ML20098H040
| ML20098H040 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 10/05/1984 |
| From: | Cain J LOUISIANA POWER & LIGHT CO. |
| To: | Eisenhut D Office of Nuclear Reactor Regulation |
| References | |
| W3A84-0133, W3A84-133, NUDOCS 8410090304 | |
| Download: ML20098H040 (110) | |
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POWER & LIGHT / NEW OALEANS. LOUSIANA 70160 9 l504] 59S-22O4 MIDDLE SOUTH UTtuTIFS SYSTEM October 5, 1984 J.M. CAIN
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President and Chief Executive Officer W3A84-0133 Mr. Darrell G. Eisenhut, Director Division of Licensing Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C.
20555
Dear Mr. Eisenhut:
On June 28, 1984, I wrote you stating that I did not intend to request a fuel loading / low power license or a full pcwer license until I am personally satisfied that all issues necessary for those phases of plant operation have been satisfactorily addressed to assure the public health and safety.
Since then, we have implemented our review program for evaluating, responding, and taking corrective action on the twenty-three issues raised in your letter of June 13, 1984, including a formal process for review by the independent Task Force and by a special subcommittee of the Waterford 3 Safety Review Committee. We have made considerable progress in this program with nearly all of the work having been completed. Much of the remaining efforts involve formal verification of the work that has been completed.
As of this date, we have filed with the NRC responses to seventeen of the twenty-three issues, with the remainder expected to be ready for submittal in the next several weeks. Those responses are subject to validation by the Tar.k Force which is reporting directly to me.
During the period when we haie been engaged in those efforts, we have also been able to accomplish and complete a number of other plant-readiness items. On April 25, 1984, we wrote Mr. Denton that Waterford 3 had essentially been completed in accordance with the application, and that the plant would be ready to load fuel by May 30, 1984. Since then, we have made considerable progress in improving plant readiness for fuel load and power operation beyond the point of mininum requirements.
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Mr. D::rr311 G. Eircnhut Paga 2 Examples of this are listed below:
1)
We have licensed five senior members of the LP&L Waterford 3 Staff, all of whom have extensive previous commercial experience.
2)
Work on our full flow condensate polisher system has progressed to the point where it will now be available early in our testing program, possibly prior to heatup.
]
This will be an improvement in our ability to optimize i
Steam Generator Chemistry Control.
s 3)
Work on a modification to install an Auxiliary Feedwater Pump and associated piping controls will also be complete early in our startup.
This also will help post heatup chemistry control and allow us to stay with our design concept of saving the Emergency Feedwater System and
. pumps for off-normal and emergency use only.
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4)
Of the original 116 systems designated by LP&L as required to be transferred to Plant Staff control prior t
to receipt of a full power License, only three remain to be transferred. These are System Supports (Hangers),
Whip Restraints, and Seismic Support.
5)
Eighty-seven percent of Mode 6 (Refueling Mode) and 53%
of Mode 3 (Hot Standby Mode) Standard Technical Specification required surveillances are now complete.
This has allowed us to field check our Technical Specifications and related procedures ahead of time and puts us in a good position for compliance with those important provisions of an Operating License.
6)
The major portion of important operations programs such as Work Control, Plant Security, Plant Planning and Scheduling, Station Modification (Design Control), and Licensee Event Reporting have now been implemented for several months. This has allowed much of the " debugging" process which goes on early in most licensed plant operations to already have taken place prior to licensing.
Although the plant is now essentially complete, and is ready for fuel load and the startup program, the implementatiori of a phased licensing
. program will allow the initiation of activities which present no risk to the public health and safety while the Task Force completes its validation process and the NRC completes its review of matters pertinent to the protection of the public health and safety.
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i Mr. D;rrall G, Ein nhut P;go 3 To facilitate a phased Licensing approach, the Waterford 3 plant operating organization has initiated a review of the impact of the twenty-three issues on each plant system to determine the significance of t
those issues cn each system which is required to be operable during:
a)
Fuel loading and pre-criticality post-core load hot functional testing, and b)
Criticality and low power testing up to 5% of full power, and power ascension up to full power.
We have identified in Attachment A hereto, among other things, all of the plant systems required by the Technical Specifications to be operable during fuel loading and pre-criticality post-core load hot functional testing. Attachment B sets out all remaining plant safety systems required to be operable for criticality and low power testing up to 5% of full power, and for full power operation. For each plant system set out in Attachments A and B, we have teamed personnel familiar with that system with other personnel who have become knowledgeable with respect to the twenty-three issues. The system knowledgeable members are from the plant operating organization. This not only lends an additional dimension to the review, but also assures that the operating staff is fully involved and familiar with the issues and their resolution. These teams have been l
charged with performing a safety review of each system paired with each of the twenty-three issues to determine what actions, if any, may be required to dispose of those issues as they may apply to the safe operation of each I
system.
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Each safety review is further reviewed by the Plant Operations Review l
Committee and the Plant Manager in the same manner that all other safety l
reviews during plant operation will be performed.
Attachment C hereto sets out the Program Plan for this procedure and a flow chart for the review process. The proper execution of this process I
will be audited by the Waterford 3 Independent Safety Engineering Group.
I The Safety Review Committee (SRC) Subcommittee is also participating in the review of the Safety Reviews in the traditional SRC role of management level overview of the PORC review of safety reviews and evaluations.
A prerequisite for fuel loading and performance of the pre-criticality testing under a phased licensing program is the completion of the above described systematic safety review of the systems in Attachment A.
Similarly, criticality and low power (5%) testing would be premised on com-pletion of the review process for the systems identified in Attachment B.
Since the two appendices include all plant safety systems, including those required for full power operation, we expect that all twenty-three issues will be resolved at least with respect to their safety review impact on plent systems in all modes, including full power operation, prior to issuance of criticality / low power authorization.
Mr. D;rrall G. Ein nhut
'P:g3 4 The independent Task Force, set up to review the resolution of the twenty-three allegation issues, recommended that the plant operating organization review the impact of the twenty-three issues on each plant system to determine the safety significance for various operating conditions. The process outlined above provides for a multilevel review, by qualified people, of the significance of the twenty-three issues on the plant hardware in a systematic review process to assure the applicable technical specifications can be met prior to each licensing phase.
Based on the seventeen responses to the twenty-three issues already filed with the NRC, the progress we have made so far on the remaining issues, and the information set out in Attachment A, we have concluded that we can now request a limited license from the NRC authorizing fuel loading and pre-criticality, post-core load hot functional testing.
