ML20033E232
| ML20033E232 | |
| Person / Time | |
|---|---|
| Site: | Dresden, Quad Cities  |
| Issue date: | 02/22/1990 |
| From: | Ross T 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-EL, REF-GTECI-SC, REF-GTECI-SY, TASK-A-09, TASK-A-44, TASK-A-46, TASK-A-9, TASK-OR GL-89-21, NUDOCS 9003090436 | |
| Download: ML20033E232 (2) | |
Text
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UNITED STATES NUCLEAR REGULATORY COMMISSION
-l wassmotow. o. c. mos February 22, 1990 l
Docket Nos. 50-254 and 50-265 j
PEMORANDUM FOR:
File FROM:
Thierry M. Ross, Pro.iect Manager l
Project Directorate !!!-2
)
Division of Reactor Projects - 111, IV, Y and Special Projects
SUBJECT:
STATUS OF IMPLEMENTATION OF UNPESOLVED SAFETY ISSUES AT CUAD CITIFS NUCLEAR POWER STATION l
The current implementation status of unresolved safety. issues (USIs) at the Quad Cities Nuclear Power Station (QCNPS) 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 sumary for each USI applicable to this fec111ty. This status sumary is based upon the licensee's response to the Ceneric Letter, discussior.s with the licensee, and my review of available NRC records and information. is a copy of the staff's data base printout for Qua.d cities.
It reflects the staff's assessment of US1 implementation for all 27 bils.
It is based on review of the licensee's response to Generic letter 89-21, and evaluation by project managers, the UST team, and NRR technical staff.
Four of the USI's which are discussed below are incomplete. The schedule for completion of these USI's is acceptable since all these items are either pending staff resolution or review, under discussion between the staff end licensee, or pending the licensee's responte which is not yet due.
l (1) Anticipated Transients Without Scram (A-09) - All ATWS related j
nodifications tcr both units were essentially complete by October 1987, Only the Technical Specifications (TS) for Alternate Rod injection (ARI) i Edison Company (Ceco)p Trip (RPT) remains to be done.
and Recirculation Pum Comonwealth plans to submit a TS license amendment by June l
1990.
However, the staff is concerned that the AR1/RPT design does not exhibit sufficient diversity from the Reactor Protection System.
This point of contention is under appeal by the BWR Owners Group with the Comission, fS A3@ f # K3 6 YSr),X/3 t
file February 22, 1990 (2) Station Blackout (SBO) A aa - The staff and CECO are still deliber-l ating over Ceco's initial response to 10 CFR 50.63. A supplemental response is expected from CECO following a second working meeting held l
December 20, 1989.
Both the staff and Ceco expect that the SB0 rule will t
necessitate station modifications. Current discussions and review efforts continue within the scope of 10 CFR 50.63.
(3) Seismic Qualification Of Equipment (A-46) - NRC staff is still reviewing Revision 1 to the Generic Implementation Procedure (GIP) submitted by the Seismic Qualification Utility Group (SOUG). Ceco is waiting for staff approval of the SQUG GIP before committing to an implementation schedule.
The target date is early 1990 for issuance of a supplemental SER. Assuming this happens, Ceco predicts final implemen-tation could be accomplished by the end of 1992 for Unit 1 and 1993 for Unit 2.
(4) Safety inspection of Control Systems-(A-47) - Ceco's response to GL 89-19 is not due until March 31, 1989.
Appropriate NRR technical branches have also reviewed the USI status summary and this memo.
(
Original sioned by Leonard 01shan j,9.,t. jht,.
F-r Thierry M. Ross, Project Manager Project Directorate 111-2 Division of Reactor Projects - 111, IV, V and ".pecial Projects
Enclosures:
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DOCUMENT NAME: [QuadCitiesUSl]
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December 8, 1989 l
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Dr. Thomas E. Murley, Director Office of Nuclear Reactor Regulation
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U.S. Nuclsar Regulatory Commission i
i Washington, DC 20555 l
t
Subject:
Quad Cities Nuclear Power Station Units 1 and 2 i
Response to Generic Letter 89 Implementation Status of USI Requirement i
Nic Decket Nea. 50-254 and 50-263 t
Reference:
Letter from J.G. Partlow to Licensees dated i
October 19, 1989 (Received October 30, 1989).
i
?
Dr..Murley i
Enclosed is the information requested by the subject Generic Letter concerning the implementation status of Unresolved Safety Issues (USI) requirements relevant to Quad Cities Station.
I The status for each relevant USI has been determined to the best of our ability in the allowed time period (a limited extension was granted by the NRR Project Manager).. Common ~nealth Edison believes the enclosed inforn tien
'=
accurately reflects the Quad Cities USI implementation status at this t! Joe.
However, should additional information become available which f.iffets fron that enclosed, the revisions will be preeptly ecmunicated to +.ht NrJL Projkct Manager.
Please contact this office if you should need further information.
Very truly yours, R. Stols Nuclear Licensing Administrator in &-$lhalfu W~Wp--
y Attachment cc A.B. Davis - Regional Administrator, RIII 5.N. Ross - Project Manager, NRR R.L. Higgins - Resident Inspector Quad Cities
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m EANEIY IEEME Ftm MIM A FIML MICAL RESDERTId4 EAS EEEE AIX1ENED gh"ZLCII1ES WITE 1 ^
2 USI/MFA NWIBER TIT 11 SIATUSlRAIE REllARES A-1 Water E - r C
12/80 Eigh reactor water level feed pump trip is lastalled. The training upgrade repirements of NUREG 0737, Item I.A.2.3 were incorporated into the training program prior to I
accreditation per J. Abel to D. Eisenhut letter dated 12/15/80. Quad cities wee accredited in 1985.
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A-2/
Asymetric Blowdown N/A MPA D-10 Loads on Reactor Primary Coolant Systems I
i A-3 Westinghouse Steam N/A Generator Tube Integrity I
A-4 DE Steam Generator Tube N/A Integrity I
A-5 B&W Steam Generator N/A Tube Integrity i
4 C - COMFLETE NC - NO CHANGES NECESSARY l
NA - NOT APPLICABLE f
I - INCOMPLETE E - EVALUATING ACTIONS REQUIRED l
2 0421T:2 1
?
USI/MPA MRIBER TITIE SIAIUSIDATE REM 6tKS A-6 Mark I Containment C
12/77 Superseded by USI A-7 Short-Term Program A-7/
Mark I long-Term C
10/87 2/15/86 letter from J.A. Zwolinski to D.L.
D-01 Program Farrar provides SERs for Fool Dynamic Imeds Review and Structural Review. MRC Request regarding Tech Spec changes required was responded to in J.R. Wojnerowski letter to N.R..
Denton dated 5/16/86. SER for drywell vacuum breakers (TAC's 57150, 57151, MFA D-20) was issued 11/26/86 and was not considered part of long term program (addressed by CL 83-08).
Modification installation work was completed in 12/82 (U-1) and 10/87 (U-2).
A-8 Mark II Containment N/A Fool Dynamic leads A-9 Anticipated Transients I
6/90*
Modifications for ARI, RFT were completed 4
Without Scram U-1 3/85, and U-2, 2/86; SIES U-1,10/87 and U-2 1/87 Tech Spec change for SLCS was completed U-1, 3/88 and U-2,12/86 (TAC's 66621 l
and 63367). RFT and ARI Tech Spec change is to be submitted to NRC by 6/90.
B.L. Siegel letter to R.E. Bliss dated 11/8/88 (TAC's 59089, 59090, 59132, and 59133) states that "the ARI and RPT designs for Dresden and Quad Cities units I
satisfy the requirements of the ATWS Rule except for diversity." 12/15/88 letter from J.A. Silady to T.E. Murley presented BWROC and l
CECO positions on this issue.
I
- Projected completion does not include NRR's response to 12/88 l tt e
er.
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USI/MPA HUMBER TIT 11 STAIUSil&TE REHARES A-10/
11/83 Feedwater sparger nozzle temperature MFA B-25 Cracking 4
modifleations were completed 1/83 (U-1) and 6/80 (U-2).
A-11 Reactor Vessel Meterial I 6/90 Tech. Spec. assendment to incorporate radiation Toughness l.
ef fects into PT curves were submitted per R.G.
1.99.
Based upon the relatively low end-of-life fluence and the small expected 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 BWR vessels dropping below the 50 ft.-Ibn limit.
Bowever, CECO plans to approach the BWBOG to establish a generic program to develop basellae data for initial unirradiated charpy upper shelf values for older BWR vessels and provide the associated documentation. Based *upon recent discussions with GE, CECO espects a report to be issued in approximstely 6 months.
A-12
. Fracture Toughness of N/A Steam Generator and Reactor Coolant Fump
+
Supports A-17 Systems Interactions Ceco's response to GL 88-20 committed to en IFE program. Water intrusion and flooding from internal sources are part of the IFE.
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USI/MPA IHJtsBER TITLE STATUS /Il&IE REll4RES A-24/
' Qualification of Class C
11/85 Inspection Report (IR) 50-254(265)/87-011 stated MFA B-60 1E Safety +Related that proposed resolutions were found to be Equipment acceptable by the NRC in final EQ SER dated 7/19/84 All EQ modifications completed 11/85.
