safety-Evaluatinn
[
Report, Mark I Containment Long
~
Term Program" 07/80
'NUREG-0661, Supplement 1 08/82 Orders for Modification to License for Applicable Licensees'.
1981 2.
IMPLEMENTATION DOCUMENTS:
TITLE NUDOCS NO.
DATE Letter from J.A. Zwolinski 8509230613 9/18/85 NRC, to D.L. Farrar (Ceco)
' dated September 18, 1985 transmitting SER on Plant
-Unique Analysis Report pursuant to NUREG-0661 and its supplement for the Mark I Long Term Program t
Dresden Nuclear Power Station 8307070176 05/83 Units 2 and 3 Plant Unique Analysis Report, Volumes 1-7 Prepared for Comonwealth Edison Company by NUTECH Engineers, Inc.
3.-
- VERIFICATION DOCUMENTS
TITLE NUDOCS NO.
DATE Inspection Reports 237/87019 8707150244 07/87
'249/87018 Inspection Reports 237/87006 8707310045 07/87 l-249/87011
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PLANT. Dresden Unit Nos. 2 and 3-DOCKET N0(S). i50-237.and 50-249 a
PROJECT MANAGER, Byron Siegel TECHNICAL CONTACT J. Mauck
~
.USI NO. A-9 TITLE ATWS per 10 CFR 50.62 MPA NO. A-20 TAC NOS.
59089/59090 74895/74896
= ISSUES
SUMMARY
4
-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 anticipeted -transients and to mitigate the consequences of an anticipated transient without scram (ATWS) event.
The-rule includes the following design-related requirements:
50.62(C)(1),
diverse and independent auyiliary feedWater initiation and turbine trip for all
.PWRs; 50.62(C)(2), diverse. scram systems for CE and B&W reactors; 50.62(C)(3) alternate rod injection (ARI) for BWRs: 50.62(C)(4); standby liquid control system (SLCS) for BWRs; and 50.62(C)(5), automatic trip of recirculation pumps under' conditions indicative of an ATHS for BWRs.
Information requirements and an implementation schedule are also specified.
IMPLEMENTATION AND STATUS
SUMMARY
ORESDEN:
A Safety Evaluation on-compliance of the ARI and RPT design with 10 CFR 50.62 was issued on November 8, 1988.
No Safety Evaluation on SLCS, however,
' inspections performed in accordance 'Sith TI-2500/20 reported in Inspection
' Reports 237/88011 and 89011 and-249/88014 and 89010 determined there were no violations.
-By-letter dated September 29, 1989 the licensee submitted Technical Specification amendments to reflect modifications to the SLCS and the addition of a RPT to comply with the requirements of 10 CFR 50.62. The staff should issue the amendments for both Dresden 2 and 3.in early 1990.
No Technical Specifications for ARI is currently required.
There is'a-diversity issue associated with the ARI and RPT analog trip units under appeal to the NRC by the BWROG which is pending resolution-before implementation can be considered conplete.
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REFERENCES:
Dresden A-9 1.c REQUIRit1ENT DOCUMENTS:
TITLE NUDOCS t.'D.
DATE-
.NUREG-0460,.and Supplements, 03/80
" Anticipated Transients Without Scram for Light Water Reactors" Federal Register flotice 49FR26045(10CFR50.62) 06/26/84 P.-
IMPLEMEllTATION DOCUMENTS:
TITLE NUDOCS tl0.
DATE Henry Bliss '(CECO) gel (NRC)'to Letter: Byron Sie 8811150025 11/8/88
" Compliance with ATWS Rule 10 CFR 50.62 Relating to'Alterate_ Rod Inspection (ARI)andReactor'CoolantRecirculation-n Pump Trip-(RPT) Systems for Dresden Units g
2 and-3 and Quad Cities Units 1 and 2".
A Letter:
"Dresden fluclear Power 8910040200 9/29/89 Station Units 2 and 3 Proposed fAnticipated Transients Without..
Scram Related Technical Specification
. Change s.
3.
-VERIFICATION DOCUMENTS:
. TITLE NUDOCS NO.
DATF.
-I
-Inspection Report 237/88011 8808050222 07/88 l
- Inspection Report 249/88014 8808290094 08/88
('
Inspection Report 237/89011 8806270430 06/89 j
Inspection Report 249/89010 8806270430 06/89-
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E.
s, PLANT; Dresden Units Nos. 2 and 3 DOCKET N0(S). 50-237 and 50-249 i
- PROJECT MANAGER ; Byron Siegel' TECHNICAL CONTACT K. Wichman USl NO. A-10
- TITLE BWR Feedwater Nozzle Cracking MPA NO. 'B-25 TAC NOS.- 08552/08483
- ISSUES
SUMMARY
This' issue was resolved in November 1980 with the publication of NUREG-0619,-
"BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking." MPA B-25 was established by NRC's Division of Licensing for implementation
~,
purposes.
Inspections of operating EWRs conducted up to April-1978 revealed cracks in the feedwater nozzles of 20 reactor vessels.
It was determined _ that cracking was due to high-cycle fatigue caused by fluctuations in water temperature within 1
the vessel in the: nozzle region.
s
- By letter. dated-November 13, 1900 Darrell G. E:senhut provided licensees with
- a copy of NUREG-0619. The letter stated that NUREG-0619 provided the'resolu-i
- tion of the staff's generic technical activity USl A-10, which resulted from the. inservice discovery of cracking in-feedwater nozzles and control rod drive return line nozzles.
NUREG-0619 describes the technical issues, General J
- Electric'and staff studies and analyses, and the staff's positions and require-ments._ Licensees were-required to. respond, pursuant to 10 CFR 50.54(f), that they would meet implementation dates indicated in NUREG-0619.
Generic Letter 81-11 was subsequently issued to provide technical clarification to the November 13,1980 letter, to clarify that it had been sent to PWR licensees for:information only, and that no response was required from PWR licensees.
IMPLEMENTATION AND STATUS
SUMMARY
DRESDEN:
NRC letter dated November 1, 1983 provided closure of A-10.
The staff concluded that if the leak rate is less.than 1.0 gpm in the triple-sleeve-sparger, the Dresden units can operate without rerouting the.RWCU. Also the_present air-operated low-flow control on the 0 - 20% power range, accomplished through-the globe-type bypass valve is acceptable. The licensee has stated-in its response-to GL 89-21 that modifications were completed for
. Dresden 2 in May 1981 and for Dresden 3 in May 1982.
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L It REFERENCESi Dresden A-10 1
L 1.s REQUIREMENT DOCUMENTS:
t-. c L
TITLE.
-NUDOCS NO.
DATE Letter from D. Eisenhut-i 7h, transmitting NUREG-0619,.
t Control-Rod Drive Return E
Line Nozzle Cracking,"
L-resolution of.A-10 to D
licensees 11/13/60 Generic Letter 81-11 "BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle
- Cracking.(NUREG-0619)"-
02/20/81 1
m:
2.
I!1PLEMEt'TATION DOCUMENTS:
TITLE.
NUDOCS N0; DATE 1
Letter " Implementation of 8108060227 7/20/81 Unresolved Safety Issue A-10,-
BWR Nozzle Cracking
Letter " Implementation of 8311040059 11/1/83
-NUREG-0619, BWR Feedwater and CR0 Nozzle Cracking."
l Letter " Response:to Generic 8912060061 11/29/89
- Letter 89-21,-Implementation Status of USI Requirements" Letter " Implementation of 8102250396-02/23/81 NUREG 0619" 3.-
VERIFICATION DOCUMENTS:
TITLE.
NUDOCS NO.
DATE i
I
\\
h
]
b'.
DOCKET N0(S).
50-237 and 50-249
[
FLANT Dresden Units 2 and 3 i
PP00ECT PANAGER Byron L. Siegel TECHNICAL CONTACT B. Elliott-6
(;
USI NO. A-11 TITLE Reactor Vessel Materials Toughness MPA NO.
TAC NOS.
ISSUES
SUMMARY
- This USI was resolved in October 1982 with the publication of NUREG-07AA,
" Pressure Vessel Material Fracture Toughness.". NUREG-0744 was issued by
- Generic Letter 82-26 and provided enly a methodology to satisfy the require-ments of 10 CFR Part 50, Appendix G.
No licensce response to Generic Letter
- 82-26 was required.
Becauseoftheremote.hossibilitythatnuclearreactorpressurevessels de:igned to the ASME B iler end 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 cperating conditions typical of current operating plants, reactor vessel fracture touchness properties provide adequate margins of safety against vessel failure; however, as plants accumulate more and more service time, neutron irradiation reduces 1
the materi_al fracture toughness and initial safety margins.
-Appendix G to 10 CFR-Part 50 requires that the Charpy upper shelf energy
'throughout the life of the vessel-be no less than 50 ft-lb unless it is demonstrated that lower values will provide margins of safety against failure-equivalent to those provided by Appendix G of the ASME code.
USI A-11 was cinitiated 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.
-HUREG-0744-provides a method for evaluating reactor vessel materials when their Charpy upper. shelf energy is predicted to fall below 50 f t-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 SUWARY DRESDEN:
The-licensee has submitted a TS amendment which incorporates the irradiation effects into the P-T curves as per the guidance contained in the current Lversion of RG 1.99.
The licensee's submittal is currently under staff review, c
s
'however, the licensee has stated that the E0L upper shelf energy for the most limiting vessel material (weld material for Dresden 3 and plate material for
' Dresden 2) meets the Appendix G 50 ft-lbm requirement. Therefore, this USI is not applicable to Dresden Unit Hos. 2 and 3 since, to date, the licensee does
-not predict either unit to fall below the 50 ft-lbm limit value.
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REFERENCES:
(., '
q Dresden z.:
A,
e 1.;
_ REQUIREMENT DOCUMENTS:.
TITLE NUDOCS NO.
DATE NUREG-0744, Revision 1, " Pressure 10/82 Yessel Material Fracture Toughness" F
Generic Letter 82-26, " Pressure Vessel Material Fracture Toughness" 11/12/82 i
u..
L
((
2.
IMPLEMENTATION DOCUMENTS:
TITLE NUDOCS NO.
DATE' p u.;
LetterR.Stols(CECO)toThomas 8911060202 10/23/89 Murley (NRC) " Application for
^ Amendment to Facility Operating Licenses DPR-19, OPR-25, DPR-29,
- and DPR-30; f-I f
3.
VERIFICATION DOCUMENTS:
TITLE NUDOCS NO.-
DATE
?
t 3
1 I
P' s
t
)
+
f
, PLANT Dresden, Units-2 and-3
=0OCKET N0(S).
50-237 and 50-249-PROJECT MANAGER Byron L.-Siegel TECHNICAL CONTACT. D. i
.:her USI NO. A-17 TITLE Systems Interactions in Nuclear Power Plants MPA NO.
TAC NOS.
ISSUES
SUMMARY
Generic Letter (GL) 89-18,- dated September 6,1989, was sent to all power
~
- reactor licensees and constitutes the resolution of USI A-17.
The generic letter.did not require any licensee actions.
GL 89-18 had two enclosures which (a) outlined the bases for the resolution of USI A-17, and (b) provided five general lessons learned from the review of the overall systems-interaction issue. The. staff anticipated that licensees would review this information in other programs, such as the Individual Plant Examination (IPE)forSevereAccidentVulnerabilities.
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 th_e resolution of this USI was the expectation that licensees would continue to review information on events at operating.
nuclear power plants in accordance with the requirements of TMI Task Action
. Plan Item I.C.5 (NUREG-0737).
IMPLEMENTATION AND STATUS
SUMMARY
DRESDEN:
- Licensee has stated that event reviews are conducted in accordance with NUREG 0737 Item I.C.5 and are controlled by administrative procedures.
In addition the effects of water intrusion and flooding from internal sources will be L
considered as'part of the level 2 IPE (10/89 response to GL 88-?O).
The licensee stated it has no record of receiving a_ letter from the staff in 1972 that would have requested Dresden-to determine whether the failure of any 4
non-Category I equipment could result in a condition such as floodina or release of chemicals, that might adversely affect the performance of safety related
- equipment required for safe shutdown or limit the consequences of an accident.
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p
REFERENCES:
Dresden A-17 1.
PEOUIREMENT DOCUMENTSi TITLE-NUDOCS NO.
DATE Generic letter 89-18 09/06/89 NUREG-1174 " Evaluation of May=1989'
. Systems Interactions in Nuclear
-Power Plants" k'
PUREG-1229 "Re9ulatory Analysis August 1909 for Resolution cf USI A-17" HUREG/CR-3922 " Survey and January 1985' Evaluation of Systen 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.
4 Instrumentation and Control Power-Supply Events"
'NRC. Letters to Licensees.
9/7?'