In addition to setting out the plant systems necessary for this phase of operations, Attachment A contains summaries of the evaluations completed to date of the safety significance of the twenty-three issues as they may affect those i
systems. A complete file of the safety reviews is available for NRC staff I
review. This review is very nearly complete, and the findings are providing a high confidence level that significant safety problems are not present.
j Accordingly, we respectfully request the NRC, upon our satisfactory completion of the safety reviews for each of the systems identified in Attachment A, and upon satisfactorily addressing the construction and system items set out in Attachment D, to issue the limited license. The items listed in Attachment D are extracted from the NRC Region IV NTOL report listing of open items required to be completed prior to fuel loading and pre-critical testing. These items are largely unrelated to the twenty-three issues. We expect to confirm to you our completion of the Attachment A reviews by October 19, 1984 and Attachment D items by October 31, 1984.
f The execution of the fuel loading and pre-critical post-core load testing activities requested in this phased licensing program will present no hazard to the public health and safety and will not significantly interfere with the resolution of any of the twenty-three issues that may
)
yet be outstanding and which may impact on plant systems not required for i
this phase of the plant startup program.
j Because the reactor will not attain criticality in the initial phase
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of the license, there will be no f.ission products generated and no decay heat. Consequently, the requested activities will pose no risk to the public health and safety.
It is on this basis that the NRC Staff recently issued a fuel loading license for the Catawba Nuclear Station, Unit 1.
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' Mr. Darrall G. Eissnhut
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'Bc' fore requesting further authority to proceed with criticality and the power ascension' program, we will complete a sitt;lar safety review of
- the-significance of the twenty-three issues on the remaining plant systems set out in. Attachment B.
1 Thank you for your consideration of this request.
Sincerely, J.M. Cain
Enclosures:
As Stated cc- (with Enclosure):
R.S. Leddick, D.E. Dobson, K.W. Cook, J.T. Collins (NRC), D. Crutchfied (NRC),
G. Knighton (NRC), G. Charnoff, L.L. Humphreys, R.L. Furgeson, S. Levine, L. Constable (NRC),
Project Files.
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ATTACHMENT A SAFETY REVIEWS OF PLANT SYSTEMS REQUIRED BY TECHNICAL SPECIFICATIONS FOR FUEL LOADING AND PRECRITICALITY p-i, f- -
W-POST-CORE LOAD HOT FUNCTIONAL j '.
TESTING 4
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LICENSING PLAN FOR FUEL LOADING AND PRECRITICALITY POST CORE LOAD HOT FUNCTIONAL TESTING A Licensing Program Plan has been structured to institute safety reviews of those plant systems required for fuel load and post fuel load testing, criricality and low power testing (to 5% power) and full power operation.
As discussed in the description of the safety review process (Attachment C), a detailed review of the technical specifications was performed to determine the listing of plant systems required 'for the initial phase of the limited license-(Table A-1).
Forty-nine plant. systems have been identified as being required to be operable by Waterford SES #3 technical specifications in modes 6, 5, 4, 3 (refueling through hot standby) and these systems are the subject of this Attachment (Attachment A).
These.are the modes involved with fuel load and pre-criticality, post fuel load hot functional testing.
This is a conservative approach because many of these requirements assume the presence of irradiated fuel and therefore are not
=0 of significance to the initial core loading and testing processes.
This program will assure LP&L management that the impact of any concern raised is properly assessed and resolved in the context of safe plant operations and-protection of the public health and safety.as will be specified in our operating license / standard technical specifications and FSAR.
Safety reviews were performed on each of the plant systems in Table.A-1, using the procedures described in Attachment C, against each of the 23 issues'(Table A-2).
Table A-3 provides a matrix indicating those safety reviews wnich have been successfully completed. Table A-4 provides the footnotes associated with the Table A-3 matrix indicating outstanding actions required to complete the matrix.
Where successful completion of the safety review is indicated in Table A-3, the safety review assures completion of those actions necessary to insure the system is constructed and functions according to the requirements of the FSAR in light of the 23 issues, without consideration of the lack of fission products (due to not having gone critical).
Should it become necessary to perform limited safety reviews (credit must be taken for lack of fission products in order to justify safety significance) in order to complete the review on specific systems the matrix will reference a footnote describing the circumstances and basis for the limited review.
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A-1
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' During the safety evaluation of these 49 fuel load systems they were categorized into subgroups that logically represent the potential issue by issue-safety! impact.'The subgroups are defined in Table A-6 as:
A_.
The issue does not have a safety related effect:on the system because:
a)L the contractor in question did not do work on the system under
)
evaluation, or b) ' the procedure or process in question did not apply to the system i
under evaluation.
B.
The issue does not have a safety related effect on the system because:
a) the contractor in question did not do any safety related work t
on the system under evaluation, or the procedure or process in question did not apply to any safety related portions of the system under evaluation, and b) any non-safety related activities performed on the system of concern does not have any significant effect on the' safety related function of the system under evaluation.
, _. The issue does have a potential safety related effect on the system C~
because:
a) the contractor in question did work of safety significance on the l-system under evaluation, or i
(
b) the procedure or process in question did apply to safety significant activities of the syseem under evaluation.
Safety evaluations were performed and' verified (as necessary) to assure LP&L management that Waterford SES #3 can be' safely operated without i
compromising the health and safety of the public. The subgroup for each system, as it relates to each of the twenty-three issues, is presented in Table A-6. In performing the evaluations, it was determined that it would be more effective to subdivide the first issue (Inspection Personnel Issues) into three subissues covering 1A - Mercury, IB - Thompkins-Beckwith 4
and 1C - Other Contractors. This resulted in effectively 25 issues being evaluated for each of the 49 plant systems.
Since this results in a total I
' of 1225 safety reviews (each consist'ing of several pages) it is not feasible to present all of the documentation in this transmittal. The full documentation of the safety reviews is on file at the Waterford SES #3 l
. On-site Licensing Unit offices for inspection and review by the NRC staff, i
The individual safety reviews were reviewed and summaries prepared, for,
l those falling within Subgroup C.
The summaries are included in this attachment-(Table A-5) for each issue and subissue.
In order to indicate the level of review performed in this process as well as to provide a correlation with the sunsmaries provided in Table A-5, several examples of the safety reviews are included following Table A-6.
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A-2
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TABLE A PLANT SYSTEMS REQUIRED BY TECHNICAL SPECIFICATIONS DURING
. FUEL LOADING AND PRE-CRITICAL POST-CORE LOAD HOT FUNCTIONAL TESTING MODE OPERABILITY t
-ACRONYM SYS. NO.