1 A-26/
Reactor Vessel Pressure N/A MFA B-04 Transient Protection l
1 j
j A-31 Residual Beat Removal N/A Shutdown Requirements A-36/
Control of Beavy Imads C
6/83 D. B. Vassalo letter to D. L. Farrar dated i
C-10, Near Spent Fuel 6/17/83 provides SER for Phase I.
C-15 Phase I Phase II C
6/85 j
Generic Letter 85-11 completes Phase II of NUREG-0612 based on improvements from l
j Phase I.
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A-39 Determination of SRV N/A A-7 addressed this for Mark I.
i Pool Dynamic Loads and Pressure Transients A-40 Seismic Design I*
Waiting final ISR acceptance of EPRI -
Criteria Seismicity owner's Group position on Items 1, l
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.
- Completion dated will be based on receipt of NRR acceptance.
0421T:5
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USI/MPA IBRIBER TITLE STAIlls/IRIE REIS&RES A-42/
Pipe Cracks in Bolling I
12/90 MFA B-05 Water Reactors Ieakage monitoring Tech. Spec. revision is being prepared and will be submitted as part of Tech. Spec. Improvement.. Tech. Spec. require-ments for ISI program will be removed from the Tech. Specs. and included to the ISI program, i
following NRR review and approval of C.L. 88-01.
A-43 Containment Emergency N/A Sump Performance A-44 Station Blackout I
The 4/17/89 CECO response proposed the t:oping I
approach in combination with some hardware i
changes. During a working meeting held on i
10/4/89, there appears to be a difference of
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opinion between the staff and Ceco regarding pla'nt acceptable blackout duration capability.
Discussions are on-going.
j A-45 Shutdown Decay Beat C
Removal Requirements USI A-45 will be addressed by IFE as implemented by GL 88-20.
Quod Cities IFE is expected to be completed by 12/93.
A-46 Seismic Qualification I*
of Equipment in Waiting for staff approval of SQUG generic inspection plan and lesmance of the final SER.
Operating Plants SQUG walkdown will be conducted following staff approval of GIF. Walkdown schedules will be l
developed following ISC approval.
A-47 Safety Implication E
Response to GL 89-19 is due 3/90.
of Control Systems on Safety Equipment i
4
- Completion date will be based upon issuance of NRR acceptsace.
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0421T:6 l
7 USI/MPA IRRBER Ti m STATUS /DATE REMARES A-48 Eydrogen Control I*
l Measures and Effects W iting for M R to issue Staff's position os of Wydrogen Burns BUR Owner's Group response to cominnetible gas control.
A-49 Pressurized Therent N/A Shock
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- Completion date will be deter !ned upon issuance of NRR position.
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l PLANT Ouad Cities DOCKETN0(S).
50-254/265 PROJECT MANAGER Thierry Ross TECHNICAL CONTACT A. Serkir USI NO. A-1 TITLE Water Hamer MPA NO.
N/A TAC NOS.
N/A ISSUES
SUMMARY
This Unresolved Safety Issue (USI) was resolved in March 1984, with the publication of NUREG-0927 " Evaluation of Water Hamer in' Nuclear Power Plants
- Technical Findings Relevant to Unresolved Safety Issue A-1."
Also on March 15, 1984, the EDO sent the Comissioners SECY 84-119 titled, " Resolution of Unresolved Safety Issue A-1, Water Hemer."
In SECY 84-119, the staff concluded that the frequency and severity of water hamer occurrences had been significantly reduced through (a) incorporation of design features such as keep-full systems, vacuum breakers, J-tubes, void detection systems, and improved venting procedures; (b) proper design of feed-water valves and control systems; and (c) increased operator awareness and training. Therefore, the resolution of USI A-1 did not involve any hardware or design changes on existing plants, it did involve Standard Review Plan (SRP) changes (forward fits) and a comprehensive set of guidelines and criteria to evaluate and upgrade utility training procrams (per TM1 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 feeowater pump trip was in place or being installed. This was necessary because calculated values h6d postulated an IC failure by water hamer that opened a direct pathway to the environment.
ItiPLEMENTAT10M AND STATUS
SUMMARY
FOR OUAD CITIES:
No changes were made to plant systems or procedures, because resolution of USI A-) did not involve any additional requirements (i.e. hardware or procedures) for Quad Cities.
Consequently, status of implementation is "NC" (no changes required).
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REFERENCES:
Quad cities A-1 1.
RE001REMENT DOCUMENTS:
i i
TITLE NUDOCS NO.
DATE Letter from Denton to Utilities, 8403150310 03/05/84
" Notice of Issuance and i
Availability NUREG-0927 Rev.1, Safety.lssue A 1" l
r 2.
IMPLEMENTATION DOCUMENTS:
\\
TITLE NUDOCS NO.
DATE NUREG-0927 " Evaluation of Water 8306060413 05/31/83 Hammer in Nuclear Power Plants-Technical Findings Relevant to Unresolved Safety issue A-1" j
NUREG-0993 Rev. 1 8306060418 March 1984 t
" Regulatory Analysis for for USI A-1, Water Hammer" SRP Sections:
3.9.3, 3.9.4,
,5.4.6, 5.4.7. 6.3, 9.2.1, 9.2.2, 10.3, and 10.4.7 SECY-84-119. " Resolution 03/15/84 of Unresolved Safety A-1,
)
Water Hammer" 3.
VERIFICATION DOCUMENTS:
TITLE NUDOCS NO.
DATE Not Applicable 1
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7 t
l PLANT Qued Cities DOCKET N0(S). 50-254/265 l
PROJECT MANAGER Thierry Ross TECHNICAL CONTACT J. Kudrick USI NO. A-6 TITLE Mark 1 Containment Short Term Program MPA NO.
TAC NOS.
ISSUES SuttMARY:
This USI was resolved in December 1977 with the publication of NUREG-0408,
" Mark I Containment Short-Term Program Safety Evaluation Report."
The ob.iectives of the !! ark I short-term progran were:
(a) to examine the containment system of each BWR facility with a Mark I containment design to verify that it would maintain its integrity and functional capability when subjected to the most probable hydrodynamic loads induced by a postulated design-basis LOCA, and (b) to verify that licensed Mark I BUR 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.
i The NRC staff used a safety factor of et least two to failure for the weakest structural or mechanical component in the Mark I containment system in judging i
that containment integrity and functions would be assured under most probable design-basis LOCA-induced hydrodynanic loads.
As indicated in NUREG-04C8, the staff reautred full implementation of the l
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 objectives of the Short-Term. Program had been satisfied, thus documenting the basis for resolving this safety issue. This issue is considered complete for all affected BWRs.
JMPLEMENTATIONANDSTATUS
SUMMARY
FORQUADCITIES:
In conjunction with the Mark 1 Short Term Program, Commonwealth Edison Company (Ceco) performed a plant unique analysis for Quad Cities 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 further recommended full implementation of a drywell to suppression chamber differential pressure, in response, Ceco submitted a license amendment application (approved by the NRC 1
staff as license amendment no. 45) to require that Quad Cities maintains a 1.20 psid drywell - suppression chamber differential pressure. This will assure integrity of the suppression chamber when subjected to post-LOCA suppression
- pool hydrodynamic forces.
i NUREG-0408 anticipated that the action taken by licensees in response to the Mark i Short Term Program would be implemented as an interim measure. Ceco has decided, as part of the Mark I Long Term Program, to retain this feature permanently.
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REFERENCES:
Quad Cities A-6 1.
RE0UIREMENT DOCUMENTS:
TITLE NUDOCS NO.
DATE L
NUREG-0408, " Mark i Containment 12/77 Short Term Program Safety Evaluation Report" (See Table I-2 for letters to BWR licensees requesting action) 2.
IMPLEMENTATION DOCUMENTS:
TITLE NUDOCS NO.
DATE A. Quad Cities Station Units 1 and 2' 08/76 "Short Term Program Plant Unique Torus Support and Attached Piping Analysis" 8.
License Amendment No. 46' 06/07/78 for DPR-29 and 30 3.
VERIFICATION DOCUMENTS TITLE NUDOCS NO.
DATE 4
0 0
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l PLANT Ouad Cities DOCKET N0(S).
50-254/265 i
PROJECT MANAGER Thierry Ross TECHNICAL CONTACT J. rudrick j
US! NO. A-7 TITLE MarkIlongTermProiram MPA NO. 0-01 TAC NOS. 7947/7949 ISSUES SUMPARY:
f i
This USI was resolved in August 1982 with the publication of Supplement I to NUREG-0661, " Safety Evaluation Report, Mark 1 Containment Leng-Term Program" and Standard Review Plan Section 6.2.1.1.C.
For operating BWRs, MPA D-01 was established for implementation purposes.
i The focus of this USI was the suppression pool hydrodynamic loads, associated with a postulated LOCA, which had not expiicitly 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, i'
published in July 1980, and in Supplement 1 to MUREG-0661, published in August 1982.
t The objective of the long-term program (LTP) was to establish the design-basis loads that are appropriate for the anticipated life of each Mark I BWR facility and to restore the originally intended design-safety margins for each Mark I
~
containment system.
The principal thrust of the LTP was the development of generic methods for defining suppression pool hydrodynamic leadings 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 i
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 r
required the licensees to assess the suppression pool hydrodynamic loads in accordance with General Electric documents and NUREG-0661 on a defined schedule.