-i Informing-Licensees of Staff c '
Concerns Regarding Potential i
Failure of Non-Category I.
j Equipment
?
'2.
IMPLEMENTATION C0CUMENTS:
TITLE NUDOCS NO.
DATE Letter " Response to Generic 8911080207 10/27/89 i
Letter 88-20 and Supplement 1"-
l 1
3.-
VERIFICATION DOCUMENTS:
TITLE NUDOC NO.
DATE 1
l
{
L 1
i
t
^
_. PLANT Dresden Units-Nos. 2 and 3 DOCKET N0(S).
50-237 and 50-249 s
s PROJECT t% NAGER : Byron L. Sieael TECHNICAL CONTACT P. Shemanski f
-USI NO. A-24 TITLE Qualification of Class 1E Equipment MPA NO. B-60 TAC _NOS.
52519/52520 ISSUES = SUtiMARY: _
This USI was resolved in July 1981 with the publication of NUREG-0568, Revision
-1, " Interim Staff Position on Environmental Qualification of Safety-Related l
' Electrical Equipment." Part I of the report is the original HUREG-0588 that-was issued for comment; that report, in conjunction with the Division of-
.l 10perating_ Reactor.(DOR) Guidelines, was endorsed by a Ccemission Memorandum and
- 0rder 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 sel" -
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 (SER). Therefore, plant-specific requirements may vary in accordance with the rule.
In summary,.the resolution of-A-24 is embodied in 10 CFR 50.49. A measure of 1
whether each licensee has implemented the resolution of A-24 may therefore be found in the-determination of complience with 10'CFR 50.49.
This was addressed
-i by 72 SERs for operatirg 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 i
= qualification inspections conducted by the NRC.
IMPLEMENTATION AND STATUS
SUMMARY
DRESDEN:
- The.NRC's SER transmitted February 12, 1986 found the licensee's EQ program to be-in compliance with 10 CFR 50.49. The SER granted an extension for Dresden-3. The modifications were completed in April 1985 for Dresden 2 and June 1986 for Dresden 3.
During the 1989 follow-up inspection, the inspector identified a concern with EQ electrical enclosures that contain taped splices or terminal blocks may not have weep holes.
Consequcntly, Dresden is currently performing walkdowns of all E0 circuits enclosures to address this enneern.
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REFERENCES:
Dresden s
A-24 1.
RE,QUIREMENT DOCUMENTS:
TITLE NUDOCS NO.
DATE r
' DOR " Guidelines.for Evaluating o
Environmental Qualification of u'
Class IE Electrical Equipment in Operating Reactors" NURE0-0588, " Interim Staff. Position on Environmental-Qualification of Safety Related Electrical Equipment" 12/79 Commissicn Memorandum'and Order, CLI-80-21,'on D0R Guidelines and NUREG-0588-05/23/80 NUREG-0588, Revision 1 07/81 10CFR50.49(48FR2730-2733) 01/21/83 Standard and Review Plan 3.11, Environmental Qualification of Mechanical and Electrical Equipment-07/81 2;-
If1PLEMENTAT10NDOCUMENTS:
TITLE NUDOCS N0.
DATE Letter " Safety Evaluations.
8602280489-02/12/86 Addressing the Environmental Oualification of. Electrical Equipment Important to Safety Letter " Response to Generic 8912060061 11/29/89 89-21, Implementation Status of USI Requirements" 3.
VERIFICATION DOCUMENTS:
. TITLE NUDOCS NO.
DATE Inspection Report-237/86013 8906200179.,
06/89
. Inspection Report 249/88009 8906200179 06/89 Inspection Report 237/86013 8609120295 09/86 Inspection Report 249/86015 8609120295 09/86
L 6
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PLANT Dresden Units 2'and-3' DOCKETN0(S).
50-?37 and 50-249 PROJECT MANAGER Byron L; Siegel' TECHNICAL-CONTACT 'J.
Wermiel
.USI NO.
A ' TITLE Control of Heavy Loads, Phases 1 & 11 MPA NO.
C-10, C-15 TAC NOS.
07986 and 52223 ISSUES
SUMMARY
This USI was resolved in' July 1980 with the publication of NUREG-0612.)" Control of Heavy Loads atLNuclear 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 1 and II, respectively, of the resolution of this issue at operating-plants..
In nuclear power plants, heevy 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 t-shutdown and continue decay heat removal.
USl 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 definition of,of heavy loads. The guidelines proposed in NUREG-0612 include safe handling 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 varicus 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 r.cnth response) and Phase II (nine month response).
Phase 1 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 crates 4.
Qualifications, training and.specified conduct of operators 5.
Special lifting devices shculd 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..
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 Il 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 'ouilding (BWR), other areas and the specific guidelines for single-failure-nroof handling systems.
As stated in Generic Letter 85-11. " Completion of Phase 11 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 1 and a Phase 11 report.
Based on the im of NUREG-0612 (provements in heavy loads handling obtained from implementation 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-l Dresden A-36 While not a requirement, NRC encouraged the implementation of any actions identified in Phase 11 regarding-the handling of heavy loads that were considered appropriate.
IMPLEMENTATION AND STATUS
SUMMARY
ORESDEN:
An SER was transmitted to the licensee on July 11, 1983 stating that the response to Phase I was. acceptable and that the guidelines contained'in NUREG-0612.were satisfied. The licensee has stated that the requirements of NUREG-0612 were completed by January 1985. A draft TER for Phase II was transmitted to the licensee on June 28, 1984 As documented in GL 85-11, Dresden 2 and 3 were part of HRC pilot review program of Phase 11. GL 85-11 concluded that based on the results of the Phase il pilot program there are not residual concerns of sufficient significance to
--demand further generic action.
REFERENCES:
1.
REOUIREMENT DOCUMENTS:
TITLE NUDOCS NO.
DATE Letter, Darrell G. Eisenhut, NRC, to all licensees, applicants for OLs and holders of cps transmitting NUREG-0612 and staff positions 12/22/80 Generic Letter 85-11, Hugh L.
Thompson,_NRC,_to all licensees for Operating Reactors, "Cempletion of Phase II of ' Control of Heavy Loads at Nuclear Power Plants' NUREG-0612" 06/28/85
~ 2.
It1PLEMENTATION DOCUMENTS:
' TITLE N_UDOCS NO._
j DATE l
Letter "NUREG-0612 8307130303 07/11/83 Phase I of Heavy Loads of Nuclear Plants" l
Letter " Control of Heavy Loads -
8407030247 06/28/84 Phase 11 i;
Phase II, NUREG-0612" Letter J. Able (Ceco) to 8105190445 05/15/81 D. Cisenhut (NRC) " Control of Heavy Loads NUREG-0612" Letter E. Swartz (Ceco) to 8109300163 09/22/81 i
D. Eisenhut-(NRC) " Response to llUREG-0612 " Control of Heavy Loads at Nuclear Power Plants"
DbjIl Akl
~~
~
L:m Letter E. Swartz (Ceco) 8112170178 12/11/81 to D. Eisenhut (NRC)." Control
. of Heavy Loads at flutlear Power Plants"
' ~
D.'Eisenhut(NRC)(Ceco)_to Letter E. Swartz 8205110447 05/04/82 3
" Control of Heavy Leads ~?!UPEG-0612" Letter.E. Swartz (Ceco) to 8205260064 05/17/82 I.
D. Eisenhut (NRC) " Control of-i Heavy Loads NUREG-0612" I
.D.Eisenhut-(HRC)(CECO)to Letter E. Swartz 821230119 11/18/62
" Control of Heavy Loads, Supplemental Response to Draft _TER"
['
3.
VERIFICATION DOCUMENTS:
i TITLE NUDOCS NO.-
DATE
~
j."
f
(.
I
- l-,'
4 I
(
'I_'
)
I n
f-
l' PLANT Dresden Unit Nos. 2 and 3 DOCKETN0(S).
50-237 and 50-249 PROJECT MANAGER Byron L. Siegel TECHNICAL CONTACT W. Koo US! NO. a 42 TITLE pipe Cracks in Boiling Water Reactors l'
MPA NO.
TAC NOS. 69132/69133 ISSUES
SUMMARY
This USI was resolved in February 1981 with the publication of NUREG-0313 P.evision 1, " Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping." That NUREG document was issued to all holders of BWR operating licenses or construction permits and to all applicants for CWR operating licenses. The staff established MPA B-05 for implementation of the resolution at operating plants.
Pipes have cracked in the heat-affected zones of welds in primary system piping in BWRs since mid-1960. These cracks have occurred mainly in Type 304 stainless steel, which is the type used in most operating BWRs. The major recognized to be intergranular stress corrosion cracking (IGSCC) problem is of austenitic stainless steel components that have been made susceptible to this failure by being " sensitized " either by post-weld heat treatment or by sensitization of a i
narrow heat affected zone near welds.
" Safe ends" that have been highly sensitired by furnace heat treatment while attached to vessels during fabrication were found to be susceptible to IGSCC in the late 1960s. Most of the furnace-sensitized safe ends in older plants heve been removed or clad with a protective material, and only a few BWRs still have furnace-sensitized safe ends in use. Most of these, however, are in smaller diameter lines.
Cracks reported before 1975 occurred primarily in 4-inch-diameter recirculation loop bypass lines and in 10-inch-diameter core spray lines.
Cracking is most often' detected during inservice inspections using ultrasonic test techniques.
Some_ piping cracks have been discovered as a result of primary coolant leaks.
NUREG-0313. Revision 1 provided the NRC staff's revised e.v eptable methods for reducing the IGSCC susceptibility of BWR code class 1, 2, and 3 pressure boundary piping of sizes identified above and safe ends.
In addition, it provided the requirements for augmented inservice inspection of piping with nonconforming materials.
As a result of further IGSCC degradations in larger piping, the staff provided licensees with additional reauirements in several NRC communications (i.e.,
Bulletins 82-03, 83-2, and 84-11). The long-term resolution of IGSCC in BWR piping (including the scope of A-42) was provided in NUREG-0313. Revision 2 which was transmitted to all helders of BWR opercting licenses via Generic Letter 88-01, 1HPLEMENTAT10N AND STATUS
SUMMARY
DPESDEN:
The requirements in GL 81-04 were superseded by the requirements in GL 84-11 (MPAB-SA). MPA B-84 was closed as being fully implemented on all BWRs as of February 12, 1988 by memorandum to Thomas E. Murley. GL 81-04 transmitted HUREG 0313, Rev. I to all BWR licensees and provided NRC requirements pertaining to resolution of USl A-42.
The requirements in GL 84-11 were in turn superseded by the requirements in GL 88-01 (MPA B-97).
[
I For the 24 BWRs that were operatino when GL 81-04 was issued, the implementa-tion document is the letter to the' licensee transmitting the staff's evaluation
. of their response to GL 81 04 For the NTOLs at the tirae, the implementation document is the SER or SSER in which the staff evaluated the applicants confor-mance to the requirements of NUREG-0313, Rev. 1.
For the Dresden facility, the closure date for A-42 is May 27, 1982.
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Dresden A-42 i
~
REFERENCES:
l t
1.
REOUIREliENT DOCUMEllTS:
TITLE NUDOCS NO.
DATE i
flVREG-0313, Revisi u 1, " Technical 07/80 i
Report on Material Selection and p
Processing Guidelines for BWR j
L Coolant Pressure Boundary Piping,"
j Generic Letter 81-04, "Implemen-2/26/81 I
tation of NUREG-0313, Rev. I for Selection and Processing Guidelines t
for BWR Coolant Pressure Boundary Piping (Generic Task A-42)"
2.
IMPLEMEllTATION DOCUMENTS:
l TITLE NUDOCS NO.
DATE D. Eisenhut (HRC)(Ceco) to Letter T. Rausch 8206040122 05/27/82 LetterD.Crutchfield(NRC) 8406290164 06/26/84 to D. Farrar (CECO)
Letter A. Schewencer (HRC) 8406210179 06/11/04 to D. Farrar Letter W. Morgan (Ceco) to 8808090125 07/29/88 USNRC Letter B. Siegel (NRC) to 8811090030 11/03/88 H. Bliss (CECO)
Letter T. Ross and B. Siegel 8905310228 05/22/89 4
(NRC) to T. Kovach (CECO)
Letter M. Richter (CECO) to 8903090204 03/01/89 to USNRC Letter M. Richter (Ceco) to 8901030310 12/21/88
_USNRC Letter M. Richter (CECO) to 8907270212 07/21/89 tlSNRC 3.
VERIFICATION DOCUMEllTS:
TITLE NUDOCS NO.
DATE I
m m-
=
m m
~
4 4
?
l'LAllT Dresden Unit Nos. 2 and 3 DOCKET N0(S).