DESCRIPTION IS REQUIRED
'DC
.02A 125v DC SAFETY MODE l-6 MT 03 SWITCHING STATION MODE l-6 ST 04 STARTUP TRANSFORMERS MODE 1-6 s
4ky 06A-4.16kv ELEC. DISTRIBUTION MODE l-6 SAFETY SSD
'07A 480v EL'EC. DISTRIBUTION SAFETY MODE l-6 LVD 08A 208/120v ELEC.' DISTRIBUTION MODE 1-6 SAFETY-
.ID 09A INVERTERS & DISTRIBUTION MODE 1-6 4
SAFETY t
10 COMMUNICATIONS MODE 1-6
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HT 13A-1 HEAT TRACE SAFETY MODE 1-6 EM 16 ENVIRONMENTAL MONITORING ALL MODES 4
1 SM 17 SEISMIC MONITORING ALL MODES b
ARM /RMC/
18-1 RADIATION MONITORING SYSTEM ALL MODES PRM 18-2 18-3 18-4 j.
18-5 SS 20 SECURITY SYSTEM ALL MODES FPD 21 FIRE DETECTION ALL MODES FP 22 FIRE PROTECTION ALL MODES CC 36-1 COMPONENT COOLING WATER MODE 1-6
.36-2 ACC 36-3 AUXILIARY COMPONENT COOLING MODE l-4 WATER l
EG 39 EMERGENCY DIESEL GENERATOR MODE 1-6 CRN 40-2 CRANE & HOIST FHB MODE 6 ONLY CCS 43A RCB CONTAINMENT COOLING MODE l-4 i
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SBV 43B SHIELD BLDG. VENTILATION MODE l-4 A-3 S
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TABLE A-1' PLANT SYSTEMS-REQUIRED BY TECHNICAL SPECIFICATIONS'DURING FUEL LOADING AND PRE-CRITICAL POST-CORE LOAD HOT FUNCTIONAL TESTING.
MODE OPERABILITY ACRONYM.
'SYS.' NO.-
DESCRIPTION' IS REQUIRED
.CVR-43E CONTAINMENT VACUUM RELIEF MODE l-4 s.
.HVC 46B CONTROL ROOM HVAC ALL MODES HVR 46D RAB HVAC MODE l-6 CHW 46E RAB CHILLED WATER MODE 1-6
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FP 46K ' ' FIRE DAMPERS' ALL MODES CB 48 LRT CONTAINMENT. VESSEL MODE l-6 PAC 49 PROCESS ANALOG CONTROL MODE 1-6 IC-50B MISC. PANELS MODE 1-6 RCS 52A REACTOR COOLANT SYSTEM MODE 1-6 52B
-52C c
.CVC 53A-CHARGING & LETDOWN MODE l-6 BAM 53B BORIC ACID MAKEUP MODE l-6 PSL.
54-9 PRIMARY SAMPLING MODE 1-5 GWM.
55A CASEOUS WASTE MANAGEMENT ALL MODES LWM
'SSB LIQUID & LAUNDRY WASTE ALL MODES SSE MANAGEMENT SI 58 SAFETY INJECTION MODE 1-6 60A 60B 60C CS 59 CONTAINMENT SPRAY MODE l-4 FHS 61 FUEL HANDLING & STORAGE MODE 6 ONLY PPS 66 PLANT PROTECTION SYSTEM ALL MODES 63-
~ENI 65A-1 EXCORE NUCLEAR INST.
MODE l-6 65A-2 CMU 71B CONDENSATE MAKEUP MODE l-3 EFW 73 EMERGENCY FEEDWATER MODE 1-3 i
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TABLE A-1 PLANT SYSTEMS REQUIRED BY TECHNICAL SPECIFICATIONS DURING FUEL LOADING AND PRE-CRITICAL POST-CORE LOAD HOT FUNCTIONAL TESTING MODE OPERABILITY ACRONYM SYS, NO.
DESCRIPTION IS REQUIRED
-SSL 75 SECONDARY SAMPLING MODE 1-4 SG 76 STEAM GENERATORS & MSIV M0'DE 1-4 TUR 88 TURBINC & TURBINE CONTROLS MODE 1-3 91 SEISMIC SUPPORTS ALL MODES 19-16 WHIP RESTRAINTS ALL MODES 19-17 SYSTEM SUPPORTS (HANGERS)
ALL MODES SEISMIC STRUCTURES ALL MODES f
O A-5
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TABLE A-2 SAFETY REVIEW ISSUES k
ISSUE NO.
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1 (A) Inspectiori Personnel Issues - Mercury
_(B) InspectioniPersonnel Issues - T&B
-(C) Inspection Personnel Issues - Other Contractors
-2 Missing NI Instrument Line Docuinentation.
3 Instrumentation. Expansion Loop Separation 4
Lower Tier Corrective Actions are not being Upgraded to NCRs
.S-Vendor Documentation - Conditional Releases
~6 Dispositioning of Nonconformance and Discrepancy Reports 7
hackfill Soil Densities
-8 Visual' Examination of Shop Welds During Hyrdrostatic Testing l
9 Welder certification 10 Inspector Qualifications (J. A. Jones & Fegles) 11 Cadwelding 12 Main Steamline Framing Restraints 13 Missing NCRs
- 14 J. A. Jones Speed Letters and EIRs 15 Welding of "D" Level. Material Inside Containment 16 Surveys and Exit Interviews of QA Personnel f
.17 QC Verification of Expansion Anchor Characteristics 18 Documentation of Walkdowns on Non-Safety Related Equipment 19 Water-in Basemat Instruments
'20 Construction Materials Testing (CMT) Personnel Qualification Records 21 LP&L QA Construction System Status and Transfer Reviews 22 Walder Qualifications (Mercury) and Filler Material Control (Site Weld) 23 OA Program Breakdown Between Ebasco and Mercury N
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' TABLE.A-4 SYSTEMS / ISSUES SAFETY REVIEW RESOLUTION MATRIX
-FOOTNOTES NOTE OUTSTANDING ACTIONS
-(l)
ISSUE lA Reinspect all N1 instrument loops. installed by Mercury (2)
ISSUE 1B Verify the qualifications of the initial Tompkins-Beckwith inspectors l
Where the initial qualifications of the^above are inadequate, verify the qualifications of the inspectors performing any over inspections l
l Where the above are not met, reinspect or justify on a case-by-case basis 4
.l (3)
ISSUE IC Allows for the inspection of additional contractors (4)
ISSUE 2 A CIWA has been initiated to rework instrumentation and correct documentation (isometric drawings) to remove class breaks in tubing from ASME III to ANSI B31.1 (5)
CIWA's have been issued (content as above)
(6)
A CIWA has been issued (content as above)
(7)
A CIWA has been issued (content as above)
(8)
A CIWA has been issued (content as above)
(9)
A CIWA has been issued (content as above)
(10)
A CIWA has been issued.(content as above) 1 (11)
ISSUE 3 Completion of a CIWA to correct tube track for PT-CA-675SAS & BS,
\\
i
\\
\\
A-11
a TABLE A-4 p
SYSTEMS / ISSUES SAFETY REVIEW RESOLUTION MATRIX
/
FOOTNOTES
~
NOTE JOUTSTANDING ACTIONS "N
ISSUE 4 NONE
\\
ISSUE 5 NONE ISSUE 6 NONE ISSUE 7 NONE ISSUE 8 NONE (12)
ISSUE 10 Verify qualifications of initial inspectors Where the initial inspector qualifications are inadequate; verify the qualifications of any over-inspection or co-signing by ECI Where the above are not met, reinspect or justify on a case-by-case basis ISSUE 11 NONE ISSUE 12 NONE (13)
ISSUE 13 (Applies to all identified systems)
Missing and voided Mercury NCR's are currently being reviewed.