For some plants, the implementation schedule was extended by a subsequent order.
IMPLEMENTATION AND STATUS
SUMMARY
FOR QUA0 CITIES:
The staff's post implementation audit and safety evaluation report dated t
February 15, 1986 concluded that all but a few of the modifications made by Consnonwealth Edison Company were in accordance with the generic acceptance criteria contained in Appendix A of HUREG-0661, Mark I Containment Long Term Program and its supplement. Deviations from the acceptance criteria specified in NUREG-0661 were considered acceptable.
Further, the NRC staff concluded that the Plant Unique Analysis Report submitted by CECO confirmed that the containment modifications did restore the original design safety margin to the Mark I containment at the Quad Cities Station, Units 1 and ?.
Installation of all Long Term Program modifications was complete by 12/82 for Unit I and 10/87 for Unit 2.
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REFERENCES:
Quad Cities F
A-7 1.
RE0V!REMENT DOCUMENTS:
I TITLE NUDOCS NO.
DATE
~
NUREG-0661, " Safety Evaluation Report, Mark I Containment Long Term Program" 07/80 NUREG-0661, Supplement 1 08/82 i
Orders for Modification to License i
for Applicable Licensees 01/13/81 L
Orders Modifying Orders dated 1/13/81 01/19/82 Modification of Order dated 1/19/82 06/30/83 2.
IMPLEMENTATION DOCUMENTS:
TITLE NUDOCS NO.
DATE i
Quad Cities Nuclear Power 05/83 Station Units 1 and 2 Plant t
Unique Analysis Report Dresden Units 2 and 3 and 06/27/84 6
Quad Cities Units 1 and 2 Request for Information from NRC R. Rybak (CECO) to H. Denton 8408310211 08/24/84
-(NRC) letter " Response to a
Questions concerning Mark I containment Plant Unique Analysis J. Zwolinski (NRC) to 8603030411 02/15/86
- 0. Farrar (CECO) Post Implementation Audited Safety Evaluation Report of Mark I long Term Program at Quad Cities Nuclear Power Station J. Wojnarowski (Ceco) to 8605220068 05/16/86 H. Denton (NRC) letter 3.
VERIFICATION DOCUMENTS:
TITLE NUDOCS NO.
DATE
r I
PLANT Ouad Cities DOCKETN0(S).
50-254/265 j
PROJECT MANAGER Thierry Ross TECHNICAL CONTACT J. Hauck l
l USl NO. A-9 TITLE ATWS per 10 CFR 50.62 MPA NO, A;20 TAC NOS.
59132 and 59133 ISSUES
SUMMARY
l This USI was resolved in June 1984 with the publication of a final rule (10 CFR i
7' 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 i
'i anticipated transients and to mitigate the consequences of an anticipated j
transient without scram (ATWS) event.
The rule includes the following design-related requirements:
50.62(C)(1),
}
diverse and inde endent 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) alternate rod injection-(ARI) for BWRs; 50.62(C)(a); standby liquid controi system (SLCS) for BWRs; and 50.62(C)(5), automatic trip of recirculation pumps under conditions indicative of an ATWS for BWRs.
Information requirements and an implementation schedule are also specified.
IMPLEMENTATION AND STATUS
SUMMARY
FOR OUAD CITIES:
Technical Specification amendments for revised SLCS requirements in accordance with 10 CFR 50.62 were approved on 12/30/86 for Unit 2 and on 3/28/88 for Unit 1.
After exchanging correspondence (see implementation documents), the NRC issued a Safety Evaluation Report dated November 8, 1988 concluding that the ARI and RPT design at Quad Cities was acceptable, except for diversity of analog j
transmitters and trip units (ATTus) used for ARl/RPT and RPS. The issue of i
acceptable diversity is presently being appealed to the Commission by the BWR Owner's Group.
All 10 CFR 50.62 modifications were ecmpleted at Quad Cities as of'10/87 for Unit 1 and 1/87 for Unit 2, except for diversity issue impli-cations. A Technical Specification amendment application to codify the requirements for the ARI and RPT is scheduled to be submitted by June 1990.
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REFERENCES:
Quad Cities A-9 1.
REQUIREi1ENT DOCUMENTS:
TITLE NUDOCS NO.
DATE NUREG-0460, and Supplements.
03/80
'" Anticipated Transients Without Scram for Light Water Reactors" Federal Register Notice r-49 FR 26045 (10 CFR 50.62) 06/26/84 2.
IMPLEMENTATION DOCtMENTS:
TITLE NUDOCS NO.
DATE Letter from J. Zwolinski (NRC) 8701140094 12/30/86 to D. Farrar (Ceco)
Letter from J. Zwolinski (NRC) 8702060153 01/27/87 to D. Farrar (Ceco)
Letter from I. Johnson (Ceco) 8710080405 09/30/87 to T. Murley (NRC)
Letter from T. Ross (NRC) 8802230353 02/18/88 to L. Butterfield (CECO)
Letter from T. Ross (NRC) 8804050127 03/28/88 to L. Butterfield_(Ceco)
Letter from I. Johnson (CECO) 8805190186 05/09/88 to T. Murley (NRC)
Letter from B. Siegel (NRC) 8811150025 11/08/88 to H. Bliss (CECO)
Letter from J. Silady (Ceco) 8812200014 12/15/88 to.T. Murley (NRC) 3.
VERIFICATION DOCUMENTS.
TITLE NUDOCS NO.
DATE e
[
PLANT-Quad Cities DOCKET N0(S).
50-254/265 L
PROJECT HAHAGER Thierry Ross TECHNICAL CONTACT K. Wichman USI NO. A-10 TITLE BWR Feedwater hozzle Cracking MPA NO. B-25 TAC NOS.
ISSUES
SUMMARY
l This issue was resolved in November 1980 with the publication of NUREG-0619, "BWR Feedwater Mozzle and Control Rod Drive Return Line Nozzle Crackina " MPA
=
B-25 was established 'uy NRC's Division of Licensing for implementation'.
purposes.
l Inspections of operating BWRs conducted up to April 1978 revealed cracks in the t
feedwater nozzles of 20 reactor vessels.
It was determined that cracking was due to high-cycle fatigue caused by fluctuations in water temperature within the vessel in the nozzle region.
By letter dated November 13. 1980 Darrell G. Eisenhut provided licensees with a copy of NUREG-0619. The letter stated that NUREG-0619 previded the resolu-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 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
FOR OUAD CITIES:
Dy letter dated April 16, 1984 the NRC concluded in a Safety Evaluation Report that modifications of the Feedwater and CRD return line nozzles at Quad Cities were performed in accordance with NUREG-0619, and were therefore acceptable.
Actual modifications were completed 1/83 (Unit 1) and 6/80 (Unit 2).
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REFERENCES:
Quad Cities A-10 1.
_ REQUIREMENT DOCUMENTS:
l TITLE NUDOCS NO.
DATE Letter from D. Eisenhut transmitting NUREG-0619, "BWR Feedwater flozzle and Control Rod Drive Return Line Nozzle Cracking,"
resolution of A-10 to licensees 11/13/80 Generic t.etter 81-11. "BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking (NUREG-0619)"
02/20/81 2.
IMPLEMENTATION DOCUMENTS:
TITLE NUDOCS NO.
DATE Letter R. Janecek (Ceco) to
-8101270481 01/22/81 D.Eisenhut(!!RC)
Letter R. Janecek (Ceco) to 8102250396 02/23/81 D. Eisenhut (NRC)
Letter T. Novak (NRC) to 8108060227 07/20/81 J. Abel (CECO) letter T. Rausch (CECO) to 8111130608 11/06/81 D.Eisenhut(NRC)
Letter D. Vassallo (NRC) to 8405160106 04/16/84 D.Farrar(CECO) 3.
VERIFICATION DOCUMENTS:
TITLE NUDOCS NO.
DATE
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' PLANT Ouad Cities DOCKET N0(S).
50-254/265 i
t PROJECT PANAGER Thierry Ross TECHNICAL CONTACT B. Elliott USI NO.
A-11_,_
TITLE Reactor Yessel Materials Toughness flPA NO.
TAC NOS.
ISSL'ES
SUMMARY
This US! was resolved in October.1982 with the publication of NUREG-0744, l
" Pressure Vessel Material Fracture Toughness.".
NUREG-0744 was issued by Generic Letter 82-26 and provided only a methodology to satisfy the require-ments of 10 CFR Part 50, Appendix G.
No licensee response to Generic Letter 82-26 was required.
Because of the remote possibility that nuclear reactnr pressure vessels designed to'the ASME Boiler and Pressure Vessel Code would fail, the desion of nuclear facilities does not provide protection against reactor vessel failure, e
Prevention of reactor vessel failure depends primarily on maintaining the i
reactor vessel material fracture toughness at levels that will resist brittle fracture during plant operation.
At service times and operating conditions typical'of current operating plants, reactor vessel fracture toughness properties provide' adequate margins of safaty against vessel failure; however, e
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 demonstreted 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 1
initiated to address the staff's concern that seme vessels were projected to
-have beltline materials with Charpy upper shelf energy less than 50 ft-lb.