50-237 and 50-249 PP00ECT MA!!AGER Byron L.Sieael TECHNICAL CONTACT P. Gill l
~
i l
USI NO. A-44 TITLE Station Blackout i
MPA NO. A-22 TAC NOS. 68539/68540 ISSUES
SUMMARY
e i
This USl was resolved in June 1988 with the publication of a new rule (10 -
l CFR 50.63) and Regulatory Guide 1.155..
L Station blackout means the loss of offsite ac power to the essential and nonessentit.1 electricc1 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 desigrated as a USI in 1980. A preposed rule was published for comment on
- l'tarch 21 1986. A final rule', 10 CFR 50.63, was published on June 21, 1988 and becameeffectiveonJuly 21, 1988.
Regulatory Guide 1.155 was issued at the same time as the rule and references an industry guidance document,
!!UMARC-8700.
In order to cceply with the A-44 resolution, licensees will be required to:
maintain onsite emergency 'ac pcwer supply reliability above a minimum level develop procedures and training foi recovery from a station blackout determine the duration of a station blackout that the plant should be able to withstand use an alternate qualified ac power source, if available, to cope with a station blackout evaluate the plant's actual capability to withstand and recover from a station blackout backfit hardware :nodifications if necessary to improve coping ability Section 50.63(c)(1) of the rule required each licensee to submit a response including the results of a coping analysis within 270 days from issuarce of an operating license or the effective date of the rule, whichever is later.
It1PLEMENTATION AND STATUS SUPPARY DRESDE!1:
The licensee responded on April 17, 1989 and a working meetina to discuss the submittal was held on October 4, 1989. The staff expects to issue a Safety Evaluatien by the end of the year.
The licensee's submittal proposed the coping approach in combination with some l
hardware changes to satisfy the rule.
However, based on the discussions during the October 4 meeting there appears to be difference of opinion between the staff and licensee regarding the plant's accep, table blackout duration capability, which could impact Dresden's ability to satisfy the rule with the i
F I"
t o
i i
REFERENCES:
Dresden i
A-44 aaproach submitted.
In a December 20, 19B9 meeting the licensee proposed l
c1anges to its submittal to address the staft concerns. These proposed changes will be formally submitted to the staff for review by the end of February 1990.
The staff safety evaluation should be issued by 6/30/90.
l The licensee's implementation date is expected to be not later than June 1992 t
based on the requirement as stated in the rule.
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 I
2.
IMPLEMENTATION DOCUMENTS:
TITLE NUDOCS NO.
DATE tetter:
M.H. Richter (Ceco) 8904240424 04/17/89 to T.E. Murley (NRC) " Response to Station Blackout Rule" l
Memorandum:
B. Siegel and 8911200002 11/09/89 T. Ross to J. Craig "Sunfary of Working Meeting with Ceco to Discuss Ceco's Station Blackout Submittted for Dresden and Quad Cities Nuclear Plants" 3.-
VERIFICATION DOCUMENTS:
TITLE NUDOCS NO.
DATE P
P
E e
PLANT Cresden Nos. 2 and 3 DOCKET N0(S).
50-237 and 50-249 L
PROJECT MA!4AGER Eyron L. Siegel TECHNICAL CONTACT P. Y. Chen USI NO. A-46 TITLE Seismic Qualification of Equipnent in Operating Plants
!^PA NO. B-105 TAC NOS.
49365 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 Qualification Utility Group (SOUG) and Electric Power Research Institute (EPRI). This approach was endorsed by the Senior Seismic Peview and Advisory Panel-(SSRAP) and approved by the NRC staff.
The scope of the review was narrowed to equipment required to bring eech 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 e walkthrough of each plant which is required to inspect equip-r:ent.
Evaluation of equipment will include:
(a) adecuacy 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 outg.rowth o.f the Systematic Evaluatic Program (SEP), the need was n
identified for reassessing design criteria and methods for the seismic cuali-fication of mechanical equipment and electrical equipment. Therefore, the seismic c,ualification 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 SQUG and the staff on a generic basis before individual member utilities proceed with plant-specific implementation.
IPPLEMENTATION AND STATUS
SUMMARY
DRESDEN For All Plants:
The Generic Implementation Procedure (GIP), Revision 0, was submitted by SQUG on June 3, 1988. The staff issued a Gercric Safety Evaluation (SE) on July 29, 1988 endorsing much of the GIF but with ebcut 70 open items to be resolved. After a series of neetings, SQUG submitted Revision I to the GIP on December 23, 1988.
Supplemental information was submitted by SOUG on March 17, 1969. The staff has prepared a supplemental SE for GIP, Revision I and has submitted it to the CRGR for review. The target date for issuance of the supplemental SE is November 1989. An additional supplement is scheduled for June 1990 and overall closecut of implementation projected for 1993.
^
SQUG walkdewns at Dresden wi11 be conducted within two outages following staff approval of GIP.
(
REFERENCES:
Dresden A-46 1
l-1.
REQUIREl'ENT DOCUMENTS:
]
TITLE NUDOCS NO.
DATE Gereric Letter 87-02, "Verifi-cation of Seismic Adequacy of Hechanical and Electric Equipment 1
i in Operating Reactors" 02/19/87 NUREG-1211. " Regulatory Analysis i
for Resolution of Unresolved Safety Issues A-46..."
02/87
~
NUPEG-1030, " Seismic Qualification of Eouipment in Operating Plants, Unresolved Safety issue A-46" 02/87 Letter attached with " Generic Safety Evaluation Report on SQUG GIP,) Revision 0,"from L. Shao (NRC to Neil Smith (SQUG) 07/29/88 2.
IMPLEt'ENTATION DOCUMENTS:
TITLE NUDOCS N0.
DATE "Gercric Implettentation Procedure (GIP for Seismic Verificaticn of Huclear Plant Equipment," Revision 0 06/88
" Generic Implementation Procedure (GIP) for Seismic Verificatinn of Nuclear Plant Equipment," Revision I 12/88 Letter." Response to Generic 8912060061 11/29/89 Letter 89-21, Implementation Status of USI Requirements" 3.
VERIFICATION DOCUMENTS:
TITLE NUDOCS NO.
DATE
.p x -
['
l E'
PLANT Dresden.UniMos.'2and3 DOCKET 40(S).
50-237 and 50 249 PROJECT MANAGCR B. Siegel 1EChMICAL CCNTACT [._Mauck USI NO. A t7 TITLE Safety Implication of Control Systems in LWR
)
Nuclear Power Plants i
MPA NO.
TAC NOS. 74937/74938 ISSUES
SUMMARY
USI A-47 was resolved September 20, 1989, with the publication of Generic Letter (GL)89-19.
i The generic letter states:
"The st'aff has concluded that all PWR plants should provide autnmatic steam generator overfill protection, all BWR plants a
shculd 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 i
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-l tions are consistent with the criteria and the risk considera-l tions of the Commission Interim Policy Statement on Technical
~
l 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 i
operators can mitigate reactor vessel overfill events that may occur via the condensate booster pumps during reduced system i
l pressure operation."
Also, page 2 of the generic letter provides for additional actions for CE and B&W plants. The generic letter provides amplifying guidance for licensees, i -
The generic letter requires that licensees provide NRC with their schedule and commitments within 180 days of the letter's date. The implementation schedule l
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 SPECIF.C):
Response due 3/90.
l u
REFERENCES:
Dresden A-47
[
1.
RE0VIREMENT DOCUMENTS 1
TITLE NUDOCS NO.
DATE Generic Letter 89-19 09/?0/89 i
" Request for Action Related to Resolution of USI A-47" NUREG-1217 " Evaluation of Safety June 1989 Implications of Control Systems in LWR Nuclear Power Plants" NUREG-1218:" Regulatory Analysis July 1989 for Resolution of USI A-47" 2.
IMPLEMENTATION DOCUMENTS:
TITLE NUDOCS NO.
DATE c
3.
VERIFICATION DOCUMENTS:
TITLE NUDOCS NO.
DATE a
t I
i PLANT Dresden Unit Nos. 2 and 3 DOCKETN0(S). 50-237 and 50-249-PROJECT MANAGER Byron L. Siegel TECHNICAL CONTACT J. Kudrick USI NO. A-48 TITLE Hydrogen Control Men'sures and Effects of Hydrogen Burns on Safety Equipnent
-MPA NO.
TAC NOS.
56579/56580 l
ISSUES
SUMMARY
{
The NRC staff concluded April 19, 1989, that USI A-48 is resolved, as stated in SECY 89-122.
t USIA-48wasinitiatedasaresultofthelargeamountofhy(TMI) accident.
drocen generated and burned within containment during the Three Mile Island This issue covers hydrogen control measures for recoverable degraded core accidents for all BWRs and those PWRs with ice condenser containments.
l Extensive research in this area has led to significant revision of the Com.
mission's hydrogen control regulations, given in 10 CFR 50.44, published December 2, 1981.
i 10 CFR 50.44 requires inertin l
method for hydrogen control. g of BWR Mark I and Mark 11 containments as a The BWR Hark I and Mark 11 reactor containments have operated for a number of years with an inerted atmosphere (by addition of im inert gas, such as nitrogen) which effectively precludes combustion of any hydrogen generated. USI A-48 with' respect to BWR Mark I and 11 containments is resolved and implemented for all but seven BWRs with Mark I containments.
1 The rule for BWRs with Mark III containments and PWRs with ice condenser containments was published on January 25, 1985. The rule required that these plants be provided with a means for controlling the quantity of hydrogen produced, but did not specify the control method.
In addition, the task action plan for USI A-48 provided for plant-specific reviews of lead plants for reactors with Hark III and ice condenser containments. Sequoyah was chosen as the lead plant for ice condenser containments and Grand Gulf for Mark 111 containments.
Both of the lead plant licensees chose to install igniter-type i
systems which would burn the hydrogen before it reached threatening concentra-tions within the containment.
Final design igniter systems have been installed not only in both lead plants Sequoyah and Grand Gulf, but in all other ice condenser and Mark !!! plants as well.
The staff's safety evaluations of the final. analyses required to be submitted by these licensees by the rule are i-scheduled for completion in 1989.
Large dry PWR containments were excluded from USI A-48 because they have a l
greater ability to acconnodate the large quantities of hydrogen associated with a recoverable degraded core accident than the smaller Mark I, II, 111 and ice condenser containments.
However, this issue has continued to be considered and, in 1989, hydrogen control for large dry PWR containments was identified as a high-priority Generic Issue (GI) 121. The resolution of G1 121 is being l
actively pursued in close coordination with more recent research findings.
i g
-.++-
l Oresden A-48
?
ISSUES
SUMMARY
(CONT.):
The NRC staff has concluded that USI-A-48 is resolved as stated in SECY 89-122.
If interested, the report should be consulted for further details regarding the relationship of A-48 to other ongoing hydrogen activities.
IMPLEMENTATION AND STATUS
SUMMARY
DRESDEN:
USI A-48 was resolved on April 19, 1989, as stated in SECY 89-122.
It is considered fully implemented at BWR Mark I and Mark 11 facilities, as these facilities use inerting as a method of hydrogen control.
There is a related issue that remains open on several Mark i facilities. This issue is associated with the requirement to have a recombiner capability at all facilities. Generic Letter 84-09, "Recombir r Capability Requirements of 10 CFR 50.44 (c)(3)(22)," provided guidance to uose l' ark I facilities that elected to rely on inerting in lieu of recombiner capability. This aspect is considered separate from A-48.
The facilities involved are Cooper, Millstone 1 Oyster Creek, Dresden, and Quad Cities.
1 On January 20, 1987 a meeting was held between the staff and utility representatives, including Commonwe61th Edison Company,. to discuss the system used in their plants for combustible gas control.
As a result of this meeting the staff requested that each licensee submit its plant specific position on its compliance with'10 CFR 50.44(g).
The licensee has not'to date provided the information requested by the staff.
In an attempt to obtain resolution of this issue the staff, in a letter to the licensee dated May 3,1989, requested that a meeting be held with Ceco to review the current status of combustible gas control at Dresden and Quad Cities. This meeting has been postponed pending a legal and technical determination by the staff if Oyster Creek, as it is currently designed, is in compliance with the requirements of 10 CFR 50.44. The staff intends to issue a position paper on the issue of compliance with 10 CFR 50.44 for the outstanding Mark I plants sometime in early 1990, i
l
{
(
I
=
i-i I
REFERENCES:
Dresden d
A-48 1.
REQUIREMENT DOCUMENTS:
t TITLE NUDOCS NO.
DATE 10 CFR 50.44, Standards for 12/81 Combustible Gas System in Light-Water-Cooled Power Reactors SECY-89-122, Resolution of 04/19/89 U51 A-48, " Hydrogen Control Measures and Effects of Hydrogen Burns on Safety t
Equipment" i
2.