A-12
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h TABLE A-5 SAFETY REVIEW SUMMARIES 9
Y 9-e O
A-14
cc.
Issue #1 - Inspection Personnel Issue _s_
i
'This issue was evaluated on a contractor basis.
Issue'i1A - Mercury Subgroup C-- Mercury did perform safety related work on the system and safety evaluations are.being performed to assure LP&L management that Waterford Steam Electric Station #3 can be safely operated without compromising the
- health and safety of the public.
Issue f1'does have a potential effect on:
' System #-
System Description
Evaluation 18-3 Radiation Monitoring System Installation of safety related instrumenta-22 Fire Protection tion was inspected by potentially unqualified 1, Component Cooling Water inspectors.
The quality of safety related instru-36-2--
. Component Cooling Water
. mentation associated with this system shall be 36-3' Aux. Component Cooling verified. Verification is Water being accomplished by reinspection of N1 instru-39 Emetgency Diesel Generator ment loops.
Satisfactory completion of this sample 43A RCB Containment Cooling of Mercury installations will be the basis for 43B.
Shield Bldg. Ventilation acceptance of the remaining installations.
43E Containment Vacuum Relief 46B Control Room HVAC 46D RAB HVAC 46E RAB Chilled Water-52A Reactor Coolant System 52B'
~52C Reactor Coolant System 53A Charging & Letdown
~53B Boric Acid Makeup A-15
System i
System Description
SSA' Gaseous Waste Management SSB
-Liquid Waste Management
-58 Safety Injection 60A Safety Injection
-O 60B Safety Injection 60C' Safety Injection 59-
' Containment Spray 66 Plant Protection System 63 Plant Protection System 71B Condensate Make-up
.73 Emergency Feedwater 76 Steam Generator and MSIVs L
A-16 s
(_
l' f
i Issue'#IB - Tompkins-Beckwith b
. Subgroup C - Tompkins-Beckwith did perform safety related work on the system, and safety evaluations are being performed to assure LP&L management that Waterford SES #3 can be safely operated without compromising the health and safety of-the.public.
' Issue il does have a potential effect on:
- System #
System Description
Evaluation 18-3 Radiation Monitoring Work performed on this system was inspected by 22 Fire Protection potentially unqualified inspectors.
To close out 36-1 Component Cooling Water the concern LP&L shall verify the qualifications 36-2 Component Cooling Water of the initial inspectors.
LP&L shall verify the 36-3 Aux. Component Cooling qualifications of the Water inspectors performing any over-inspection. A 39 Emergency Diesel Generator determination for any reinspection will be 43B Shield Bldg. Ventilation evaluated on a case-by-case basis.
43E Contcinment Vacuum Relief 46B Control Room HVAC 46D RAB HVAC 46E RAB Chilled Water 48 LRT Containment Vessel 52A Reactor Coolant System 52B Reactor Coolant System 52C Reactor Coolant System 53A Charging and Letdown 53B Boric Acid Makeup 54-9 Primary Sampling SSA Gaseous Waste Management SSB Liquid and Laundry Waste Management I
1 A-17
I
/
i i
System #
System Description
I-SSE Liquid and Laundry Waste Management i
58
. Safety Injection
-60A Safety Injection l
60B Safety Injection 60C Safety Injection 59 Containment Spray 61 Fuel Handling and Storage
'65A-1 Excore Nuclear Instrument 7IB Condensate Make-up.
73 Emergency Feedwater 76 Steam Generator and MSIV 88 Turbine and Turbine Controls 19-16 Whip Restraints 19-17.
System Supports A-18
Issue #1C -'Other Contractors The safety evaluation is being performed and will be finalized later.
O e
S A-18a
Issue.#2 - Missing N1 Instrument Line Documentation
_ Subgroup C - Instrumentation installations that were identified to have adequate documentation to support the quality of the installations but a decision was made to rework the installations to comply with ASME III
- documentation requirements are contained-in this system and a safety evaluation-was performed to assure LP&L management that Waterford SES #3 can be safely operated without compromising the health and safety of the public.
Issue.#2 does have an effect on:
System i
System Description
Evaluation 36-1
. Component Cooling Water These systems are directly affected by systems #36-3, 36-2 Component Cooling Water
~ and #39 and therefore instrument rework and documentation correction is required
- to demonstrate system operability.
36-3.
Aux. Component Cooling These systems require
- Water instrument rework to correct documentation to 4
39 Emergency Diesel Generator demonstrate system operability and remove 43B Shield Building Vent 11a-tube class bceaks from-tion from ASME III to ANSI B31.1.
66Property "ANSI code" (as page type) with input value "ANSI B31.1.</br></br>66" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process. Plant Protection System 63 Plant Protection System 73 Emergency Feedwater 76 Steam Generator and MSIV i
- The removal of tubing class breaks was not specifically required due to lack of documentation, but was decided upon to assure timely closure of the issue. The safety review assumed this action was necessary for conservatism.
A-19 i
.3
[
)
Issue #3 - Instrumentation Expansion Loop Separation Su'ugroup C - It has been determined that there is identified installation deficiency regarding tubing separation criteria in the system and a safety evaluation was performed to assure LP&L management that Waterford SES #3 can be safely operated without compromising the health and safety of' the public.
Issue #3 does have a potential effect on:
~
System i
System Description
Evaluation 66 Plant Protection System New tube, tracks and.
supports were installed to 63 Plant Protection System correct the-deficiencies.
Accordingly, this issue does not serve as a constraint to the safe operation of these systems, and has been resolved and closed out by LP&L.
l A-20 a
G L
Issue #4 - Lower Tier Corrective Actions Are Not Being Upgraded to NCR's Subgroup C - DCN's.'FCR's, EDN's and T-B DN's have been reviewed and it was determined that some documents should have been upgraded to NCR's.