NUREG-0744-provides a method for evaluating reactor vessel materials when their Charpy upper shelf eneroy is predicted to fall below 50 ft-lb. Plants will use the prescribed method when analysis of irradiation damage predicts that the charpy upper shelf energy is below 50 ft-lb.
IMPLEMENTATION AND STATUS
SUMMARY
FOR OUAD CITIES:
i-Commonwealth Edison Company (in consultation with General Electric Company) l does not expect that the charpy upper shelf energy for reactor vessel materials of older BWRs at end-of-life will be below the 10 CFR 50 App G threshold value of 50 ft.-lb.
Consequently, this US1 does not have any implementation require-i ments applicable to Quad Cities.
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REFERENCES:
Quad Cities A-11 i
1.-
REOUIREMENT DOCUMENTS:
I TITLE NUDOCS N0.
DATE i
NUREG-0744, Revision 1, " Pressure 10/82_
Vessel Material Fracture Toughness" f-Generic Letter 82-26, " Pressure Vessel Materia 1' Fracture Toughness" 11/12/82 7
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IMPLEMENTATION DOCUMENTS:
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TITLE h0 DOCS NO.
dATE Not Applicable 3.
VERIFICATION DOCUMENTS:
TITLE NUDOCS NO.
DATE Not Applicable f
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I PLANT Quad Cities DOCKETN0(S).
50-?54/P65 PROJECT MANAGER Thierry Ross TECHNICAL CONTACT D. Thatcher l
051 NO. A-17 TITLE Systems Interactions in Nuclear Power Plants i
PPA NO.
TAC N05.
ISSUES
SUMMARY
GenericLetter(GL)89-18,datedSeptember6,1989,wassenttoallpower reactor licensees and constitutes the resolution of USl b17. The generic letter did not require any licensee actions.
GL89-18hadtwoenclosureswhich(a)outlinedthebasesfortheresolutionof USI A-17, and (b) provided five general lessons learned from the review of the nyerall 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 concernino water intrusion and flooding from internal sources, as described in the appendix to NUREG-1174, would be considered in the IPE program. Also considered in the resolution of this USl was the expectation that licensees would continue to review information on events at operating nuclear power plants in accordance with the requirements of TM1 Task Action Plan Item I.C.5 (NUREG-0737).
IMPLEMENTATION AND STATUS
SUMMARY
FOR QUAD CITIES:
Commonwealth Edison Company will evaluate the issues associated with USI A-17.(i.e. water intrusion and flooding) as part of the Individual Plant Examination process described by GL 88-20. There were no specific i
implementation requirements for Quad Cities.
With regards to an event at Quad Cities, Unit 1, the NRC issued letters to all licensees in 1972 cencerning system interactions (e.g. internal flooding).
Ceco addressed these concerns for the. Quad Cities Station in a letter dated November 10, 1972. All required modifications were completed by April 1975.
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REFERENCES:
Quad Cities f
A-17 1
1.
REOUIREMENT DOCUMENTS:
TITLE NUDOCS NO.
DATE Generic Letter 89-10 09/06/89
]
- Evaluation of May 1989 1
Systems Interactions in Nuclear 4
Power Plants" i
NUREG-1229 " Regulatory Analysis August 1989 for Resolution of USI A-17"
)
NUREG/CR-3922 " Survey and January 1985 l
t Evaluation of System Interaction.
Events and Sources" NUREG/CR-4?61 " Assessment of June 1986 j
System Interaction Experience in Nuclear Power Plants" NUREG/CR-4470 " Survey and August 1986 Evaluation of Vital Instrumentation and Control Power Supply Events" NRC Letters to Licensees 9/72 Informing Licensees of Staff Concerns Regarding Potential failure of Non-Category 1 Equipment 2.
IMPLEMENTATION DOCUMENTS:
TITLE NUDOCS N0.
DATE Letter from D. Butterfield (Ceco) 11/10/72 to A. Giambusso (flRC) 3.
VERIFICATION DOCUMENTS:
TITLE NUDOC NO.
DATE Not Applicable L
PLANT Ouad Cities DOCKETN0(S).
50-254/265 PROJECT MANAGER Thierry Ross TECHNICAL CONTACT P. Shemanski USI NO. A-24 TITLE Qualification of Class'1E Equipment MPA NO. B-60 TAC N05. 42478 and 42479 q
ISSUES
SUMMARY
4 This USI was resolved in July 1981 with the publication of NUREG-0588, Revision 1, " Interim Staff Position on Environmental Qualif. cation of Safety-Related i
Electrical Equipment. Part I of the report is the original NUREG-0588 that was issued for comment; that report, in conjunction with the Division of Operating Reactor (00R) Guidelines, was endorsed by a Comission Memorandum and i
Order as the interim position on this subject until " final" positions were established in rule making. On January 21, 1983 the Comission 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, j
including NUREG-0588.
The rule is based on the DOR Guidelines and NUREG-05BB. These provide guidance on (a) how to establish environmental service conditions, (b) how to select methods which are considered approp(riate for qualifying the equipment inc) such different areas of the plant, 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 i
completed as a result of Comission 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, t
In sumary, the resolution of A-24 is embodied in 10 CFR 50.49.
A measure of whether each licensee has implemented the resolution of A-24 may therefore be found in the determination of compliance with 10 CFR 50.49. This was addressed by 72 SERs for operating plants issued shortly after publication of the rule and subsequently in operating license reviews pursuant to Standard Review Plan Section 3.11.
This was further addressed by the first-round environmental qualification. inspections conducted by the NRC.
IliPLEMEt:TATION AND STATUS
SUMMARY
FOR QUAD CITIES:
f The NRC staff's final-Safety Evaluation Report (SER) dated January 11, 1985, concluded that Comonwealth Edison Company's Environmental Qualification e
l Program at Quad Cities for electrical equipment important to safety was in compliance with the requirements of 10 CFR 50.49.
Furthermore, the resolutions l
proposed by letter dated March 30, 1984 for each of the environmental cuali-fication deficiencies identified by a previous staff SER dated January 18, 1983 l
were acceptable and that continued operation of Quad Cities Units 1 and 2 would not present undue risk to the public health and safety. Ceco certified by letter dated February 14, 1985 that the EQ program at Quad Cities complied with 10 CFR 50.49.
Environmental Qualification of safety-related electrical equipment was fully implemented at. Quad Cities by the end of 1985 in accordance with 10 CFR 50.49.
Region III conducted a special team inspection (documented by a report dated September 1, 1987) that confirmed implementation of the EQ program.
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REFERENCES:
Quad Cities A-24 1.
REQUIREMENT DOCUMENTS:
s TITLE NUDOCS NO.
DATE 00R " Guidelines for Evaluating Environmental Qualification of Class IE Electrical Eauipment in Operating Reactors" NUREG-0588, " Interim Staff Position on Environrental Qualification of Safety Related Electrical Equipment" 12/79 Commission Pemorandum and Order, CL1-80-21, on D0R Guidelines and 05/23/80 NUREG-0588 MUREG-0588,. Revision 1 07/81 10 CFR 50.49 (48 FR 2730-2733) 01/21/83 Standard and Review Plan 3.11, Environmental Qualification of Mechanical and Electrical Equipment 07/81 2,
IMPLEMENTATION DOCUMENTS:
TITLE NUDOCS N0.
DATE Letter from J. Abel (CEC) to 8011200062 10/30/80 J. Keppler (NRC) s Letter from T. Novak (NRC) to 8103260149 03/17/81 J. Abel (CECO)
Letter from T. Ippolito (NRC) 8106100435 06/03/81 to J. Abel (Ceco)
-Letter from T. Rausch (CECO) to 8109100411 09/04/81 H.Denton(NRC)
Letter from D. Vassallo (NRC) 8301250202 01/18/83 to L. Del George (CECO) letter from B. Rybach (Ceco) to 8404090255 03/30/84 H. Denton (NRC)
Letter frnm 8. Rybach (CECO) 8405300130 05/21/84 to H. Denton (HRC)
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REFERENCES:
Quad Cities l
A.P4 Letter from D. Vassallo (NRC) 8502010816 01/11/85 to D. Farrar (Ceco)
Letter'from D. Farrar (CECO) 8502200269 02/1A/85 to H. Denton (NRC) 3.
VERIFICATION DOCUMENTS:
TITLE NUDOCS NO.
DATE
'l Inspection Report 6709090154 09/01/87 i
50-?54(265)/87-011
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PLANT Ouad Cities DOCKET N0(S).
50-254/265 I
PROJECT MANAGER Thierry Ross TECHNICAL CONTACT J. Wermiel i
USI NO. A-36 TITLE Control of Heavy Loads. Phases I & 11 MPA NO.
C-10, C-15 TAC NOS.
I i
ISSUES
SUMMARY
I This USI was resolved in July 1980 with the publication of NUREG-0612,)* Control of Heavy Loads at Nuclear Power Plants," and Standard Review Plan (SRP Section j
9.1.5.. The staff established MPAs C-10 and C-15 for the implementation of i
Phases 1 and II, respectively, of the resolution of this issue at operating l
plants.
In nuclear power plants, heavy loads may be handled in several plant areas.
If these loads were to drop in certain locations in the plant, they may impact
.i spent fuel, fuel in the core, or equipment that may be required to achieve safe shutdown and continue decay heat removal.