IMPLEMENTATION DOCUMENTS:
TITLE NUDOCS NO.
DATE l
LetterfromT.Rausch.(CECO) 8209220196 09/15/82 to D. Eisenhut (NRC)
Letter from B. Ryback.(CECO) 8407030092 06/25/84 to H. Den' ton (NRC)
-Letter from D. Vassallo (NRC) 8509300568 09/26/85 tn D. Farrar (Ceco) l Letter from J. Wojnarowski (Ceco) 8511220234 11/14/85 to D. Vassallo (NRC)
Letter from R. Bernero (NRC) to 8608190542 08/11/86 D. Farrar (Ceco)
LetterfromJ.Wojnarowski(CECO) 8912200424 09/16/86 to J. Zwolinski (NRC)
LetterfromM.Grotenhuis(NRC) 8705140076 04/24/87 to D. Farrar (Ceco)
LetterfromG.Holahan(NRC) 8905110238 05/03/89 to C. Reed (Ceco) 3.
VERIFICATION DOCUMENTS:
TITLE NUDOCS NO.
DATE
Ecciosure 3 Fage No.
1
{
02/01/90 LISI!N6 0F INCOMPLETE US! DATA i
FOR INPUT TROM PROJECT RANA 6EkS
!!$UE ISSUE DESCRIPi!VE NAME IMPLEMENT IMPLEMENT LICENSEE COMMENT STAFF COMMENT NUMBER DATE STATUS t
8I PLANT NAME: DRE! DEN 3
. A 01 NATER HAMMER
//
NC A-02 ASYMMETRIC DLONDONN LCADS ON
//
N/A PNR ONLY REACf0RPRIMARYCOOLANTSYSTEMS P
A-03 NESTIW6 HOUSE STEAM BENERATOR TUDE //
N/A NESI!N6 HOUSE ONLY INTEGRITY A-04 CE STEAM BENERATOR TUBE INTE6R!if //
N/A CE PLANTS ONLY A-05 66N STEAM GENERATOR TUDE
/!
N/A HNPLANTSONLY INTEBRITY A-06 MAFK ! SHORT TERM PROBRAM 02/31/76 C DELTAPCONTROL A 07 MARK 1 LONE-TERM PRDERAM 06/31/94 C ENFORCEMENT IAKEN A-09 MARK !! CONTAINMENT POOL DYNAMIC
/ /
N/A MK!!BNRONLY LOAD!-LON6-TERMPROGRAM A09 ATNS
//
1
$!VERSITY 6 T/S A 10
$NRFEEDNATERN0l!LECRACK!N6 05/31/B2 C A-11 REACTOR VESSEL MATERIALS
//
NC 70U6HNESS A12 FRACTURE TOU6HNEES OF STEAM
/ /
N/A CP AFTER 83 DNLY 6ENEF.ATOR AND REACTOR COOLANT PUMPSUPPORTS A 17 SYSTEMSlhiERACTION
//
NC IPE NOREQUIREMENTS A :4 00ALIFICATIONOFCLASS1E 06/30/66 C SAFETYRELATEDEQUIPMENT A 26 REACTOR VESSEL PRESSURE TRANS!ENT //
N/A PNR DNLY F101Eti!DN A!!
RE SHUiDONN REQUIREMENTS
//
N/A NEN PLANTS ONLY. SRP.
A-36 CCNTRCL OF HEAVY LOADS NEAR SPENT 01/31/t5 C 6L-65!! ENDED FUEL A 39 DETERMINAi!0NOFSAFETYRELIEF
//
NC SEEA-07 VALVE POOL CYNAM!C LOADS AND TEMFERATURE LIMITS A 40 SEISMIC DES!6N CRlIER!A -
//
NC SUBSUMMED BY A-46 SHORT TERM PR06 RAM A 42 P!FE CRACKS IN 80!LIN6 NATER
//
NC P!PE REP 1985/1986 REACTORS A-43 CONTAINMENT EMER6ENCY SUMP
//
NC INFO ONLY PERFORMANCE A 44 STAi!CN BLACK 0UT 06/30/92 1 SER6/30/90 A 45 SHUIDCNN DECAY HEAT REMOVAL
//
NC SUBSUMEDBYSEVEREACC REQUIREMENTS A 46 SE!$MIC QUAllFICAT10N OF
//
I RED UNDER DEVEL l
EQUIPMENT IN OPERAi!N6 PLANf6 l
A-47 SAFETY IMPLICATIONS OF CONTROL 03/31/90 E NENREQUIREMENTS l'
SYSTEMS l
A 49 HYI,R06EN CONTROL MEASUEES AND
//
NC lhERTED BUT NO CAD SYS EFFECTS OF HYDR 06EN BURNS ON SAFETY EDUlPMENT A 49 TRES$URl!EDTHERMALSH0CK
//
N/A PNR ONLY
p u n.9 UNITED STATES I
NUCLEAR REGULATORY COMMISSION i
,}
wAsmwatow, o. c. mss
\\.... + /'
February 13, 1990 Docket Nos. 50-373 and 50-374 MEMOPANDUM FOR:
File FROM:
Paul C. Shemanski, Project Manager Project Directorate 111-2 Division of Reactor Projects - III, 1Y, Y and Special Projects
SUBJECT:
STATUS OF IMPLEMENTAT!0N OF UNRESOLVED SAFETY ISSUES The current implementation status of unresolved safety issues (USIs) at the LaSalle facility is set forth in the enclosures to this memorandum. contains a copy of the information provided by the licensee in its response to Generic Letter 89-21.
In addition, Enclosure 2 contains a status summary for each USI applicable to this facility. This status sumary is based upon the licensee's response to the Generic Letter, discussions with the licensee, and my review of available NRC records and information.
. is a copy of the staff's data base printout for LaSalle facility.
It reflects the staff's assessment of US1 implenentation for all 27 USIs.
It is based on review of the licensee's response to Generic Letter 89-21, and evaluation by project managers, the US! team, and NRR technical staff.
For those items that are incomplete my assessment of the schedular significance is as follows:
A-9:
Ceco is upgrading the RPT logic and, the controversy between the NRR technical staff and the owner's group on the subject of diversity continues. Expected completion date is 6/91 for Unit I and 3/92 for Unit 2.
A-11:
Ceco evaluating actions required, expected response date S 2/01/90. This is acceptable.
A-44:
SE scheduled for 3/31/91. This is acceptable.
A-47:
Ceco response to GL 89-19 is due 3/19/90. This is acceptable.
Appropriate NRR technical review branches have also reviewed this USI summary and memo.
0C 5
=h Paul C. Shemanski, Project Manager Y / d M [, 8 N T p (t/ Q g Project Directorate 111-2 Division of Reactor Projects - Ill,
Enclosures:
As stated cc w/ enclosures:
K. Eccleston
.q INDEX i
US1 NO.
PAGE NO.
US1 MASTER FORM PAGE 1 1.0 U51 MASTER FORM PAGE 2 2.0 A-1 3.0/4.0 i
A-2 5.0/6.0 A-3, 4, 5 7.0/8.0 A-6 9.0/10.0 A-7 11.0/12.0 A-8 13.0/14.0 A-9 15.0/16.0 A-10 17.0/18.0 A-11 19.0/20.0 A-12 21.0/22.0 A-17 23.0/24.0 A-24 25.0/26.0
-A-26 27.0/28.0 A-31 29.0/30.0 A-36 31.0/32.0 A-39 33.0/34.0 A-40 35.0/36.0 A-42 37.0/38.0 A-43 39.0/40.0
- A-44 41.0/42.0 A 45 43.0/44.0 A a6 45.0/46.0 A-47 47.0/48.0 A-48 49.0/50.0 A-49 51.0/52.0 5.
. ~
}
4
'p.
/
\\ COmm0RWOSith Edit 0R
> n wem Asems sn et. en.cago. nno EMcLorogE i
(
Adotget Reply to. Poet Othee 50s 767 i
cnicago, nn:ns 60690 0767
'i November 29, 1989 1
Dr. Thomas E. Murley, Director Office of. Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C.
20555 i
Subject:
LaSalle County Station Units I and 2 i
Response to Generic Letter 89-21 (Request for information concerning-
-i status of implementation of unresolved safety issue (USI) requirements)
NRC Docket Nos. 50-373 and 50-374 Reference (a) Generic Letter 89-21 dated October 19, 1989 5
Dear Dr. Murley:
.i o
b Reference (a) requested that holders of operating licenses and I
construction permtts for nuclear power reactors review and report the status of implementation of Unresolved Safety Issues (USI) for which a final technical resolution has been achieved and which are applicable.to their respective facility.
The following attachment provides Commonwealth Edison LaSalle County Station Units I and 2 status of implementation for the USI.
If you have any additional questions regarding this matter please contact this office.
Very truly yours, W
rg Nuclear Licensing Administrator l
l cc:
A.B.' Davis - Regional Administrator, RIII P.C. Shemanski - Project Manger NR V Senior Resident Inspection - Lasalle County Station O L A 1 YVs f,
1
[]
[
NY) w w
~
,,.. - - -.... ~.
t STATUS OF UNRESOLVED SAFETY JSSUES LA1Altr mWTY STATION UNITS I and 2 i
f USI NUMBER STATUS /DATE REMARKS A-1 C / 3-81 NUREG 0519 SER dated 3/81, App. C, Page i
C-7&B A-2 NA A-3 NA A-4 NA A-5 NA A-6 NA A-7 NA A-8 C l'U 4-82 U-1, NUREG 0519 supplement #6 dated U 12-83 11/83 page 3-2 & IR 373/82-49. U-2 E. Adensam letter to D.L. Farrar dated 8-29-86 & IR 374/87018 A-9 C / U 6-88 N. Morgan letter to NRC dated 4-28-89 U 8-88 in response to NRC SER of 1-5-89, additional upgrade to RPT logic is to i
be done on Unit-1; 6-91 and Unit-2; 3-92.
A-10 C / 3-81 NUREG 0519 SER, App. C, Page C-10.
A-11 E
Evaluating actions required, expected response Date: 1-15-90 A NA A-17 NC Generic Letter 89-18 closes this issue.
i A-24 C / U 11-85 Unit 1 IR 373/87003.
U 11-85 Unit 2 IR 374/88026.
A-26 NA
'l A-31 NC NUREG 0519 SER dated 3/81 section 5.4.2 A-36 C / 3-85 A. Schwencer letter to D. L. Farrar~ dated 3-12-85 e
A-39 C / U 4-82 NUREG 0519 supplement #6 dated 11/83 U 12-83 page 3-2 A-40 C
NUREG-1233 dated 9/89
.0417T:2 L
STATUS OF UNRESOLVED SAFETY ISSUES F
I; k
USI NUMBER STATUS /DATE REMARKS A-42 C / 7-29-88 Unit 1. IR 373/85035.
Unit 2. IR 374/87002.
A-43 NC / 3-81 NUREG 0519 SER (Resolution in SECY 85-349 required no changes).
A-44 I
M. Richter letter to T. Murley, dated 4-17-89 projected completion date U-1:
refueling outage starting May 1991.
U-2:
refueling outage starting December 1991-A-45 NC SECY 88-260 transferred action for this issue to the IPE Program.
A-46 NA A-47 E
Evaluating actions required, expected r
response da.te 3-19-90.
A-48 C / 3-81 SECY 89-122 A-49 NA C - Complete NC - No Changes Necessary NA - Not Applicable I - Incomplete E - Evaluating Actions Required L
l i
l 0417T:3 1
4 ENCLOSURE 2 i
k i
g STATUS
SUMMARY
OF US1's APPLICABLE TO LASALLE COUNTY
[
STATION, UNITS 1 AND 2 DECEMBER 11,1989 PAUL C. SHEMANSKI 8
e I
l 4
e i
@J 6
PLANT LaSalle, Units 1 and 2 DOCKET N0(S).
50-373/374 PROJECT MANAGER Paul C. Shemanski TECHNICAL CONTACT A. Serkir USI NO.
A-l' TITLE Water Hammer HPA NO. N/A TAC NOS.
ISSUES
SUMMARY
ThisUnresolvedSafetyIssue(US1)wasresolvedinMarch1984 with the publicationofNUREG-0927."EvaluationofWaterHammerinNuclearPowerPlants-
- Technical Findings Relevant to Unresolved Safety Issue A-1."
Also on March 15, 1984, the EDO sent the Commissioners SECY 84-119 titled, " Resolution of Unresolved Safety Issue A-1, Water Hammer."
In SECY 84-119, the staff concluded that the frequency and severity of water herrer occurrences had been significantly reduced through (a) incorporation' of design f eatures such as keep full systems, vacuum breakers, J-tubes - void detection systems, and improved venting procedures; (b) proper design of feed-water valvrs and control systems; and (c) increased operator awareness and
. training. Therefore, the resolutinn of USI A-1 did not involve any hardware or design changes on existing plants.