A safety
. evaluation was performed to assure LP&L management that Waterford SES #3 can be safely operated without compromising the health and safety of the public.
Issue #4 does have a potential effect on all systems in Table A-6.
The Evaluation reveals that a statistically acceptable number of lower tier documents were reviewed showing no significant quality impact (no cases were
. detected which were safety significant and would be reportable under 10CFR50.55e). Therefore it is possible to conclude with a 95% confidence
' level that 95% of the unsampled documents contain no significant deficiencies.- Accordingly, this issue does not serve as a constraint to safe operation of the systems.
6 6
d*
O A-21
Issue 15 - Vendor Documentation - Conditional Releasesl
\\
Subgroup C - With a review of QA/QC records it is concl'uded that there are no unresolved items which affect the systems, however Issue #5 does have a potential effect on all systems in Table A-6.
The Evaluation reveals that during the review of QA/QC re' cords conditional release items which affected systems were evaluated and closed out by LP&L with receipt of the " unconditional" paperwork.
No items exist to affect tihe
' safety function of the systems.
O N
A-22
o Issue #6 - Dispositioning of Non-Confotiaance and Discrepancy Reports Subgroup C - It was noted during a review of NCR's that some of the reports had questionable dispositioning potentially rendering the quality of installation indeterminate.
Issue #6 does have a potential effect on all systems in Table A-6.
The Evaluation included a combination screening and sampling method to review
' ERASCO NCR's including NCR's identified by the NRC and no items were
' identified which had significant safety impact on the systems. Mercury NCR's were reviewed for upgrade and sampled to determine reportability to support the conclusion that the safety review is not effected.
O e
p
=
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A-23
/
I
/
I I.
Issue.#7 - Backfill Soil Densities t
l
- Subtroup C - Data frcmi the in-place density tests on the class A fill was
_potentially not traceable relative to the technical adequacy of the l
placements, ' therefore the -impact on the the quality of the system may have been indeterminate. A safety evaluation was performed to assure LP&L management.that Waterford SES #3 can be safely operated without ceripromising the health and safety of the public.
- Issue #7 does have a p'otential effect on all systems in Table A-6.
The Evaluation reveals that the data for the in-place density tests performed
.on the class A. fill has been located and has been transmitted to the QA records vault. ' Peview and analysis of the records indicates that the Class A backfill soil densities are in accordance with specifications and FSAR
-requirements except for analytically non-significant deficencies and does provide the required design structural capacity for the plant under seismic loadings.- Accordingly, this issue does not serve as a constraint to safe operation of the system, and has been resolved and closed out by LP&L.
e' 4
5 e
A-24
h Issue #8 - Visual Fxamination of Shop Welds During Hydrostatic Testing Subgroup C - The system does include ASME Class 1 & 2 welds (shop and field) that were inspected during total system hydro in the field. A safety evaluation was performed to assure LP&L management that Waterford SES #3 can be safely operated without compromising the health and safety of the public.
Issue #8 does.have a potential effect on:
System #
System Description
Evaluation 18-1 Radiation Monitoring System ASME Class 1 & 2 welds (shop and field) were 18-2 Radiation Monitoring inspected and documented on ASME N-5 code data reports 18-3 Radiation Monitoring
'during total system hydro in the field.
The ASME 18-4 Radiation Mcnitoring Class 1 & 2 welds (shop and field) were tested and 18-5 Radiation Monitoring inspected in accordance with ASME code, in the 36-1 Component Cooling Water field.
There is no devia-tion from FSAR require-36-2 Component Cooling Water ments. Accordingly, this issue does not serve as a 36-3 Aux. Component Cooling restraint to safe operation Water of these systems, and has been resolved and closed 52A' Reactor Coolant System out by LP&L.
52B Reactor Coolant System 52C Reactor Coolant System 53A Charging And Letdown 53B Boric Acid Makeup 54-9 Primary Sampling 55A Gaseous Waste Management SSB Liquid and Laundry Waste Management SSE Liquid and Laundry Waste Management 58 Safety Injection A-25 s
~
'Systes f
System Description
60A-Safety Injection 60B Safety Injection 60C Safety Injection 59 Containment Spray 0
715 Condensate Makeup 73 Bnergency Feedwater.
76 Steam Generator and MSIV O
O l-A-26 s
I R____.--___________--.__._.
i M. g.
Issue #9 _ Welder certification The safety evaluation is being performed and will be finalized later.
O
)
A-26a
.e Issue #10 - Inspector Qualifications - (J.A. Jones and Fegles)
Subaroup C - J.A. Jones and Fegles were responsible for the construction 'of the basemat and all structural concrete on the basemat. A safety evaluation was performed to assure LP&L management that Waterford SES #3 can be safely l,
operated without compromising the health and safety of the public.
Issue #10 does have a potential effect on:
System #
System Description
Evaluation Seismic Structures To close out the concern LP&L shall verify qualifi-cations of the initial inspectors involved.
LP&L shall verify the qualifications of the inspectors performing any over-inspection. A deter-l' mination for any reinspec-tion will be evaluated on a case-by-case basis.
l l
h s
A-2f L
Issue #11 - Cadwelding Subtroup C - Data from the cadweld testing program was potentially not traceable relative to the technical adequacy; therefore the impact on the system could have been indeterminate. A safety evaluation was performed to assure LP&L management the Waterford SES No. 3 can be safely operated without compromising the health and safety of the public.
Issue #11 does have a potential effect on all systems in Table A-6.
O The Evaluation of cadweld records concluded that discrepancies noted were not significant to safety and would not have had any effect on the structural capability of the NPIS during operation and safe shutdown. The probability of an accident previously evaluated in the FSAR is not increased.
Accordingly, this issue does not serve as a constraint to the safe operation of the systems, and has been resolved and closed out by LP&L.
W e
6 e
e 4
4 l
A-28
- ~
_.. _ ~ _ - _. _ - - -.
/
i i
Issue #12 - Main Streamline Framing Restraints Subaroup C - Apparent failure to inspect the installation of the main streamline framing restraints may rendered the quality of the system indeterminate! A safety evaluation was performed to assure LP&L management that Waterford SES #3 can be safely operated without comprocising the health and safety of the public.
Issue #12.does have a potential effect on:
System #
System Description
Evaluation 76 Steam Generators and The deficiencies noted MS1V during the reinspection have been corrected and 91 Seismic Supports all hardware corrective actions have been completed 19-16 Whip Restraints and verified by LP&L.
Accordingly, this issue 19-17 System Supports does not serve as a
'Mansers) constraint to safe operation of these systems.
Seismic Structures and has been resolved and closed out by LP&L.