USl A-36 was established to systematically. examine staff licensing criteria and the adecuacy of measures in effect at operating plants, and to recommend necessary changes to ensure the safe handling of heavy loads. The guidelines proposed in NUREG-0612 include definition of safe load paths, use of load handling procedures, training of crane operators, guidelines on slings and special lifting devices, periodic j
inspection and maintenance for the crane, as well as various alternatives.
By Generic Letters dated December 22, 1980, and February 3, 1981 (Generic Letter 81-07), all utilities were requested to evalcate their plants against i
the guidance of NUREG-0612 anti to provide their submittals in two parts: Phase I (six month response) and Phase II (nine month response).
Phase I responses were to address Section 5.1.1 ef NUREG-0612 which covered the following areas:
l 1.
Definition of safe 1 cad 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 11 responses were to address Sections 5.1.2 thru 5.1.6 of NUREG-0612 which covered the need for electrical interlocks / mechanical stops, or 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 ' Control of Heavy l
Loads at Nuclear Power Plants' - NUREG-0612," all licensees have empleted the
-requirement to perform a review and submit a Phase I and a Phase li report.
Based on the improvements in heavy loads handlirg obtaineo from implementation of NUREG-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.
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[E 0x PLANT Ouad Cities DOCKETN0(S).: 50-254/265 1
- ni;o 1
gjE PROJECT; MANAGER Thierry Ross TECHNICAL CONTACT J.-Wermiel hen USI NO. A-36
- TITLE Control of Heavy' Loads, Phases I & 11 f0 HPA NO.
C-10, C-15 TAC NOS.
While not a requirement, NRC' encouraged the implementation of any actions identified in Phase II regarding_the handling of heavy loads that were considered appropriate, IMPLEMENTATION AND STATUS SUPMARY FOR 00AD CITIES:
1
. Commonwealth. Edison Company submitted a number of letters in response to the Generic Letters that transmitted NUREG-0612. The NRC staff, in a Safety.
Evaluation Report dated June 27, 1983, concluded that Phase I for " Control of
. Heavy Loads" was acceptable at Quad Cities and that the guidelines contained in-NUREG-0617 were satisfied.
Phase 1.1 of NUREG-0612 was completed by Generic Letter 85-11..
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REFERENCES:
Quad Cities c
A,
_1. -
REQUIREMENT D'0CUMENTS W
TITLE NUDOCS NO.
DATE Letter, Darrell G. E'isenhut, 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' l
NUREG-0612" 06/28/85 2.
IMPLEMENTATION DOCUMENTS:
TITLE-NUDOCS NO.
DATE Letter _from J. Abel (Ceco) 8105190445 05/15/81' s
.to D. Eisenhut'(NRC)
~
LetterfromE.Swartz(CECO) 8106290327 06/22/81 to D. Eisenhut (NRC)
Letter from E. Swartz (Ceco) 8112170178 12/11/81
- to D. Eisenhut (NRC).
LetterfromE.Swartz(Ceco) 8205110447 05/04/82_
to D. Eisenhut (NRC)
LetterfromE.Sw'artz(CECO) 8211230119 11/18/82 to D. Eisenhut:(NRC)
Letter from B. Rybach (CECO) 8304220567 04/15/83 toD.-Eisenhut(NRC)
LetterfromD.Vassallo(NRC)-
8307180026 06/27/83 to D. Farrar (Ceco) t 1
3.
VERIFICATION DOCUMENTS:
TITLE NUDOCS N0.
DATE n
/
4
6 PLANT Quad cities-DOCKET N0(S).
50-254/265 PROJECT MANAGER Thierry Ross TECHNICAL CONTACT J. Kudrick
' USI NO. A-39 TITLE Determination of SRV Pool Dynamic Loads and Temperature Limits HPA-NO..
TAC NOS.
ISSUES SUMMAP,Y:
This USI was resolved with the publication of Standard Review Plan (SRP)
Section 6.2.1.1.C, in October 1982. In addition, NUREGs 0763, 0783 and 0802 were issued for Mark I, Mark II, and Mark III containments, respectively.
.j BWR plants are equipped with safety / relief valves (SRVs) to protect the reactor from overpressurization.
Plant operational transients,-such as turbine trips, will-actuate the'SRV. Once'the SRV opens, the air column within the partially submerged discharge line is compressed by the high-pressure steam released from the reactor. The compressed air discharged into the suppression pool prodm es high-pressure bubbles. Oscillatory expansion and contraction of these bubt;ies create hydrodynamic loads on the containment structures, piping, and equipment
~inside containment.
NUREG-0802 presents the results of the staff's evaluation of SRV loads. The t
evaluation, however, is limited to the cuencher devices used in Mark 11 and III L
containments. With respect to Mark I containments, the SRV acceptance criteria are presented in NUREG-0661, " Safety Evaluation Report, Mark'I Containment and Long-Term Program," and are dealt with as part of USI A-7.
SRP'Section 6.2.1.1.C addresses the applicable review criteria, since all Mark II and III containment designs are understood to have completed their operating
-license (OL). reviews subsecuent to resolution of this USI and reflection of the L
resolution in the SRP.
IMPLEMENTATION AND STATUS
SUMMARY
FOR~ QUAD CITIES:
t-L Ou'ad Cities is a Mark I containment f acility. Consequently, any implemeMation requirements associated with this USI were addressed by A-7 (Mark I Long Term
~ Program).
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REFERENCES:
Quad Cities A-39 1.
REQUIREMENT DOCUMENTS:
TITLE NUDOCS NO.
DATE SRP 6.2.1.1.C, Pressure Suppression Type CWR Containments NUREG-0802, " Safety / Relief Yalve Quencher Loads:
Evaluation for BWR Mark 11 and III Containments, Generic Technical Activity A-39" 1982 HUREG-0661, " Safety Evaluation Report -
7/80
. Mark I;Long Term Program" g.
2.
IMPLEMENTATION DOCUMENTS:
TITLE NUDOCS NO.
DATE Not Applicable i
4 3.
' VERIFICATION DOCUMENTS:
TITLE NUDOCS NO.
DATE Not applicable
- *The applicable SRP revision number would depend on the date of the evaluation for each specific plant.
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PLANT Quad cities DOCKET N0(S).
50-254/265 PROJECT MANAGER Thierry Ross
. TECHNICAL CONTACT H. Ashar i
l USI NO.
A TITLE Seismic Design Criteria MPA_NO.
TAC NOS.
-i ISSUES-
SUMMARY
The staff has resolved USl A-40 as documented in NUREG/CR-5347, "Recommenda-1
'tions for Resolution of Public Comments on USI-A-40," issued in June 1989, and NUREG-1233,= " Regulatory Analysis for US1 A-40," issued in September 1989.
For plants not covered under the scope of USI A-46, " Seismic Qualification of Equipment in Operating Plants," the staff concluded that tanks-in plants that were subject to licensing review by the staff after 1984 had been reviewed to current requirements and found acceptable.
For tanks in plants reviewed during 1980-1984, the staff identified four plant sites (six units) that were not s
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) to recommend modifications to the Standard Review Plan (SRP) if changes are justified.
Lawrence Livermore National Laboratory (LLNL) completed the study.and published its findings in NUREG/CR-1161, " Recommended Revisions-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 major 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) location and reductions of input ground notion 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 requ_ ired for these items.
However, the revised provkions could be used for margin studies and reevaluations or individual plant examination for external events (IPEEE).
The participant utilities in the Seismic Qualification Utility Group (SQUG) agreed to implement the changed criteria for flexible vertical tanks for their plants.
For the four plants where this issue has to be resolved on an indi-vidual basis a 10 CFR 50.54(f) request-for-information letter has been sent to the affected utilities.
If the information received indicates that large above-ground vertical tanks do not meet the new criteria, plant-specific backfits will be considered.
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PLANT-Quad Cities-DOCKET N0(S)'.
50-254/265
' PROJECT MANAGER Thierry Ross' TECHNICAL CONTACT H. Ashar USI NO. A-40 TITLE: Seismic Design Criteria MPA N0..
TAC NOS.
IMPLEMEllTATION AND STATUS
SUMMARY
FOR QUAD CITIES:
. Quad Cities was licensed prior to origination'of.USI A-40 in 1977.
Commonwealth Edison Company is a participant utility in SQUG.
Implementation
-requirements associated with flexible vertical tanks'will be addressed as part of USI A-46, therefore no changes are necessary for this USI.
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REFERENCES:
Quad Cities A-a0 1.-
REQUIREMENT DOCUMENTS:
TITLE
'NUDOCS NO.
DATE Regulatory Analysis for NUREG-1233 Sept. 1989 i
USI A 40 Recommendations for Resolution NUREG/CR-5347-June 1989 of Public Coments on USI A-40 Standerd 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 Earthquake Excitation Engineering Characteri-NUREG/CR-3805 Feb.-Aug. 1986 zation of Ground Motion, Vols. 3,4,5 Proceedings of the NUREG/CR-0054 June 1986 Workshop on Soil-Structure Interaction
'Value Impact Assessment NUREG/CR-3480
.Aug. 1984 for Seismic Design Criteria' Seismic' Hazard Analysis NUREG/CR-1582 Oct. 1981
. Application of. Methodology, Results and Sensitivity Studies,xVol.'4 t
Recomended Revision to NUREG/CR-1161 May 1980 l
Huclear Regulatory Commission c
Seismic Design Criteria Power Spectral' Density Functions NUREG/CR-3509 June 1988 Compatible with NRC R.G. 1.60 l
Response Spectra l
2.