ItdidinvolveStandardReviewPlan(3RP).
changes-(forward fits) and a comprehensive set of guidelines and criteria to evaluate and upgrade utility training programs (per THI Task Action Plan item I.A.2.3).
In addition, the assumption was made that for BWPs 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 hanmer that opened a direct pathway to the etivironment.
1tjyLEMENTATION AND STATUS
SUMMARY
(PLANT SPECIFIC):
1.
By letter dated November 29, 1989, Commonwealth Edison indicated that TM1 Task Action Plan Item I. A.2.3 was implemented in March 1981.
2.
LaSalle County Station does not have an isolation condenser.
i 9
__j
l l
RFFERENCES:
LaSalle Unit I and 2 i
j.-
A-1 1.
PEOUIREMENT DOCUMENTS:
' TITLE NUDOCS.NO.
DATE Letter from Denton to Utilities.
8403150310 03/05/84 l
" Notice of Issuance and Availability NUREG-0927 Rev. 1 Safety issue A-1" l
2.
IMPLEFEtlTATION DOCUMENTS:
TITLE NUDOCS NO.
DATE j
NUREG-0927 " Evaluation of Water 8306060413 05/31/83 Hammer in Nuclear Power Plants-Technical Findinos Relevant to Unresolved Safety issue A-1" NUREG-0993 Rev. I 8306060418 March 1984
" Regulatory Analysis for for USI A-1, Water Hammer" i
SRP Sections:
3.9.3, 3.9.4, S.A.6, 5.4.7, 6.3, 9.2.1, 9.2.2, i
10.3, and 10.4.7 i
SECY-84-119, " Resolution 03/15/84 of Unresolved Safety A-1, Water Hamer" l
i.
LetterfromW.E. Morgan (CECO) 8912970245 11/29/89 i
to T. E. Furley - Response to GL 89-21 NUREG 0519 SER, App. C.
8103230641 Page C-7 & 8 Parch 1981 3.
VERIFICATION.0OCUMENTS:
TITLE NUDOCS.NO.
DATE e
I
r l
sPLANT LaSalle. Units 1 and ?
DOCKETN0(S).$0-373/374 5
PROJECT MANAGER Paul C. Shemanski TECHNICAL CONTACT J. Kudrick l
051 NO. A-8 TITLE Mark 11 Containment Pool Dyna 6 tic Loads MPA 1:0.
TAC NOS.
ISSUES
SUMMARY
6.
US1 NO. A-8 TITLE: Mark 11 Containment Pool Dynamic Loads This US1 was resolved in August 1981 with the publication of NUREG-0808, " Mark 11 Containment Program Load Evaluation and Acceptance Criteria," and Standard ReviewPlan(SRP)Section6.2.1.10. The requirement is that the 11 BWRs having the Mark 11 containment shall meet the requirements of GDC 16.
As stated in NUREG-0808, the original design of the Mark 11 containment system considered only those loads normally associated with design-basis accidents that were known at the time.
These included pressure and temperature loads associated with a LOCA, seismic loads, dead loads, jet impingement loads, hydrostatic loads due to water in the suppression chamber, overined pressure test loads, and construction loads. However, since the establishment of the original design criteria, additional loading conditions were identified that must be considered for the pressure-suppression containment-system design.
In the course of performing large-scale testing of an advanced design pressure-suppression containment (Mark III), and during inplant testing of Mark I containments, new suppression-pool hydrodynamic loads were identified that had not been included explicitly in the original Mark !! containment-design basis.
These additional loads result from dynemic effects of drywell air and steam being rapidly forced into the suppression pool during a postulated LOCA and from suppression-pool response to safety / relief valve (SRV) operation; these are generally associated with plant transient operating conditions.
Because these new hydrodynamic loads had not been considered, the NRC staff determined that a detailed reevaluation of the Mark 11 containment system was required.
The issuance of NUREG-0808, NUREG-0802, Safety Relief Valve Quencher loads:
Evaluation for BUR Mark 11 and 111 Containments," and NUREG-0487, " Mark 11 Containment Lead Plant Program Load Evaluation and Acceptance Criteria,"
i documented acceptable methods for calculating the hydrodymanic loads associated i
with plant transient conditions. Specifically, the loads referenced in these
[
1 HRC staff reports, as modified by the acceptance criteria, constituted the l
resolution of USI A-8.
SRP Section 6.2.1 has been modified to reflect the l
applicability of these reports to Mark 11 containment evaluations.
l Implementation is believed to be complete for all Mark 11 BWRS. As part of the licensing process, the staff required that the applicants utilize the new l
calculation methodology defined in the reference documents before a full power y
license was issued.
1MPLEMENTAT10N AND STATUS
SUMMARY
(PLANT SPECIFIC):
1.
By letter dated November 29, 1989 Comonwealth Edison indicated that US1 A-8 was completed on 4/28 and 12/83 for Units 1 and 2, respectively, l-
,w
REFERENCES:
LaSalle Units 1 and 2 A-R 1.
RE001REMENT DOCUMENTS:
TITLE NUDOCS NO.
DATE GDC-16 Containment Design NUREG-0B08 " Mark 11 Containment Program Load Evaluation and Acceptance Criteria" August 1981 Standard Review Piant 6.2.1.1.c.
t
" Pressure Suppres'. inn Type BWR Containmente/-
Revision 1-4 NUREG-0487, " Mark 11 Containment November 1978 Lead Plant Program Load Evaluation and Acceptance Criteria" a.
Supplement 1 September 1980 b.
Supplement 2 February 1981 j
NURG-0802, " Safety Relief Valve October 1982 Quencher Loads: ' Evaluation for BWR tiark 11 and 111 Containments" 2.
IMPLEPEllTATION DOCUMENTS:
9 TITLE NUDOCS NO.
DATE NUPEG-0519 Supplement #6; 8312220175 11/83 page 3-2 E. Adensam letter to D.'t. Farrar 8609050115 8/29/86 3.
VERIFICATION. DOCUMENTS TITLE NUDOCS NO.
E D
Inspection Report 8212030021 1982 i
373/82-49 Unit 1 Inspection Report 8707210751 1987 374/87018. Unit 2
i i
PLANT LaSalle. Units 1.and 2 DOCKET N0(S).
50-373/374 i
PROJECT P.AliAGER Paul C. Shemenski TECHNICAL CONTACT J. Mauck.
]
USI NO. A-9 TITLE ATWS per 10.cFR 50.67 o
MPA NO.
TAC N05. 59107/59108 i
ISSUES
SUMMARY
This USI was resolved in June 1984 with the publication of a final rule (10 CFR l
50.62) to recuire improvements in plants to reduce the likelihood of failure of the reactor protection system (RPS) to shut down the reactor following j
anticipated transients and to mitigate the consequences of an anticipated transientwithoutscram(ATWS) event.
The rule includes the following design-related requirements:
50.52(C)(1),
diverse and independent auxiliary feedwater initiation and turbine trip for all PWRs; 50.62(C)(2), diverse scram systems for CE and B&W reactors; 50.62(C)(3) i alternate rod in.iection (ARI) for BWRs; 50.62(C)(4); standby licuid control system (SLCS) for BWRs; and 50.62(C)(5), automatic trip of recirculation pumps under conditions indicative of an ATHS for BWRS.
Information requirements and an implementation schedule are also specified.
TitPLEMFl!TATION AND STATllS.SllMMARY (PLANT SPECIFIC):
Commonwealth Edison implemented the ATWS modifications (alternate rod injection, recirculation pump trip, and standby liquid control system) on 6/88
~
and 8/8B for Units 1 and 2, respectively. Additional upgrade to RPT logic is to be done on 6/91 and 3/92 for Units 1 and ?, respectively, in addition, controversy between the NRR technical staff and the owner's group on the issue
-of diversity continues.
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i
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'q
REFERENCES:
LaSalle, Units 1 and 2 A-9 1.
RE001REMENT DOCUMENTS:
TITLE NUDOCS.WO.
DATE NUREG-0460, and Supplements.
03/B0
" Anticipated Transients Without Scram for Light Water Reactors" Federal Register Notice 49FR26045(10CFR50.62) 06/26/84 2.
IMPLEMENTATION DOCUMENTS:
TITLE NUDOCS NO.
DATE Letter from W. Morgan to 8905020241 4/28/89 NRC in response to NRC SER 8901110211 1/5/89 3.
VERIFICATION DOCUMENTS:
i TITLE NUDOCS NO.
DATE f
n 9
?
i PLANT LaSalle. Units 1 and 2 DOCKET N0(S). 50-373/374 PROJECT mat!AGER Paul C. Shemanski TECHNICAL CONTACT K. Wichman US) NO. A-10 TITLE PWR Feedwater Nozzle Cracking MPA NO.
B-25 TAC N05.
ISSUES
SUMMARY
1 This issue was resolved in November 1980 with the publication of NUREG-0619,
- BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking." MPA j
B-25 was established by NRC's Division of Licensing for implementation purposes.
Inspections of operating BWRs conducted up to April 1978 revealed cracks in the feedwater nozzles of 20 reactor vessels.
It was determined that cracking was due to hich-cycle fatigue caused by fluctuations in water temperature within the vessel in the nozzle region.
By letter dated November 13, 1980, Darrell G. Eisenhut provided licensees with a copy of NUREG-0619. The letter stated that NUREG-0619 provided the resolu-tion of the staff's generic technical activity US1 A-10, which resulted fe m the inservice discovery of cracking in feedwater nozzles and control rod drive return line nozzles.
NUREG 0619 describes the technical issues, General
. 1 Electric and staff studies and analyses, and the staff's positions and require-ments. Licenseeswererequiredtorespond,Pursuantto10CFR50.54(f),that they would meet implementation dates indicated in NUREG-0619.
Generic Letter 81-11 was subsequently issued to provide technical clarification to the November 13, 1980 letter, to clarify that it had been sent to PWR licensees for information only, and that no response was required from PWR licensees.
_1PPLEMENTAT10N Af40 STATUS
SUMMARY
(PLAlti SPFCIFIC):
NUREG-0619 includes and evaluation of the GE Topical Report NEPE-21821 that addresses solutions to the generic feedwater nozzle cracking concern. LaSalle feedwater spargers are stainless steel headers with six headers served through six feedwater nozzles, each fitted with triple thermal sleeves which meets NUREG-0619. The control rod drive return nozzle has been capped and the line eliminated thus eliminating the thermal cycling problem identified for these nozzles. Connonwealth Edison states in NUREG-0519, 3/18, App. c, Page C-10 that LaSalle complies with NUREG-0619.
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REFERENCES:
LaSalle, Units 1 ard ?
i A-10 i
i 1.
RE001REMENT DOCUMENTS:
i i
TITLE NUDOCS NO.
DATE i
letter from D. Eisenhut l
transmitting NUREG-0619, "BWR Feedwater Nozzle and Control Rod Drive Return Line N0221e Cracking,"
j resolution of A-10 to licensees 11/13/80 l
i Generic Letter 81-11. "BWR Feedwater Norrle.and Control Rod Drive Return Line Nozz7e l
CracUng(NUREG-0619)"
02/?0/81 I
2.
IMPLEMENTATION DOCUMENTS:
{
TITLE NUDOCS.NO.
D,Al i
i;UREG-0519 SER 8103230641 3/81 App. C, Page C-10 i
3.
VERIFICATION 00Ct!MENTS:
TITLE NUDOCS.NO.
DATE 5
i
)
i l
l
_~.
l I
PLANT::LaSalle. UAit 1 and 2
'DOCKETN0(S).
50-373/374 1
E Re to Vesse tr 1 ne
-MPA NO.-
TAC NOS.
ISSUES
SUMMARY
This-USI was resolved in' October 1982 with the publication of NUREG-0744,
" Pressure Vessel Material Fracture Toughness.". NUREG-0744 was issued by l 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
- c 82-26 was required.
Because of.the remote possibility that._ nuclear reactor pressure vessels
-designed to the ASME Boiler and Pressure Vessel Code would feil, the design of nuclear, facilities does not provide protection against reactor vessel f ailure, j
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 Lproperties 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 p
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.
m
~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 belovi 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.
'IMPLEMENTATIDN AND STATUS
SUMMARY
'(PLANT SPECIFIC):
Commonwealth Edison is evaluating actions that are required. The expected response date:is 2/01/90.
BasedJon minimum acceptable impact values of 20' foot-pounds and fabrication
- techniques employed on the La Salle vessels, the licensee conservatively-estimates that the total fluence over the design life would result in the end
~of. fracture toughness above the minimum charpy impact requirement of 50 foot-pounds.
In addition, the surveillance program required by Appendix H of
'10 CFR Part 50 will afford an opportunity to reevaluate the fracture toughness periodically during the first half of the design life.