1 s
4 O
s l
I A-29
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t 1
Issue #13 - Missing NCRs Subtroup C - It was noted that there were' missing reports in the consecutively numbered EBASCO and Mercury NCRs implying missing NCRs that may have rendered system quality indeterminate. A safety evaluation was performed to assure LP&L management that Waterford SES #3 can be safely operated without compromising the heath and safety of the public.
Issue #13 does have a potential effect on all systems in Table A-6.
The Evaluation includes a review program to evaluate missing and voided Mercury NCR's.
O A-30 m
Issue #14 - J.A. Jones Speed Letters and EIRs Subaroup C - Contractors' performing safety related work generated EIRs and Speedy Menos which transmitted design information that could potentially affect system quality. A safety review was performed to assure LP&L management that the system can be safely operated without compromising the health and safety of the public.
Issue #14 does have a potential effect on all systems in Table A-6.
The' Evaluation included a sampling program to evaluate informal documents requesting engineering information from safety related contractors.
Of all the samples reviewed those that resulted in design change deficiency had no safety significance. The program provides reasonable assurance that informal
' documents were not used to transmit design changes which have safety significance.
L A-31 i
s
_______________-__.-_-_________________J
Issue #15 - Welding of "D" Level Material Inside Containment Subaroup C - Class "D" material installation inside containment does have a potential effect on:
System #
System Description
Evaluation 08A 208/120v Elec. Distribution During the evaluation of Safety Class "D" material installation inside 17 Seismic Monitoring containment the work and
~~~'_.__ material under review was 18-1 Radiation Monitoring verified by LP&L.
System Contractor QA is of satisfactory quality, and 18-2 Radiation Monitoring this issue does not have System an adverse effect on the safety analysis, system
(
18-3 Radiation Monitoring operability or margin to
[
System safety on these systems.
18-4 Radiation Monitoring System 18-5 Radiation Monitoring System 21 Fire Detection 22 Fire Protection 36-1 Component Cooling Water 36-2 Component Cooling Water 40-2 Crane & Hoist FHB 43A RCB Containment Cooling 43E Containment Vacuum Relief 48 LRT Containment Vessel l
1 52A Reactor Coolant System
$2B Reactor Coolant System 52C Reactor Coolant System 53A Charging & Letdown 54-9 Primary Sampling A-32
J >
System #
System Description
.58 Safety Injection 60A Safety Injection 60B Safety' Injection 60C-Safety -Inj ection 59 Containment Spary 61 Fuel Handling & Storage
.65A-1 Excore Nuclear Inst.
-Excore Nuclear Inst.
-715 Condensate Makeup 76 Steam Generators & MSIV 91 Seismic Supports 19-16 Whip Restraints 19-17 System Supports (Hangers)
Seismic Structures k
i A-33
' Issue #16 - Surveys and Exit Interviews of QA Personnel Subaroup C - An interview program was insti uteil by LP&L to provide an additional avenue of communication to elicit'information on quality concerns from personnel prior to leaving the Waterford SES No. 3 project. The concern was that the LP&L program may not have promptly or thoroughly examined the specific areas of concern and the programmatic implications of these systems.
Issue #16 does have a potential effect on all systems in Table A-6.
- The Evaluation reveals that all concerns are being reviewed under an improved quality concern program.
Where there are issues not previously identified with potential safety related consequences, these issues are promptly reported'to LP&L management. These concerns are properly addressed under LP&L required and approved management programs in a timely fashion. The program does not involve unreviewed safety issues.
p 4
4 A-34
Issue #17 - QC Verification of Expansion Anchor Characteristics Subgroup C - Mercury, the subject of this concern, did install safety related instrumentation expansion anchors in these systems.
A safety evaluation was performed to assure LP&L management that the system can be safely operated without compromising the health and safety of the public.
Issue #17 does have a potential effect on:
System f.
System Description
Evaluation 18-1 Radiation Monitoring Inspection forms were used 18-2 System that do not explicitly 18-3 cover all inspection i
18-4 attributes. Verification 18-5 of acceptability will be accomplished via the 36-1 Component Cooling Water reinspection program.
36-2 Component Cooling Water 36-3 Aux. Component Cooling Water 39 Emergency Diesel Generator 43A RCB Containment Cooling 43B Shield Bldg. Ventilation 43E Containment Vacuum Relief 46B Control Room HVAC 46D RAB HVAC 46E RAB Chilled Water 50B Misc. Panels 52A-Reactor Coolant System 52B Reactor Coolant System 52C Reactor Coolant System 53A Charging and Letdown 53B.
Boric Acid Makeup 55A Gaseous Waste Management A-35
. ~..
/i
/
/
i i
Issue #18 - Documentation of Walkdowns on Non-Safety Related Equipment Subgroup C - Documentation of walkdown on non-sa'fety related equipmlent does have a potential effect on:
i System #
System Description
Evaluation 02A.
125v DC Safety Area inspections where the system is present indicate 06A 4.16kv Elec.
no interactions of safety l
Distribution Safety significance. Accordingly, this issue does not serve 07A 480v Elec.
as a re: raint to safe Distribution Safety operatie of these systems, and has on resolved and 08A 208/120v Elec.
closed out by LP&L.
Distribution Safety 09A Inverters &
Distribution Safety l
10 Communications 13A-1 Heat Trace Safety J'
16 Environmental Monitoring 17 Seismic Monitoring i
L 18-1 Radiation Monitoring System 18-2 Radiation Monitoring System 18-3 Radiation Monitoring System 18-4 Radiation Monitoring System 18-5 Radiation Monitoring System 20 Security System 21 Fire Detection 22 Fire Protection A-36
l f
/
}
i System #
System Description
l 36-1 Component Cooling Water 36-2 Component Cooling Water 36-3.
Aux Component. Cooling Water 39
' Emergency Diesel Generator 40-2 Crane & Hoist FHB 43A RCB Containment Cooling 43B Shield Bldg. Ventilation 43E Containment Vacuum Relief
'46K Fire Dampers 48
.LRT Containment Vessel 49 Process Analog Control 50B Misc. Panels 52A Reactor Coolant System 52B-Reactor Coolant. System 52C Reactor Coolant System 53A Charging & Letdown 53B' Boric Acid Makeup 54-9 Primary Sampling
- 55A Gaseous Waste Management A-37
.i
gatem #
System Description
-SSB Liquid & Laundry Waste Mana3ement 55E Liquid & Laundry Waste Management
~58 Safety Injection 60A Safety Injection 60B Safety Injection 60C Safety Injection 59 Containment Spray 61 Fuel Handling & Storage 66 Plant' Protection System 63 Plant Protection System
-65A-1 Excore Nuclear Inst.