IMPLEMENTATION DOCUMENTS:
TITLE NUDOCS N0.
DATE L
Not applicable 3.
VERIFICATION DOCUMENTS:
TITLE NUDOCS NO.
DATE Not applicable
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PLANT Ouad'CiEies DOCKET N0(S).
50-254/265 s
PROJECT MANAGER Thierry Ross TECHNICAL CONTACT W. Koo USI H0.: A-42' TITLE Pipe Cracks in Boiling Water Reactors MPA NO.: B-5/97 TAC NOS.
69154/69155 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 HUREG' document was issued to all-holders of BWR operating licenses or construction permits and to all applicants for BWR operating licenses. The staff established MPA B-05 for
' implementation of the resolution at operating plants.
Pipes have cracked in the heat-affected zones.of welds in primary system piping in BWRs since mid-1960. These cracks have occurred mainly in Type 304 stainless
. steel, which.is the type used in most operating BWRs. The major problem is recognized to:be intergranular stress corrosion cracking-(IGSCC) of austenitic stainless steel components that have been made susceptible to this failure by being " sensitized," either by post-weld heat treatment or by sensitization of a narrow heat affected zone near welds.
" Safe ends" that have been highly sensitized by furnace heat treatment while attached to vessels during fabrication were found to be susceptible to IGSCC in
_the late 1960s. Most of the furnace-sensitized safe ends in older plants have.
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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.
HUREG-0313, Revision 1 provided the NRC staff's revised acceptable methods for reducing the'IGSCC susceptibility of BWR code class 1, 2, and 3 pressure-1 boundary piping of sizes identified above and safe ends.
In addition, it previded the requirements for augmented inservice inspection of piping with nonconforming materials.
As a result of further IGSCC degradations in larger piping, the staff provided licensees with additional requirements in several NRC communications (i.e.,
Bulletins 82-03,83-2,and84-11). The long-_ term resolution of IGSCC in BWR piping (including.the scope of A-42) was provided in NUREG-0313, Revision 2 which was transmitted to all holders of BWR operating licenses via Generic Letter 88-01, a
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L LANT-Quad ' Cities 00CKETN0(S).
50-254/265 P
4 PROJECT PANAGER Thierry Ross TECHNICAL CONTACT W. Koo i
LUSI NO. A-42 TITLE. Pipe Cracks in Boiling Water Reactors MPA NO.~ B-5/97 TAC N05, 69154/69155 IMPLEMENTATION AND STATUS
SUMMARY
FOR QUAD CITIES:
t
' The requirements in GL 81-04 were superseded by the requirements in GL 84-111
-(MPA B-84). MPA B-84 was closed as being fully implemented on all BWRs as of i
February 12, 1988 by a. 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 USI A-42.
The requirements in GL 84-11.were in
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 NT0Ls at-the time, the implementation
. document is the SER'or SSER in which the staff evaluated the. applicants'confor-mance to'the reouirements'of NUREG-0313, Rev. 1.
For the Quad cities-facility, the closure date for A-42.is 7/29/88..
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REFERENCES:
Quad Cities
%g)'
A-42 i
l.'
REQUIREMENT DOCUMENTS:-
-TITLE.
NUDOCS N0.
DATE t
HUREG-0313;_ Revision-1 " Technical Report on Material Selection.and Processina Guidelines for BWR Coolant Pressure Boundary Piping,"
07/80
_ Generic Letter 81-04; "Implemen-02/26/81 tation of NUREG-0313, Rev. 1 for
. Selection and Processino Guidelines for BWR Coolant-Pressure Boundary Piping (GenericTaskA-42)"
2.
IMPLEMENTATION DOCUMENTS:
TITLE NUDOCS NO.
DATE c
LetterfromL.De1 George (Ceco) 8107140437 07/07/81
-toD..Eisenhut-(NRC)
Letter'from G. Laines (NRC) 8211030535 10/28/82 to L.~De1 George (Ceco)
Letter from T. Rausch (CECO) 8301140193 01/07/83
- to D.- Eisenhut (NRC)
Letter from T..Rausch (CECO) 8301250134 01/18/83 to H. Denton,(NRC)
Letter from A. Schewncer (NRC) 8406210179 06/11/84
.to D. Farrar (NRC)
Letter from B.- Ryback (Ceco) 8406110246 06/04/84 to H. Denton (NRC)
Letter from B. : Ryback (Ceco) 8408060108 07/30/84 to H. Denton'(NRC)
Letter from D. Vassallo (NRC) 8408160465 08/02/84-to D. Farrar (Ceco)
LetterfromW.. Morgan'(Ceco) 8808090125-07/29/88 to U.S.NRC LetterfromT.Ross(NRC) 8905310228 05/22/89
-and B.:Siegel (NRC) to T.:Kovach (CECO)
Letter from M. Richter 8907270336 07/21/89 to U.S.NRC 3.
VERIFICATION DOCUMENTS:
TITLE NUDOCS NO.
DATE
, Inspection Repnrts 50-254/85-16 8507020466 06/18/85 and 50-265/85-08 Inspection Report 50-254 & 265 8911280128 11/16/89 89-21
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PLANT Quad Cities-DOCKET N0(S).
50-254/265 PROJECT MANAGER _ Thierry Ross TECHNICAL CONTACT A. Serkiz USI NO.
A-43~
TITLE Containment Emergency Sump Performance MPA NO.
TAC NOS.
ISSUES SUWAy:
- 19. USI NO. A a3 TITLE: Containment Emergency Sump Performance The resolutinn of this USI was presented to the Connission in October 1985 in SECY-85-349.
NUREG-0897, Revision 1, " Containment Emergency Sump Performance,"
presents the results of the staff's technical findings. These findings estab-lished a need to revise current licensing guidance on these matters.
RG 1.82 Revision 0 and' Standard Review Plan Section 6.P.2, " Containment Heat Removal' Systems" were revised to reflect this new guidance.
No licensee actions were J
required.
Initially, an issue existed concerning the availability of ade tioncoolingwaterfollowingaloss-of-coolantaccident~(LOCA)quaterecircula-
'when long-term recirculation of cooling water from the PWR containment sump, or the BWR
- residual heat removal system (RHR) suction intake, must be initiated and maintained to prevent core melt.
The technical concerns evaluated under USI A-43 were:
(a) post-LOCAadverse l
-: conditions resulting from potential vortex formation and air ingestion ~end subsequent pump failure, (b) blockage of sump screens with LOCA generated insulation debris causing inadequate net positive suction head (NPSH) on pumps, and (c) RHR and containment spray pumps inoperability due to possible air, debris, or particulate ingestion on pump seal and bearing systems.
This revised guidance applies only to future construction permits, preliminary design approvals,-final design approvals, standardized designs, and applica-tions for' licenses to manufacture. The staff-performed a regulatory analysis 1.0 determine if this new guidance should be tpplied 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
FOR QUAD CITIES :
Results of USI did not warrant any new generic requirements for present licensees.
.-.-..2
-]
REFERENCES:
Quad Cities A-43
- 1. -
REQUIREMENT DOCUMENTS-TITLE-NUDOCS H0.
DATE-
'NUREG-0869, Rev. 1, "USI 10/85
- A-43 Regulatory Analysis" NUREG-0897, Rev. 1,=" Containment 10/85 Emergency Sump Performance" GL 85-22, " Potential for loss 12/03/85 of Post-LOCA Recirculation
. Capability Due to Insulation Debris Blockage" 2..
IMPLEMENTATION DOCUMENTS:
TITLE NUDOCS NO.
DATE Not applicable 3.
VERIFICATION DOCUMENTS:
m
-TITLE NUDOCS NO.
DATE Not applicable 5
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PLANT Ouad Cities DOCKETN0(S).
50-254/265 L
--PROJECT MANAGER Thierry Ross TECHNICAL CONTACT P. Gill USI NO. -A-44 TITLE Station Blackout-j_
MPA NO.
A-22 TAC NOS.
68590 and 68591 H
ISSUES-
SUMMARY
This USI was resolved in June 1988 with the publication of a new rule (10 CFR 50.63) and Regulatory Guide 1.155.
c Station blackout _means the loss of offsite ac power to the essential and nonessential electrical buses concurrent with turbine trip and the unavailability of the redundant onsite emergency ac power systems. WASH-1400 showed that station blackout could be an-important risk contributor, and operating experience has indicated that the reliability of ac power systems might be.less than originally anticipated.
For these reasons station blackout was designated as a USI in 1980. A proposed rule was published for comment on March 21. 1986. A final rule. 10 CFR 50.63, was published on June 21, 1988 and became effective on July 21, 1988.
Regulatory Guide 1.155 was issued at the same time as the rule and references an industry guidance document, NUMARC-8700.