~
NUREG-0519 SER dated 3/81 App. C Page C-10 describes the staff's evaluation of this' issue.
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REFERENCES:
.LaSalle, Units 1.and 2:
A-11~
.x
' 1, REQUIREMENT' DOCUMENTS:
TITLE NUDOCS NO.
DATE
'j NUREG-0744, Revision 1. " Pressure 10/82 Vessel Meterial Fracture Toughness" f
Generic Letter 82-26,'" Pressure vessel Material Fracture Toughness" 11/12/82
?.
IMPLEMENTATION DOCUMENTS:
' TITLE NllDOCS HO.
DATE
-i i
3.
VEP1FICAT10N DOCUMENTS:
TITLE' NUDOCS NO.
DATE
-n t
i PLANT t.aSalle. Units'1.and 2 DOCKETN0(S).
50-373/374 PROJECT MANAGER-Paul C. Shemanski.
TECHNICAL CONTACT D. Thatcher.
j US1 NO.
A TITLE Systems Interactions in Nuclear. Power. Plants...
t tiPA NO.
TAC NOS.
-ISSUES
SUMMARY
Generic letter (GL) 89-18, dated September 6, 1989, was sent to all power reactor licensees and constitutes the resolution of USI A-17.
The generic
. letter did not require any licensee actions.
GL 89-18 had two eucMurts which (a) outlined the bases for the resolution of USl A-17, and (b) 1%vided l'ive general lessons learned from.the review of the overall systems intuaction issue. The staff anticipated that licensees would review this informatiOn in other prograa>. tuch as the Individual Plant-Examination (IPE)forSevereAccidentVulner6bilities. 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 t
IPE~ program. Also considered'in the resolution of this US1 was the. expectation that licensees would continue to review information on events at operating nuclear'. power plants in accordance with the requirements of Tt11 Task Action Plan item 1.C.5 (NUREG-0737).
ItiPLEMENTATION AND STATUS SUNNARY (PLANT SPECIE 1C):
No licensee action is required. GL 89-18 closes this issue.
Resolution of USl A-17 assumed that flooding issues were reviewed and resolved as necessary. Section 10.6 of the SER related to the operation of LaSalle E
(NUREG-0519,3/81) finds that in the event of failure of the circulating water system,. flooding will not affect safety-related equipment. Appendix C to this
'SER also addresses this issue.
The licensee has no record of receiving'the 9/72 NRC letter concerning potential failure of non-category 1_ equipment..
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REFERENCES:
LaSalle, Units l'and 2 A-17 1.
REQUIREt1ENT DOCUMENTS:
TITLE NUDOCS NO.
DATE l
Generic Letter 89-18 09/06/89 NUREG-1174 " Evaluation of May 1989 Systems Interactions in Nuclear Power Plants" NUREG-1229 " Regulatory Analysis August 1989 for Resolution of USI A-17" NUREG/CR-3922 " Survey and January 1985 Evaluation of System Interaction Events and Sources" NUREG/CR-4261 " Assessment of June 1986 System Interaction Experience in Nuclear Power, Plants" NUREG/CR-4470 " Survey and August 1986 Evaluation of Vital Ir.strumentation and Control Power' Supply Events" NRC Letters to Licensees 9/72 Informing Licensees of Staff Concerns Regarding Potential Failure of Non-Category I Equipment 2.
IMPLEMENTATION-DOCUMENTS:
TITLE NUDOCS NO.
DATE L
L l
3.
VERIFICATION DOCUMENTS:
1 TITLE NUDOC NO.
DATE 1
l-
m, i
PLANT L'aSalle, Units 1 and 2 DOCKET'N0(S),
50-373/374 PROJECT MANAGER Paul C. Shemanski TECHNICAL CONTACT P. Shemanski USI NO.
A-24' TITLE Qualification of Class 1E Equipment P
MPA NO.-
B-60 TAC NOS. 42536/42537 ISSUES
SUMMARY
1 This USI was resolved in July 1981 with the publication of NUREG-0588, Revision 1,," Interim Staff Position on Environmental Qualification of Safety-Related 1
-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 m
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 l
s
. methods in national standards, regulatory guides, and certain NRC publications, including NUREG-0588.
The rule is based on the DOR Guideline.s and HUREG-0588. These provide guidance on (a) how to establish environmental service conditions, (b) how to select i
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 (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 p
whether each licensee has implemented the resolution of A-24 may therefore be l-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):
- 0n March 28, 1985 the staff issued an SER to extend the schedular requirements for specific eouipment subject to the environmental qualification rule to
-November 30, 1985.
For all other EQ equipment, the staff concluded that the licensee demonstrated compliance with the requirements of 10 CFR 50.49.
The licensee completed implementation of A-24 on November 30, 1985.
An EQ-inspection in 1987 resulted in Inspection Reports IR 373/87003 and IR 374/88026 4
for Units 1 and 2, respectively.
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REFERENCES:
Plant Name.
- A-24 I
'1.
.REOUIREMENT DOCUMENTS:
TITLE NUDOCS NO.--
~DATE
.D0R " Guidelines for Evaluatina 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.
l
'CLI-80-21, on 00R Guidelines and HUREG-0588 05/23/80 i
NUREG-0588, Revision 1 07/81 10CFR50.49.(48FR2730-2733) 01/21/83 l:
Standard.and Review Plan 3.11 EnvironmentalQualificationof L
Hechanical and Electrical Equipment 07/81 2.
IMPLEMENTATION DOCUMENTSi I'
TITLE NUDOCS NO.
DATE
.NUREG-0519, SER 8308290385 3/83 L
-Supplement #5, Section 3.11~
L 4
l' Letter D. L. Farrar to 8502200269 2/14/85 NRC, 10 CFR 50.49 certification l
Letter A. Schwencer to D. Farrar 8504080036 3/28/85 L
.(CECO), SER for 10 CFR 50.49 L
compliance ~
3 '..
VERIFICATION DOCUMENTS:
TITLE NUDOCS N0.
DATE Inspection Reports IR 373/87003 8703240396 1987 IR 374/88026 8812130133 1988 k
PLANT LaSalle. Units 1 and 2
.DOCKETN0(S). 50-373/374 1
PROJECT 14ANAGER Paul C. Shemanski TECHNICAL CONTACT
'R. Jones i
USI~NO. -A-31 TITLE RHR Shutdown Reovirements j
1
- HPA NO.
TAC NOS.
L' ISSUES
SUMMARY
This USI was resolved in May 1978 with the publication of Standard Review Plan j
(SRP) Section 5.4.7..
Only those plai.ts 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 4
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 applications were docketed before January 1,1978, and for which an Operating License (OL) issuance was l
expected on or after January 1, 1979. Class 3 affected all operat1ng reactors and all other plants for which issuance of the OL was expected before January 1, 1979. The extent to which Class 3 plants would require implementation was based on'the combined staff review of related plant features.
In general, the L.
outcome of these evaluations were that'only plants receiving an OL'after January 1, 1979 were affected by this UST resolution, and there were no backfits to operating plants that'had received an operating license before January 1,1979.
ItiPLEMEt' TAT 10N-AND STATUS
SUMMARY
(PLANT SPECIFIC):
O Section 5.4.2 cf the SER (NUREG-0519, 3/81) related to the operation of LaSalle' County Station states the' residual heat removal system is designed to i
seismic Category I requirements.
It is protected against the effects of ficoding, tornadoes, hurr.icanes, and other natural phenomena by the reactor i
l building in which it is housed (as discussed in Section 3.8 of the SER). which
?
conforms with the requirements of Regulatory Guide 1.29 " Seismic Design p
Classification," and Criterion 2 of the General Design Criteria. The containment itolation requirements of Criteria 55, 56, and 57 of the General Design. Criteria are discussed in Section 6.2 of the SER.
Systems used for-cooling the residual heat removal system conform to the requirements of Criteria 44, 45, and 46 of General Design Criteria.
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REFERENCES:
LaSalle, Units.1 and 2' A-31.
-i
~1.-
RFOUIREMENT:OOCUMENTS:L j
~ TITLE-NUDOCS NO.
DATE:
3 NUREG-0800." Standard Review Plan,"
5/78; SRP Section 5.4.7 NUREG-0606 " Unresolved Safety Issues Summary"
. Regulatory Guide 1.139, " Guidance 1
for Residual Heat Removal" Regulatory Guide.1.113-2.
IMPLEMENTATION DOCUMENTS:-
L 1 TITLE NUDOCS NO.
DATE
~
NUkEG-0519, SER 8193230641 3/81 Section 5.4.2 3.
VERIFICATION DOCu ENTS:
TITLE NUDOCS.NO.
DATE v
i I{
l
-=
a ww-
PLANT _ tasalle. Units'1 and 2-DOCKET N0(S).
50-373/374 PROJECT HAHAGERf Paul C. Shemanski TECHNICAL CONTACT J. Wermiel US1 fl0. A-36 TITLE Control of Heavy Loads. Phases 1 & 11 MPA NO..C-10 'C-15 TAC N05, 52240 (Unit 1) i 1SSUES
SUMMARY
This USl 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 11, respectively, of the resolution of this issue at operating pla nt s'.
t 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, em aquipment that may be required to achieve safe a
shutdown and continue. decay b' emoval.
USI A-36 was established to
- ing criteria and the adequacy of measures in systematically examine staff i
effect at operating plants, a.ic m recommend necessary changes to ensure the p
safe handling of-heavy loads. The guidelines proposed in NUREG-0612 include definition of safe-load paths, use of load handling procedures, training of crene operators, guidelines on slings and special lifting devices, periodic inspection and maintenance for the crane, as well as various alternatives.
7
.By Generic Letter's dated December 22, 1980, andFebruary3,1981(Generic Letter 81-07), all utilities were requested to evaluate their plants against the cuidance of HUREG-0612 and to provide their submittals in two parts: Phase 1 (six month response) and Phase II (nine month response). Phase 1 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..
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 11 of ' Cont,rol 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 HUREG-0612 (Phase 1), 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 11 was considered completed.
t F
'.While.not a requirement, NRC. encouraged the: implementation of any actions
. identified in Phase 11 regarding the handlirg of heavy. loads that were
. considered appropriate.
IMPLEMENTATION.AND. STATUS.
SUMMARY
'(PLANT SPECIFIC):
.in Supplenent No. I to our Safety Evaluation Report, we stated'that for all applicants-for an operating license are required to implerent the interim actions specified in our December 22, 1980 generic letter prior to the final implementation:of-the NUREG-0612 guidelines.
Ceco has implemented these
= interim actions; and therefore, meets our reovirements.for operating licenses.
However, we conditioned the operating license for Unit 2 to indicate'that prior to startup after the first refueling outape, the licensee shall incorporate:the NRC staff's requirements after completion of an acceptable review regarding the-guidelines of Sections 5.1.2-through 5.1.6 of NUREG-0612 (Phase 11 - nine month responses to the NRC generic letter dated December 22, 1980).
By letter dated 3/1?/85 (A. Schwencer to D. L. Farrar) this. issued was considered to be closed.
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t LaSalle, Units l'and 2
REFERENCES:
A-36 1.;
REQUIRFMENT. DOCUMENTS:
fTITLE-NUDOCS NO.=
DATE Letter, Darrell G. Eisenhut, NRC, to all licensees, applicants for OLs and holders of CPS transmitting NUREG-0612 and staff positions 12/22/80' Generic Letter 85-11, Hugh L.
Thompson, NRC, to all licensees for
' Operating Reactors, " Completion of: Phase 11 of ' Control of Heavy Loads at Nuclear Power Plants'
'NUREG-0612"'
06/28/85 2.
IMPLEMENTATION DOCUMENTS:
TITLE-NUDOCS NO.
DATE NUREG-0519 SER 8103230641 3/81-NUREG-0519 SER 8308290385 3/83 Supplement #5 page 9-3 3.
VERIFICATION. DOCUMENTS:
TITLE-
- NUDOCS NO.
DATE NRC letter A. Schwencer 8503190225 3/12/85 to D. L. Farrar, Heavy Loads l*
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' PLANT LaSalle..UnitsLl'.and 2 DOCKETN0(S).
50-373/374 PROJECT MANAGERT Paul C.-Shemanski TECHNICAL CONTACT J. Kudrick-..
USl NO. A-39 TITLE Determination of SRV Pool Dynamic Loads and
-Temperature Limits.
- MPA NO.
.10 5.
i ISSUES
SUMMARY
-This USI was resolved with the publication of Standard Review Plan (SRP)
Section 6.P.1.1.C. in October 1982. In addition, NUREGs 0763, 0783 and 0802 were issued for Mark I, Mark II, and Mark III containments, respectively.