65A-2 Excore Nuclear Inst.
71B Condensate Makeup 73 Emergency Feedwater 75 Secondary Sampling 76 Steam Generators & MSIV 91 Seismic Supports
.19 Whip Restraints 17 System Supports (Hangers)
Seismic Structures
\\
A-38
Issue #19 - Water in Basemat Instruments Subgroup C - Water in basemate instruments does have a potential effect on:
' System #'
System Description
Evaluation 08A 208/120 v Elec. Distribution The present analysis for Safety moderate energy pipe rupture flooding per the
~*~
'10-Communications FSAR envelopes the concern for water seepage since 13A Heat Trace Safety this flow rate would be minimal. Accordingly, 17 Seismic Monitoring this issue does not serve as a restraint to safe 18-1 Radiation Monitoring operation of these System systems, and has been resolved and closed out 18-2 Radiation Monitoring by LP&L.
System
.18-3 Radiation Monitoring System 18-4 Radiation Monitoring System 18-5 Radiation Monitoring System 20' Security System 36-1 Component Cooling Water 36-2 Component.Cocling Water 36-3 Aux' Component Cooling Water-43A RCB Containment Cooling 46D-RAB HVAC 46E-RAB Chilled Water l
.53A Charging & Letdown 53B.
Boric Acid Makeup A-39
l
. System #
System Description
'SSA Gaseous Waste Management SSB Liquid & Laundry Waste Management SSE Liquid & Laundry Waste Management
+
i l
58-
. Safety Injection 60A Safety Injection.
60B Safety Injection 60C Safety Injection
-59 Containment Spray 71B.
Condensate Makeup 73 Emergency Feedwater Seismic Structures d
A-40
,;+.
Issue #20 - Construction Materials Testing (CMT) Personnel Qualifications Records Subgroup C - Construction Material Testing (CMT) personnel did do work on the system and a safety evaluation was performed to assure LP&L management that Waterford SES #3 can be safely operated without compromising the health and safety of the.public.
Issue #20 does have a potential effect on:
' System #-
System Description
Evaluffion Seismic Structures An Engineering Evaluation of CMT for backfill soils indicates no defective work of safety significance was accepted as a result of testing personnel actions.
All documentation of.
qualifications for GEO personnel involved with the concrete testing are being reviewed for completeness and verification.
E 1
n L
A-41
.l '
~~
- Issue #21 -'LP&L QA Construction System Status and Transfer Reviews Subgroup C - Open walkdown comments did have a potential impact on the system even though startup and system engineering evaluated the walkdown concerns and determined that there is no adverse impact on system / testing or operability.
Issue #21 does have a potential effect on:
System #
System Description
Evaluation 71 Condensate Makeup All open walkdown comments have been resolved / closed.
91 Seismic Supports All significant construction QA findings have been identified and properly dispositioned.
Accordingly, this review does not serve as a constraint to safe operation of these systems, and has been resolved and closed out by LP&L.
4 '
A-42
l 4
/
'i Issue #22'- Welder Qualifications (Mercury) and Filler Materials Control (Site Wide)
Subgroup C - The LP&L review of qualifications status documentation for all Mercury welders has been cotcpleted and the program does have a potential impact on the system. The weldment filler material controls did apparently
' deviate from code requirements.
. Issue #22 does have a potential effect on all systems in Table A-6.
The Evaluation contains a clarification of the review finding on welder qualifications, and there are no potential unreviewed safety questions pertinent to this issue.
"Rebaking" of low hydrogen electrodes was not practiced on the site and engineering justification demonstrates that while there were limited deviations from code specifications however this did not cause degradation of quality of weldment filler material.
9 6
e e
o E
A-43
t Issue #23 - QA Program Breakdown Between EBASCO And Mercury The concern is not directly related to the systems under review and is
. considered to be programmatic in nature.
There are no Subgroup C systems.
e' O
A-43a
. _ -.,.. _. --.. _ _.. _,, _ - -.. ~ _.. _ _...... _.... _ _ _ _ _. _.. _ _ _ _. _. -__ _ ____._.. _.._
n.
e TABLE A-6 ISSUES No. 1 Es. 2 No. 3 No. 4 No. 5 No. 6 No. 7
%. 8 Inspection M8==8=r 32 Instrumsa-Lower Tier Vendor Docu-Disposition-Backfill Visual Exam-Personnel h t=rian Ex-Corrective mentation - ing of Non-Soil ination of Issues 1.ime h pension % Actions are Conditional conformance Densities Shop Welds Esperation mot being Releases and Discrep-During SYSTEM Opgraded to ancy Baports Rydrostatic (A)(B)(C)
NCBs.
Teating t
02A - 125v DC Safety A B A
C C
C C
A 03
- Switching Station A B C
C C
C A
04
- Startup Transformers A B A*
C C
C C
A 06A - 4.16ky Elec.
A B A
C C
C C
A Distribution Safety 07A - 480v Elec.
A B A
C C
C C
A Distribution Safety 08A - 208/120e Elec.
A B C
C C
C A
4 Distribution Safety 09A - Invertsrs &
A B A
C C
C C
A Distribution Safety J
]
10
- Communications A B A
A C
C C
C A
1 13A Heat Trace Safety A B A
A C
C C
C A
, 16
- Environmental A B A
C C
C C-A Monitoring i -
17
- Seismic.' Monitoring A B
.i A
C C
C ~
C A
i 4
i A-44
TABLE A-6 ISSUES s-No. 1 No. 2 No.3-No. 4 No. 5 No. 6 No. 7 No. 8 Inspection Missing N1 Instrumen-Lower Tier. Vendor Docu-Disposition-Backfill Visual Exam-Personnel Instrument tation Ex-Corrective mentation - ing of Non. Soil ination of Issues Line Docu-pension Loop Actions are Conditional conformance Densities Shop Walde mentation Separation not being Releases and Discrep-During Upgraded to ency Reports Hydrostatic SYSTEM (A)(B)(C)
NCRs Testing a
18 Radiation Monitoring B B A
A C
C C
C C
System 18-2 3 B A
A C
C C
C C
18-3 C* C*
A A
C.
C C
C C
1 18-4 A B A
A C
C C
C C
18-5 A B A
A C
C C
C C
l 20
- Security System A B A
A C
C C
C A
21
- Fire Detection A B A
B C
C C
C A
22
- Fire Protection --
C* C#
A B
C C
C C
B 36 Component cooling Water C* C*
C*
5 C
C C
C C
36-2 C* C*
C*
B C
C C
C C
36 Aux Component Cooling C* C*
C*
5 C
C C
C C
l Water 39
- Emergency Diesel Cenerator C# C*
C*
B C
C C
C B
l 40 Crane & Holst FHB A B A
A C
C C
C A
f 43A - RCB Containment Cooling C* B A
B C
C C
C A
9
]
The issue does have a potential effect on the system; the defined a
activity necessary to close out the concern has not Leen completed to date.