In order to comply with the_A-44 resolution, licensees will be required to:
maintain onsite emergency ac power supply re' liability above a minimum level develop proce' ures and training for recovery from a station blackout d
~ determine the duration of a station blackout that the plant should be able to withstand use an alternate qualified ac power source, if available, to cope with a station blackout evaluate the plant's actual capability to withstand and recover from a station blackout backfit hardware modifications if necessary to improve coping ability Section-50.63(c)(1) of the rule required each licensee to submit a response including the-results of a coping analysis.within 270 days from issuance of an operating license or the effective date of the rule, whichever is later.
IMPLEMENTATION AND STATUS
SUMMARY
FOR QUAD CITIES:
Commonwealth Edison Company (CECO) responded to 10 CFR 50.63 with a letter
. dated April 17, 1989. During a working group meeting on October 4, 1989 to discuss Quad Cities' compliance with the rule on Station Blackout a number of staff concerns surfaced regarding CECO's approach. A subsequent meeting
-between CECO and NRC was held December 20, 1989.
CECO is preparing to supplement their original response to address staff concerns.
Resolution of A-44 is still being evaluated by the staff and should be completed by June 1990. The licenseee's implementation date is expected to be mot later than June 1992 based on the requirement stated in the rule.
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' Quad Cities-x REFEREllCES:
A-44 1.
REQUIREMENT DOCUMENTSi TITLE' NUDOCS HO'.
DATE 10-CFR 50.63, " Loss of All-Alternating Current Power" 06/21/88 Regulatory Guide 1.155,.
m
?
" Station Blackout" 08/88 2.
. IMPLEMENTATION DOCUMENTS:
TITLE
~NUDOCS NO.
DATE Letter from M. Richter (CECO) 690424042 04/17/89 to T.;Murley-(NRC) l e
3.
-VERIFICATION DOCUMENTS:
TITLE-
-NUDOCS NO.
DATE q
il i
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L n
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s I$
' PLANT Quad Cities
- DOCKETN0(S).
50-254/265 g
PROJECT MANAGER. Thierry Ross TECHNICAL CONTACT R. Jones USI NO.- A TITLE Shutdown Decay Heat Removal Requirements
'MPA'NO.
TAC NOS.
ISSUES
SUMMARY
USI A-45 was resolved by SECY 88-260, " Shutdown Decay Heat Removal Requirements.
I
'(USI-A-45),". issued September 13, 1988, without imposing any new licensing jrequirements other than the Individual Plant Examination (IPE), as described below. At the same time the staff issued NUREG-1289, " Regulatory and Backfit 1
' Analysis:: USI A-45."
Since all of the significant USI A-45 results have been
-l found to be highly plant specific, the Commission decided-it was not
]
appropriate to propose a single generic corrective action to be applied uniformly to all plants.
)
The Commission is currently implementing the Severe Accident Policy (50 FR 1
32138) and will require all plants presently-operating or under construction to l
undergo a systematic examination termed the IPE. The reason for this examina-i tion is to identify any plant-specific vulnerabilities to severe accidents.
The -IPE analysis intends to examine and understand the plant emergency pro--
cedures, design, operations, maintenance, and surveillance, in order to identify vulnerabilities..The analysis will examine both the decay heat removal systems-and those systems used for other related functions. This includes CE plants-without power-operated relief valves.
NRC has decided to subsume A-45 into the IPE program as the rost effective way of achieving resolution of specific plant concerns associated with A-45.
IMPLEMENTATION AND STATUS
SUMMARY
.FOR QUAD CITIES:
l No implementation requirements are associated with this USI.
Resolution of plant: specific concerns have:been subsumed by the IPE Program (GL 88-20). Quad Cities happened to be one of the plants selected for site visit by the USI A-45 program team from NRC and Sandia National Laboratory. Arrangements and expectations of the team's visit were communicated to Commonwealth Edison Company in a letter dated October 9, 1984.
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REFERENCES:
Quad Cities
._ A-45 1.
REQUIREMENT-DOCUMENTS ci-TITLE-NUDOCS NO.
DATE Federal Register Notice "10 CFR Part 50, Shutdown Decay Heat Removal Requirements"-
NUREG/CR-5230 " Shutdown Decay Heat April 1989 Removal Analysis:
Plant _ Case Studies and Special Issues Summary Report" NUREG-1289 " Regulatory and Backfit 11/30/88
-Analysis for the Resolution of USI.A-45" SECY-88-260." Shutdown Decay Heat 09/13/88 Removal Requirements 2.
IMPLEMENTATION DOCUMENTS:
TITLE NUDOCS NO.
DATE
~
.LetterfromD.Vassallo(NRC) 8410310691 10/09/84 to D. Farrar (Ceco)
-3.
VERIFICATION DOCUMENTS:
TITLE-NUDOCS NO.
DATE Not applicable e
7
o s
PLA'NT Ouad Cities?
DOCKETN0(S).
50-254/265 PROJECT MANAGER Thierry Ross-TECHNICAL CONTACT P. Y.-Chen
'USI NO. A-46 TITLE Seismic Qualification of Equipment in Operating Plants MPA NO. B-105 TAC NOS.
69476/69477 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 (SQUG) and Electric Power Research l
Institute (EPRI). This approach was endorsed by the Senior Seismic Review and Advisory Panel (SSRAP) and approved by the NRC staff.
f 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.
.L As an outgrowth of the Systemat-ic Evaluation Program (SEP), the need was
. identified for reassessing design criteria and methods for the s6ismic quali-
'fication of mechanical eouipment and electrical equipment. Therefore, the seismic oualification 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 scismic 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 adecuacy 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.
r IMPLEMENTATION AND STATUS
SUMMARY
FOR QUAD CITIES:
The Generic. Implementation Procedure (GIP), Revision 0,(was submitted by SQUG on June 3, 1988. The staff issued a 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 1.3, 1988. Supplemental information was submitted by SOUG on March 11, 1989. The staff _has prepared a supplemental SE for GIP, Revision 1 and has submitted it to the CRGR for review. The target date for issuance of the supplemental SE is the end of-1989. An additional supplement is scheduled for January 1990 and overall closecut of implementation projected for 1993. Commonwealth Edison Company is waiting for staff approval of the SOUG GIP and issuance of a final SER before scheduling plant walkdowns at Quad Cities. Assuming the staff issues a SER in early 1990, CECO anticipates final implementation of A-46 could be accomplished by end of 1992 for Unit 1 and 1993 for Unit P.
CECO had originally scheduled (by letter dated October 7, 1988) final implementation for
'1991 but the staff's evaluation of the GIP has slipped.
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REFERENCES:
Quad Cities A
'1.
REOUIREMENT DOCUMENTS: '
. TITLE NUDOCS NO.
DATE Generic Letter,87-02,'"Verifi-cation of Seismic Adequacy of Mechanical and Electric Eouipment-i in Operating. Reactors" 02/19/87 NUREG-1211, " Regulatory Analysis for Resolution of Unresolved Safety-i Issues A-46..."
02/87 NUREG-1030, " Seismic Qualification of Equipment in Operating Plants, Unresolved Safety' Issue A-46" 02/87 Letter attached with " Generic Safety Evaluation Report on'SQUG GIP,) Revision 0," from L.-Shao (NRC to Neil' Smith (SQUG) 07/29/88
- 2.
IMPLEMENTATION DOCUMENTS:
TITLE NUDOCS NO.
DATE
" Generic' Implement'ation Procedure
- (GIP for Seismic _ Verification of
-Nuclear-Plant Equipment," Revision 0.
06/88
. Letter from L. Shao (NRC) to 8808030335 07/29/88 N.--Smith _(SQUG)
~
Letter from 1. Johnson.(CECO) 8810120140 10/07/88 to T. Murley (HRC)-
I
" Generic Implementation Procedure
~(GIP) for Seismic Verification of Nuclear: Plant Eouipment," Revision I 12/88 3;
. VERIFICATION DOCUMENTS:
TITLE NUDOCS N0.
DATE I
s L
. =
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~
~
c PLANT. Quad Cities DOCKET N0(S).
50 254/265 p
PROJECT MANAGER Thierry Ross-TECHNICAL CONTACT J. Mauck US1 NO.. A-47 TITLE Safety implication of Control Systems in LWR Nuclear Power Plants MPA.NO.
TAC NOS.
74986/74987 ISSUES
SUMMARY
USI A-47 was resolved September 20, 1989, with the publication of Generic
?
Letter (GL)89-19.
.The generic letter states:
"The staff has concluded that all PWR plants should provide 1
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 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 recomenda-tiens are consistent with the criteria and the risk considera-tions of the Comission -Interim Policy Statement on Technical
~
Specification improyecznt.- In addition, the staff recommends that all BWR recipients reassess and modify,-if needed, their-operatvng-procedures and operator training to assure-that the operators can mitigate reactor vessel overfill events-that may occur via the condetsate booster pumps during reducea 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-comitments within 180 days of the letter's date. The implementation schedule for actions on which comitments are made should be prior to startup after the first< refueling. outage, but no later than the second refueling outage, l
beginning 9 months after receipt of the letter.
IMPLEMENTATION AND STATUS
SUMMARY
FOR QUAD CITIES:
L E
Commonwealth Edison Company's initial response is not due until March 31, 1990.
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~
REFERENCES:
Quad Cities
=A,
- 1..