BWP plants are equipped with safety / relief valves (SRVs) to protect the reactor
- from overpressurization. Plant operational transients, such as turbine trips, will actuate the SRV. Once the SRV opens, the air column within the partially submerged discharge line is compressed by the high-pressure steam released from the reactor.- The compressed air discharged into the suppression pool produces high-pressure bubbles. Oscillatory expansion and contraction of these bubbles create hydrodynamic loads on the containment structures, piping, and equipment 1nside containment.
HUREG-0802 presents the results of the staff's evaluation of SRV loads. The evaluation, however, is limited-to the quencher devices used in Mark 11 and 111 containments. With respect to Mark I containmen.ts, the SRV acceptance criteria are presented in HUREG-0661, " Safety Evaluation Report. Mark 1-Containment and Long-Term Program," and are dealt with as part of USl A-7.
SRP Section 6.2.1.1.C addresses the applicable review criteria, since all Mark 11 and III containment designs are understood to have completed their operating license (OL) reviews subseouent to resolution of this USI and reflection of the resolution in the SRP.
IMPLEMENTATION'AMD STATUS
SUMMARY
.(PLANT SPECIFIC):
.in a-letter from C. W. Schroeder to H. R. Denton dated September 13, 1983, Ceco provided a report entitled, " Evaluation of NUREG-0808 Load Definition for LaSalle County' Station - Units 1 and 2".
Based on the results of the assessment performed by Ceco in appendices H and I of the Design Assessnent Report and on the results of the additional reevaluation reported in the September 13, 1983 letter, the NRC staff concluded that Ceco has satisfactorily demonstrated that the piping and supports in the LaSalle Facility have been adequately designed to withstand the suppression pool hydrodynamic loads associated with the BWR Mark 11 containment design.
- Thus, the confirmatory item identified in Section 3.9.3.1 of Supplement No. 5 of the
- Safety Evaluation Report and the license condition 2.C(16) as specified in the Unit 1 license associated with the ability of the LaSalle piping systems to accommodate steam condensation oscillation and chugging loads per NUREG-0808 is considered to-be closed.
i:
i PEFERENCES:-
LaSalle, Units 1 and 2 A-39 1.
RE001REMENT DOCUMENTS:
TITLE NUDOCS.NO.
DATE SRP 6.2.1.1.C. Pressure Suppression i
Type BWR_ Containments i
NUREG-0802, " Safety / Relief Valve Quencher Loads:
Evaluation for EWR Mark 11 and'111 Containments, Generic Technical Activity A-39"
. 1982
.NUREG-0661, " Safety Evaluation Report -
7/80 Mark 1 Lono Term Program" L2.
IMPLEMENTATION DOCUMENTS:
TITLE NUDOCS NO.
DATE NUREG-0519 SER 8312220175 11/83
~ Supplement #6, page 3-1 i.
3.
VERIFICATION DOCUMENTS:
TITLE NUDOCS NO.
DATE l-l L
- The applicable SRP revision number would depend on the date of the evaluation l-for esco -specific plant.
l-e e
S
E
- PLAN _T LaSalle; Units l' and 2 DOCKET N0(S).
50-373/374 g
PROJECT MANAGER Paul C. Shemanski TECHNICAL CONTACT H. Ashar l
US1 NO. A 40 TITLE Seismic Desion Criteria MPA NO, TAC N05.
- ISSUES
SUMMARY
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 s
- NUREG-1233, " Regulatory Analysis for USI A-40," issued in September 1989, 1
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 curront requirements and found acceptable.
For tanks in plants reviewed during 1980-1984, the staff' identified four plant sites (six units) that were not explicitly reviewed to current requirements. The four plants (Callaway 1/2, Wolf Creek, Shearon Harris 1, and Watts Bar 1/2) are being handled on a plant-specific basis.
' USI A-40 originated in 1977. The basic objectives were (a) to study the seismic design criteria, (b) to quantify the conservatism associated with the criteria,and.(c).torecommendmodificationstotheStandardReviewPlan(SPP)if changes are justified. ' Lawrence Livermore National Laboratory (LLNL) completed the study and. published its findings in NUREG/CR-1161, " Recommended Revisions s
to USNRC - Seismic Design Criteria," dated May 1980.
The report recommended specific changes to the Standard Review Plan (SRP).
NRC staff reviewed the
+
- report and developed some other changes that would reflect the present state of seismic design practices. The resulting SRP changes were issued for public
. comment in June 1988, and the final SRP changes are to be published in October 1989.
- The majbr SRP changes consist of (a) clarification of development of site specific spectra, (b)-justification for use of single synthetic time-history by.
power spectral density function, (c) locatinn and reductions of input ground motion for soil structure interaction, and (d)' design of above-ground vertical tanks. Except for item (d), these items do not constitute any additional requirements for current' licenses and applications, and thus, no backfitting is
'being required for these items. However, the revised provisions could be used for margin studies and reevaluations or individual plant examination for j
external events (IPEEE).
-The participant utilities in the Seismic Qualification Utility Group (SQUG) agreed to implement the changed criteri6 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 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.
IMPLEMENTATION AND STATUS -
SUMMARY
(PLANT SPECIFIC):
Resolved by NUREG-1933 9/89. No additional requirements for LaSalle County Station.
+4 m
REFERENCES:
LaSalle, Units l'and 2-A-40 1.
PFOUIREMENT DOCUMENTS:
TITLE-NUDOCS NO.
DATE
' Regulatory Analysis for NUREG-1233 Sept. 1989 USI A-40 Recommendations for Resolution NUREG/CR-5347 June 1989
~ f Public Corrments on USI A-40 o
e Standard Review Plan
'NUREG-0800 To be issued Sections 2.5.2, 3.7.1, 3.7.2, 3.7.3 (Revision 2)
Response of Seismic NUREG/CR-4776 Feb. 1987 Category 1 Tanks to Earthouake-Excitation Engineering Characteri-NUPEG/CR-3805 Feb.-Aug. 1986 zation of Ground Motion, Vols. 3,4,5 f.'-
Prnceedings of the NUREG/CR-0054 June 1986 Workshop on Soil-Structure Interaction.
j Value Impact Assessment NUREG/CR-3480 Aug. 1984
.f or Seismic Design Criteria Seismic Hazard Analysis NUREG/CR-1582 Oct. 1981 i
f.pplication of Methodology, 1
Fesults and Sensitivity 1
Studies, Vol. 4 Recommerded Revision'to NUREG/CR-1161 May 1980 Nuclear Regulatory Commission
. Seismic Design Criteria 1
Power Spectral Density Functions NUREG/CR-3509 June 1988 l
Compatible with NRC R.G. 1.60 l
L Response Spectra j
1 l~
2.
I!1PLEMENTAT10N DOCUMENTS:
TITLE NUDOCS NO.
DATE l
Recuest for Information Letters Docket Nos.
May 1989 to Owner's of Callaway 182, Wolf 483, 486, 482, L
Creek 1, Shearon Harris 1, Watts 400, 390, 391 Bar 1&2 1
3.
VERIFICATION DOCUMENTS:
L TITLE NUDOCS N0.
DATE l
Y PLANT-LaSalle, Units 1 and 2' DOCKETN0(S).
50-373/374 m
PROJECT MANAGER Paul C. Shemanski TECHNICAL CONTACT W. Koo
~ USI NO. A-42 TITLE Pipe Cracks in Boiling Water Reactors MPA NO..
TAC N05, 69141/69142 4
ISSUES
SUMMARY
This USI.was resolved in February 1981 with the publication of NUREG-0313, Revision 1,." Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping." That NUREG document was issued to all holders of BWR operating licenses or construction permits and to all applicants for BWR operating licenses. The staff established MPA B-05 for
. implementation of the resolution at operating plants.
Pipes have cracked in the heat-affected zones of welds in primary. system piping in BWRs since mid-1960. These cracks have occurred mainly in Type 304 stainless steel, which is the type used in most operating BWRs. The major problem is i
recognized to be intergranular stress corrosion cracking (IGSCC) of austenitic stainless steci components that have been made susceptible to this failure by being_" sensitized," either by post-weld heat treatment or by sensitization of a narrow heat affected zone near welds.
" Safe ends" that have been highly sensitized by furnace heat treatment while
-attached to vessels during fabrication were found to be susceptible to IGSCC in the late 1960s. Most of the furnace-sensitized safe ends in older plants have
-been removed or clad with a protective material, and only a few BWRs still have.
furnace-sansitized safe ends in use. Most of these, however, are in smaller diameter lines.
Cracks reported before 1975 occurred primarily in 4-inch-diameter recirculation loop bypass. lines and in 10-inch-diameter core spray lines.
Cracking is most often detected during inservice-inspections using ultrasonic test techniques.
Some piping cracks have been discovered as a result of primary coolant -leaks.
1 L
NUREG-0313, Revision 1 provided the NRC staff's revised acceptable methods for reducing the IGSCC susceptibility of BWR code class 1, ?, and 3 pressure boundary' piping of sizes-identified above and safe ends.
In addition, it provided the reouirements for augmented inservice inspection of piping with nonconforming materials.
p l-As a result of further IGSCC degradations in larger piping, the staff provided l
licensees with additional requirements in several NRC communications (i.e.,
L
~ Bulletins 82-03,83-2,and84-11).
The long-term resolution of IGSCC in BWR
-piping (includingthescopeofA-42)wasprovidedinNUREG-0313, Revision 2 l
which was transmitted to all holders of BWR operating licenses via Generic Letter 88-01.
L
-IMPLEMENTATION AND STATUS
SUMMARY
(PLANT SPECIFIC):
By letter dated November 29, 1989 Commonwealth Edison indicated USI A-42 was o
completed on 7/29/88.
Inspection Reports IR 373/85035 and IR 374/87002 were issued for Units 1 and 2, respectively. GL 81-04 transmitted NUREG 0313, Rev.
I'_to all-BWR licensees and provided NRC requirements pertaining to resolution H
1 of USI:A-42. -The requirements in GL 81-04 were superseded by the requ'irements in GL 84-11'(MPA B-84). MPA B-84 was closed as being fully implemented on all i
.BWPts as February 12. 1988.e-The basis was the memorandum'to Thomas E. Murley from' Frank J. Miraglia and James H. Sniezek dated February 12, 1988.4 The-requirements in GL 84-11 were in turn superseded by the requirements-in GL
- 88-01(MPAB-97).
For the 24 BWRs that were operating when GL 81-04 was issued, the implementa-tion document is the letter to the licensee transmitting the staff's evaluation of their response to GL 81-04.
For the NT0Ls at the time, the implementation document is the SER or SSER in which the staff evaluated the applicants confor-mance to the requirements of NUREG-0313, Rev. 1.
P 4
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P.EFERENCES:
LaSalle, Units-1 and 2 A-42 1.-
REOUIREMENT-DOCUMENTS:
TITLE-NUDOCS NO.
DATE
~HUREG-0313, Revision 1, " Technical Report on Material Selection and Processing Guidelines for BWR Coolant' Pressure Boundary Piping,"-
07/80 Generic Letter 81-04, "Implemen-tation of NUREG-0313, Rev. 1~for Selection and Processing Guidelines for BWR Coolant Pressure-Boundary.
Piping (Generic Task A-42)"-
2.
IMPLEMENTATION DOCUMENTS:
TITLE NUDOCS NO.
DATE Letter W. Morgan to 8808090125 7/29/88 NRC -' Response to GL 88.
n 3.
VERIFICATION DOCUMENTS:
TITLE NUDOCS NO.
DATE inspection Reports IR 373/85035 8663270073 1985 1R'374/87002 Not Found.
1987 i.
i e
03 n,
j b
PLANTlLaSalle. Units 1and2-DOCKETN0(S).~50-373/374 1
PROJECT MANAGER Paul'C.-Shemanski' TECHNICAL CONTACT A. Serkir-
)
MPA NO.-
TAC HOS, ISSUES
SUMMARY
+
The resolution of this'US1 was presented to the Commission in October 1985 in
- SECY-85-349. NUREG-0897, _ Revision 1, " Containment Emergency Sump Performance,"
-presents the results of the staff's technical findings. These findings estab-
=11shed a need to revise current licensing guidance on these matters.
- Revision 0 and Standard Review Plan Section 6.2.2, " Containment Heat Removal Systems" were revised to reflect this new guidance.
No licensee actions were required.
Initially, an issue existed concerning the availability of adequate recircula -
tion cooling witer following a loss-of-coolant accident (LOCA) when long-term 1
. recirculation of cooling water from the PWP, containment sump, or the BWR residual heat removal system (RHR) suction intake, must be initiated and 4
maintained to prevent core melt.
-The technical concerns evaluated under USI A-43 were:
(a) post-LOCAadverse conditions resulting from potential vortex formation and air ingestion and subsequent pump failure, (b) blockage of sump screens with LOCA generated insulation debris causing inadequate net positive suction head (NPSH) on pumps,
- and (c) RHR and containment spray pumps inoperability due to possible air, debris, or particulate ingestion on pump
- seal and bearing systems.
This~ revised guidance applies only to future' construction permits, preliminary design approvals, final design approvals, standardized designs, and applica-tions for licenses to manufacture. The staff performed a regulatcry analysis to determine if this new guidance should be applied-to operating plants. The results of this analysis were reported in NUREG-0869 Revision 1 "USI A-43 Regulatory Analysis," issued in October 1985. The staff concluded that the
- regulatory analysis does not support any new generic requirements for present licensees to perform debris assessments.
IMPLEMENTATION AND STATUS
SUMMARY
(PLANT SPECIFIC):
NUREG-0519 SER, 3/81 h
Resolution in SECY 85-349 required no changes.
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I 1
il l
L
REFERENCES:
LaSalle, Unitsel and 2~
A.43:
RE0lllREMENT DOCUMENTS j
-1.
i.
TITLE NUDOCS NO.-
DATE
)
o NUREG-0869, Rev. 1, "USI 10/85 A-43 Regulatory Analysis"
.itVREG-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-s L
- 3.
VERIFICATION -DOCUMENTS:
i TITLE NUDOCS NO.
DATE i
5 9
r e
h LPLANTE LaSalle. Units 1 and 2 DOCKET N0(S),
50-373/374 1 PROJECT MANAGER' Paul C. Shemenski TECHNICAL CONTACT R. Jones USI NO.
A-45~
. TITLE Shutdown Decay Heat P.emoval Reovirements HPA NO.
TAC N05.
ISSUES SllMMARY:
o USl A-45 was resolved by SECY 88-260, " Shutdown Decay Heat Removal Requirements (USI-A-45),". issued September 13,1988, without imposing any.new licensing requirements other than the Individual-Plant Examination-(IPE), as described below. At the same time the staff issued NUREG-1289, " Regulatory and Backfit Analysis: USI A-45."
Since all of the significant USI A-45 results have been.
f ound to be highly plant specific, the Commission decided it was not appropriate to propose a single generic corrective action to be applied uniformly to all plants.
~
t 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 terned the IPE. The reason for this examina-tion is.to identify any plant-specific vulnerabilities to severe accidents.
The IPE analysis intends to examine ano understand the plant emergency pro-
'cedures, design, operations, maintenance, and surveillance, in order to identify vulnerabilities.. The analysis will examine both the decay heat removal' systems and those systems used for other related functions. This
. includes CE plants without power-operated relief valves.
NRC has decided to subsume A-45 into the IPE program as the most effective way
- of achieving resolution of specific plant concerns associated with A-45.
1MPLEt'ENTA110N AND STATUS
SUMMARY
(PLANT SPECIFIC):
S CY 88-260 transferred action for this issue to the IPE Program; hence, there are no additional requirements for LaSalle County Station with regard to USI.
A-45 at this time..
't 1
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e.
4e l
i
REFERENCES:
LaSalle, Units 1 and 2 A-45 1.
REQUIREMENT DOCUMENTS 1
TITLE NUDOCS NO.
DATE Federal Register Notice "10 CFR Part-50,, Shutdown Decay Heat.
Removal Requirements" NUREG/CR-5230 " Shutdown Decay Heat April 1989 1
Removal' Analysis: Plant Case Studies and Special Issues.Surnary Report" NUREG-1289 " Regulatory and Backfit
-11/30/88 Analysis for the Resolution of USI A-45" SECY-88-260" Shutdown'DecayHeat 09/13/88 Removal Requirements:
2.
IMPLEMENTATION DOCUMENTS:
TITLE NUDOCS NO.
DATE
.- h-t b
3.
VERIFICATION DOCUMENTS:
TITLE NUDOCS NO.-
DATE i
e 4
t i
PLANT' LaSalle, Units 1-and 2
.DOCKETN0(S),
50-373/374 Paul C. Shemanski-TECHNICAL CONTACT P. Y. Chen PROJECT MANAGER USI NO. A-46 TITLE Seismic Qualification of Equipment in Operating Plants g
t itPA'NO. B-105 TAC N05.
_1SSUES SUMMAR_Y_:
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 Qualification Utility Group-(SQUG) 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)seismicsystemsinteration.-
As an outgrowth of.the Systematic Evaluation Program (SEP), the need was identified for reassessing design criteria and methods for.the seismic quali-1 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 SQUG and the steff on a generic basis'before individual member utilities proceed with
. plant-specific implementation.
IMPLEMENTATION AND STATUS
SUMMARY
(PLANT SPECIFIC):
The Generic. Implementation' Procedure (GIP), Revision 0, was submitted by SQUG on June 3, 1988. The staff issued a Generic Safety Evaluation (SE) on July 29, 1988 endorsing much of the GIP but with about 70 open items to be resolved.
- After a series of meetings, SQUG submitted Revision 1 to the GIP on December 23,.1988. Supplemental information was submitted by SQUG on March 17, 1989.
The staff has prepared a supplemental SE for GIP, Revision 1 and has submitted
-it to the CPGR for review. The target date for issuance of the supplemental
.SE was November 1989. An additional supplement is scheduled for June 1990 and overall closeout of implementation projected for 1993.
-l i
r
REFERENCES:
L LaSalle, Units 1 and 2 A-46 1.
- RE001REttENT DOCUMENTS:
TITLE NUDOCS NO.
DATE Generid Letter 87-02, "Verifi-cation of Seismic Adequacy of Mechanical and Electric Equipmer.t in Operating Reactors" 02/19/87 NUREG-1211. " Regulatory Analysis for Resolution of Unresolved Safety Issues A-46..."
02/87 i
NUREG-1030, " Seismic Qualification of Equipment in Operating Plants, Unresolved Safety Issue A-46" 02/87 Letter attached with " Generic Safety Evaluation Report on SQUG
. GIP, Revision 0," from L. Shao
.(NRC).to Neil Smith-(SQUG) 07/29/88 2.-
IMPLEMENTATION DOCUMENTS:
TITLE NUDOCS NO.
DATE
'" Generic Implementation Procedure (GlP for Seismic Verification of fluclear Plant Equipment," Revision 0 06/88'
" Generic Implementation Procedure
-(GlP)forSeismicVerificationof Nuclear Plant Equipment," Revision 1 12/88 3.
VERIFICATION DOCUMENTS:
TITLE NUDOCS NO.
DATE l
s 1
m 3
PLANT- 'L'aSalle. Units.1-and 2 DOCKET N0(S). 50 373/374 PROJECT MANAGER Paul C. Shemposki TECHNICAL CONTACT J. Mauck.
USl NO. A-47 TITLE Safety Implication of Control Systems in LWR Nuclear Power Plants MPA NO.
TAC N05.
1SSUES
SUMMARY
1US1.A-47 was resolved September 20, 1989, with the publication of Generic Letter (GL)88-19.'
The generic: letter states:
~
"The staff has concluded that ell 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, l
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 ennsistent with the criteria and the risk considera-tions of_ the Commission Interim Policy Statement on Technical-Specification Improvement.
In addition, the staff recommends that all:BWR recipients reassess and modify, if needed, their operating procedures and operator training to assure that the operators can mitigate reactor vessel overfill events that may occur via the condensate booster pumps during reduced system pressure-operation."
.Also, page 2 of the generic letter provides for additional actions for CE_and B&W plants. The generic letter provides amplifying guidance for licensees.
The generic letter. requires that licensees provide NRC with their schedule'and commitments within 180 days of the letter's date.
The implementation schedule for actions on which commitments are made should be prior to startup after the first'. refueling outage, but no later than-the second refueling outage, beginning 9 months after receipt of the letter.
IMPLEltENTATION AND STATUS SLTftARY (PLANT. SPECIFIC):
Genetic Letter 89-19 was recently issued to Commonwealth Edison. A response to the September 20, 1989 generic letter is due by March 19, 1990.
REFEREllCES:'
LaSalle, Units 1 and 2 A-47
-l
- c..
l
- 1..
REQUIREMENT 00ClNENTS-
. T_1TLE NUDOCS NO.
DATE Generic Letter.89-19 09/20/89-
" Request for Action Related' i
to Resolution of USI A-47" J
'NUREG-1217 " Evaluation of Safety June-1989 Implications of Control Systems in LWR Nuclear Power Plants" NUREG-1218 " Regulatory Analysis July 1989 for Resolution of USI A-47" 2.
IMPLEMENTATION DOCUMENTS:
TITLE-NUDOCS NO.
,DATE 3.
VERIFICATION DOCUMENTS:
I TITLE NUDOCS NO.
DATE k
9
1
.5 PLAtlT LLaSalle. Units 1 and 2-DOCKET N0(S). 50-373/374 PROJECT itANAGER Paul C. Shemanski TECHNICAL CONTACT-J. Kucrick USI NO..A-48 TITLE Hydrogen Control Measures and Effects of Hydrogen Burns on Safety Eouipment tiPA NO.
TAC NOS.
1SSUES
SUMMARY
The NRC staff concluded April 19, 1989, that USI A-48 is resolved,-as stated in
~ SECY 89-122.
a I
USI A-48 was-initiated as a result of the large amount.of hydrogen generated
.and burned within containment during the Three Mile Island (TMI) accident.
This issue covers hydrogen control measures for recoverable degraded core U
accidents for all BWRs and those PWRs with ice condenser containments.
Extensive research in this area has. led to significantcrevision of'the Com-mission's hydrngen control regulations, given in 10 CFR 50.44, published December 2, 1981.
10 CFR 50.44 requires inerting of BWR Mark'I and Mark 11 containments as a
- method for hydrogen control. The BWR Mark I and Mark 11 reactor containments have operated for a number of years with an inerted atmosphere (by addition of an inert gas, such as nitrocen) which effectively precludes combustion of any hydrogen generated.
USI A-48 with respcet to BWR Mark I and 11 containments is not only. resolved but understood to be fully implemented in the affected plants.:
The rule for BWRs with Mark 111 containments and PWRs with ice condenser containments was published on January 25, 1985. The rule required that these plants be provided with a means for controlling the quantity of hydrogen t
produced, but did not specify the control method.
In' addition,'the task actinn plan for USI A 48 provided for plant-specific reviews of lead _ plants for reactors with Mark Ill and ice condenser containments.
Sequoyah was chosen as the lead plant for ice condenser containments and Grand Gulf for Mark III containments. Both of the lead plant licensees chose to install igniter-type systems which would burn the hydrogen before it reached threatening concentra-tions within the containment.
Final design igniter systems have been installed
- not only-in both lead plants, Sequoyah and Grand Gulf, but in all other ice condenser and Mark III plants as well. The staff's safety evaluations of the final analyses required to be submitted by these licensees by the rule are scheduled for completion in 1989.
Large dry PWR containments were excluded from USI A-48 because they have a greater ebility to accommodate the large quantities of hydrogen associated with a recoverable decraded core accident than the smaller Mark I, II, III and ice condenser containments. However, this issue has continued-to be considered and, in 1989, hydrogen control for large dry PWR containments was identified es a high-priority Generic Issue (GI) 121. The resolution of G1 121 is being actively pursued in close coordination with more recent research findings.
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' IMPLEMENTATION AND STATUS
SUMMARY
-(PLANT SPECIFIC):
Section 6.2.5-of NUREG-0519, SER (3/81) concluded-that the Combustible Gas
' Control System which includes the Containment Hydrogen Recombined is.
acceptable.- SECY 89-122 further resolved USI A-48,
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9
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LaSalle, Units 1 and 2 A-48
_15$UESStEttARY(CONT.I:
i:
The NRC staff has concluded that USI A-48 is resolved as-stated in SECY L
89-122.
If interested, the report should be consulted for further details regarding the relationship of A-48 to other ongoing hydrogen activities.
L 1MPLEMENTAT10N AND STATUS
SUMMARY
(PLANT SPECIFIC):
L REFER.ENCES:
l-l 1.
REOUIREMENT 00CtHEN15:
TITLE NUDOCS NO.
DATE 10 CFR S0.44, Standards for 12/81 3' '.
Comt.ustible Gas System 'n 1.ight4ftter-Cooled Power
^
Reactors SECY-M 10Z, Pesolution of
,4 US1 A.49, "Hyd sget. Control A "z Meest, ret 6 rid Ef f t. cts of T
Hyttogen BJens on S6fety 1-Equi peer.:
- 04/19/89 2.
ItiPLEtdFitTATION Of'ClNENYS:
TITLE NUDOCS NO.
DATE 3.
VERIFICATION DOCUMENTS:
TITLE NUDOCS NO.
DATE
(-
1 l.
e