REQUIREMENT DOCUMENTS TITLE NUDOCS NO.
DATE Generic Letter 89-19' 09/20/89
" Request.for Action Related to Resolution of USI'A-47" NUREG-1217 " Evaluation of Safety June 1989 Implications of Control Systems in LWR Nuclear Power' Plants" NUREG-1218 "Regulatcry Analysis July 1989 for Resolution of.USI A-47" 2'
IMPLEMENTATION DOCUMENTS:
TITLE NUDOCS NO.
DATE i
3.,
VERIFICATION DOCUMENTS:
TITLE NUDOCS NO.
DATE g-
.l
}
u L
1
e PLANT Quad Cities DOCKETN0(S).
50-254/265 PROJECT MANAGER Thierry Ross TECllNICAL CONTACT J. Kudrick lUSINO. A-48
< TITLE Hydrogen Control Measures and Effects of Hydrogen 3
Burns on Safety Equipment-
54017/54018 ISSUES
SUMMARY
The NRC staff concluded April 19, 1989, that USI A-4B is resolved, as stated in SECY 89-122.
t USI A-48 was initiated as a result of-the large amount of hydrogen generated-and burred within containment during the Three Mile Island (TMI) accident, i
This issue covers hydrogen control measures for recoverable degraded core i
accidents for all=BWRs and those PWRs with ice condenser containments'.
Extensive research in this area has led to significant revision of the Com-mission's hydrogen control regulations, given in 10 CFR 50.44, published.
December 2, 1981.
10 CFR 50.44 requires inerting of BWR Mark I ar.d Mark II 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 nitrogen) which effectively precludes combustion of any hydrogen generated. ' USI A-48 with respect to BWR Mark I and 11 containments is resolved and fully implemented for all but seven Mark BWRs.
The rule for BWRs with Mark III containments and PWRs with' ice condenser 1
containments was published on January 25, 1985. The rule required that these
-plants'be provided with a means for controlling the quantity of hydrogen produced, but did not specify the control method.
In addition, the task action (plan for USI A-48 provided for plant-specific reviews of lead plants for
- reactors with Mark III and ice condenser containments. Sequoyah was chosen as the lead plant for ice condenser containments and Grand Gulf for Mark III containments. Both of the -lead plant licensees chose to install igniter-type systems which would burn.the hydrogen before it reached threatening concentra-tions within the containment.
Final design igniter systems have been installed i
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, i
Large dry PWR containments were excluded from USI A-48 because they have a j
greater ability to accommodate the large cuantities of hydrogen associated with a recoverable degraded core accident than the smaller Mark I, II, III and ice condenser containments. However, this issue has continued to be considered and, in 1989, hydrogen control for large dry PWR containments was identified as a high-priority Generic issue (GI) 121. The resolution of GI 121 is being
-nctively pursued in close coordination with more recent research findings.
IMPLEMENTATION AND STATUS
SUMMARY
FOR OUAD CITIES:
USI A 48 was resolved on April 19, 1989, as stated in SECY 89-122.
It is considered fully implemented at BWP Mark I and Mark Il facilities, as these facilities use inerting as a method of hydrogen control.
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-PLANT Ouad Cities DOCKETN0(S).
50-254/265 F
r PROJECT MANAGER Thierry Ross TECHNICAL CONTACT J. Kudrick j
J USI NO. A-48 TITLE Hydrogen Control Measures and Effects of Hydrogen Burns on Safety Equipment MPA NO. A-19 TAC NOS. 54017/54018 There is'a related issue t' hat remains open on several Mark _I facilities.
This issue is associated.with the requirement to have a recombiner. capability at all facilities. Generic Letter 87-04, "Recombiner Capability Requirements of L10 CFR 50.44(c)(3)(22)," provided guidance to those Mark 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.
t On January 20, 1987 a meeting was held between the staff and utility representatives, including, Commonwealth 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 ont 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.
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REFERENCES:
Quad Cities A-48
REFERENCES:
1.
REQUIREMENT DOCUMENTS:
TITLE NUDOCS NO.
'DATE 10 CFR 50.44, Standards for.
-12/81
.I Combustible Gas System in 1
'Li ht-Water-Cooled Power 9
-Reactors SECY-89-122, Resolution'of 04/19/89 USI A-48, " Hydrogen Control
. Measures and Effects of
' Hydrogen Burns on Safety Ecuipment" 2.
.1MPLEMENTATION DOCUttENTS:'
TITLE NUDOCS NO.
DATE Letter.from T. Rausch (Ceco) 8209220196 09/15/82 to D. Eisenhut-(NRC) 3 Letter from B. Rybach (Ceco) 8407030092 06/?5/84
.toH.~Denton-(NRC)
Letter from D. Vassallo (NRC) 8509300568~
09/26/85 to'D.-Farrar(CECO)
Letter from J. Wojnarowski 8511220234 11/14/85 (Ceco) to D. Vassallo (NRC)
Letter from R. Bernero (NRC) 8608190542 08/11/86 to D. Farrar (Ceco)
Letter from J. Wojnarowski 8912200424 09/16/86
- (CECO) to' J. Zwolinski- (NRC)
Letter from ti. Grotenhuis'(NRC) 8705140076 04/24/87
.to D. Farrar-(Ceco) j Letter from G. Holahan (NRC) 8905110238 05/03/89 to C. Reed (CECO) 3.
VEP.1FICATION DCCUMENTS:
TITLE NUDOCS NO.
DATE
.y g.
1 p ;.
{ b) C L 0 i U S E.-
3 1Page No, 1
6 h
- 0:!07/?0 LISTING OF INCCMPLETE USl DATA' FOR INPUT FR3M PROJECT MANAGERS
!! SUE ISSUE I,ESCRIPTIVE NAME IMPLEMENT IMPLEMENT LICENSEE C09 MENT STAFF COMMENT NUMPER-DATE STATUS j
r t
- 18 FLANT NAME: QUAD Cli!ES 2 A-01 WATER NAPMER
/ /
NC A-)2'-
ASYMMETRICPLONDOWNLOADSON
/ /
N/A PWR ONLY REACTOR PP! MARY COOLANT SYSTEMS A-03 -WESilN6H0VSE STEAR 6ENERATOR TUPE / /
N/A WESi!NSHOUSEONLY
' INTEGRITY, j!
A-04" -CE STEAM BENERATOR TUBE INTE6RITY / / 'N/A CE PLANTS ONLY-A-05: B&W STEAM SENERATOR TUPE
/ /
N/A C1W PLANTS ONLY
'!NTESRITY A*06-
.9 ARK 1 SHORT TERM PRO 6 RAM-06/07/7S C DELTA P CONTROL is A-071 MARK l LONG-TERM FR03 RAM 10/31/87 C A 08 PARK !! CCNTAINMENT POOL DYNAMIC- / /
N/A MK 11 PWR DNtY.
LDADS:- LONG-TERM PROGRAM A 09 ATWS
//
DIVERSITY i T/S
.A 101 BWR FEEDWATER N0!!LE CRACKING 06/30/90 C
=A-!!
REACTOR VESSEL MATERIALS
/ /
NC 10U6HNESS A-12 FRACilMETOU6HNESSOFSTEAM
// ' N/A CP AFTER 93 CNLY GENERATOR AND PEACTOR C30LANT
.FUMP SUFFORTS I A 175 SYSTEPS INTERACTION -
/ :/
.NC IPE NOREQUIREMENTS-
- A 24 QUALIFICATION CF CLASS 1E 02/14/85 C
-SAFETY-RELATED EDU!PMENT.
A 26 REACTOR VESSEL PRESSURE TRANSIENT / /
N/A PWR ONLY FR0iECTICN
-A SHR SHUTDOWN REQUIREMENTS'
/ /
N/A NEW PLANTS ONLY. SRP..
- A-36.
CONTROL OF HEAVY LDADS NEAR SPENT 06/27/83 C GL 25-11 ENDED FUEL-A 79 DETERMINAi!ON CF SAFETY RELIEF :
/ /
NC SEE A-07 VALVE POOL DYNAMIC LOADS AND
. TEMPERATURE LIMITS A-40' - SEISMIC DES!6N CRITERIA -
/ /
NC SUBSUMMED BY A-46 SHORT-TERM PROGRAM A-4 P!PE CRACKS IN BOILING WATER
/ /
C FEACTORS.
- A-43.CCNTAINMENT EMER6ENCY SUNP
/ /
NC INFO ONLY FERFORMANCE A-44 STATION PLACK 0UT 06/30/92 I SER 6/30/90
- A-45 SHUTDOWN DECAY HEAT REMOVAL
/ /
NC SUBSUMED BY SEVERE ACC FEQUIREMENTS
.A-46 SE!SMIC QUALIFICAi!DN OF 12/31/93 i RED UNDER DEVEL EQUIPMENT IN OPERATING PLANTS L
A-47 SAFETY IPPLICATIONS OF CONTROL 03/31/90 i NEW REGUIREMENTS SYSTEMS o
-'A-40 HYDROGEN CONTROL MEASURES AND
/ /
NC
!NERTED PUT NO CAD SYS EFFECTS OF HYDROGEN BURNS ON SAFETY EQUlFMENT L
A-49 FRESSURIZED THERMAL SHOCK
/ /
N/A PWR ONLY
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