A-45 t
a
4 TEB12 &-6 m
No. 1 No. 2 No. 3 En. 4 No. 5 No. 6 No. 7
.No. 8 Inspection Missing N1 Instrismsm-Jammar Tier Vendor Docu-Disposition-Backfill Visual Exam-Personnel Instrument tation En-. hive mentation - ing of Non-Soil ination of Issues Line Docu-pension W a ** - are Conditional confa.ance Densities Shop Welds mentation Separart-ust king-Releases and 1,iscrep-During W to ancy Reports Rydrostatic SYSTEMS (A)(B)(C)
Mar Testing 438 - Shield Bids. Ventilation C* C*
C*
B C-C C
C A
43E - Containment Vacuum Relief C* C*
A B
C' C
C C
A 468 - Control Roon HVAC C* C*
A B
C C
C C
A B
C C
C C
A 46E - RAB Chilled Water C* C*
A B
C C
C C
B 46K - Fire Despers A B A
A C
C C
C A
48
- 1.RT Containment Vessel A
C*
A B
C C
C C
A 49
- Process Analog Control A B A
B C
C C
C A
505 - Misc. Panels A B A
B C
C C
C A
1 i
52A - Reactor Coolant System C* C*
A B
C C
C C
C l
J
$2B C* C*
A B
C C
C C
C l
1 52C C* C*
A B
C C
C C
C I
The issue does have a potential ef fect on the systems the defiand activity necessary to close out the concern has not been coupleted to date.
]
A-46 r
p i
I l
i i
i
TABLE A-6 ISSUES No. 1 No. 2 Wo. 3 No. 4
'No. 5 No. 6 No. 7 No. 8 Inspec tion Missing N1 Instruen-Lower Tier Vendor Docu-Disposition-Backfill Visual Exam-Personut Instrument tation Ex-Corrective. mentation - ing of Non-Soil ination of Issues Line Docu-pension Loop Actions are Conditional conformance Densities Shop Welds mentation separation not being Releases and Discrep-During Upgraded to ancy Reports Bydrostatic STSTDis (A)(B)(C)
NCRs Teoting 534 - Charging & 14tdown C* C*
A B
C C
C C
~C 535 - Boric Acid Makeup C* C*
A B
C C
C C
C 54 Primary Sampling B
C*
A B
C C
C C
C 55A - Ceseous Weste Management C* C*
A B
C C
C C
C O
553 - Liquid & Laundry Waste C* C#
A B
C C
C C
C Management 55E B
C*
A B
C C
C C
C 58
- Safety Injection C* C*
A B
C C
C C
C 60A C* C*
A B
C C
C C
C
/
605 C* C*
A B
C C
C C
C 60C C* C*
/
A B
C C
C C
C i
59
- Containment Spray C# C#
A B
C C
C C
C C1
- Fuel Handling & Storage A
C*
A B
C C
C C
B The issue does have a potential effect on the system; the defined activity necessary to close out the concern has not been completed to date.
A-47
\\:
i
TABLE A-6 IeEWa No. 1 No. 2 No. 3 No. 4 Rs. 5 No. 6 No. 7 No. 8 Inspection Missing N1 Instrumen-Lower Tier Vendor Docu-Disposition-Backfill Visual Exam-Personnel Instrument tation Ex-Corrective ammtatamm - ing of Non-Soil ination of Issues-Line Docu-pension Loop Actions are w ei - E comformance Densities Shop Ifelde j
aantation Separation mot being n=1-and Discrep-During Upgraded to ancy Reports Hydrostatic j
STSTEMS (A)(B)(c)
NCBs Testing i
i i
j 66
- Plant Protection System C* B Ce C*
C C
C C
A 63 C* B C*
C*
C C
C C
A i
65A Excore Nuclear Inst.
A C*
A A
C C'
C C
A 65A-2 A B A
A.
C C
C C
A 715 - Condensate Makeup C* C*
A B
C C
C C
C 73
- Emergency Feedwater C* C*
C*
5 C
C C
C C
75
- Secondary Sampling 7 5 A
B C
C C
C B
76
- Steam Cenerators & MSIV C* C*
C*
B C
C C
C C
88
- Turbine & Turbine Controls B
C*
A B
C C
C C
A 91
- Seismic Supports A B A
A C
C C
C A
19 Idhip Restraints A
C*
A A
C C
C C
A 19 System Supports (Hangers)
A C*
A A
C C
C C
A
- - Seismic Structures A B A
A C
C C
C A
The issue does have a potential effect on the system; the defiwd activity necessary to close out the concern has nnt been completed to date.
A-48 i
t l
k I
i
.l TABLE A-6 ISSUES No. 9 No. 10 No. 11 Iso. 12 Ilo. 13 Iso. 14 No. 15 No. 16 Walder Cer-Inspector Cadwelding Mein Steam-Missing NCRs J.A. Jones Welding of Surveys and tifier. tion qualifica-line Framing Speed-
"D" Level Exit Interviews tions (J.A.
Restraints Letters Material
. of QA Personnel 1
Jones &
and EIRs Inside Sm Fegles)
Containment i
02A - 123v OB W A
C A
C*
C A
C l
03
- SeiteMme 3rarian A
C A
C*
C A
C 04
- Startup Transformers A
C A
C*
C A
C 06A - 4.16kw Elec.
A C
A C*
C A
C Distri he h Safety 1
07A - 480v Elec.
A C
A C*
C A
C Distribution Safety 1
08A - 200/120e Elec.
A C
A C*
C C
C N
Distri h h Safety j
09A - Inverters &
A C
A C*
C A
C Distribution Safety s.
s I
10 r - icattees A
,C A
C*
C A
C 13A Ileat Trace Safety A
C A
C*
C A
C 16
- hvirosmaatal A
C A
C*
C A
C Mneitoring 17
- Seismic Manitoring A
C A
C*
C C
C j
a Ibe issue does have a potential ef fect on the systems the defined j
activity necessary to clope out the concern has not been completed to de*e.
f, A-49 i
i i
i 1
i
e TARLE A-6 i
ISSUES q
No 9 No. 10 llo. 11 No. 12 Ho. 13 Ilo. 14 Es. 15 No. 16 '
Walder Cer-Inspector Cadwelding Mein Steam-Missing DCRs J.A.* Jones. Dolding of Surveys and tification Qualifica-line Framing Speed N " lawel Exit Interviews tions (J.A.
Restraints Intters Esserial of QA Personnel Jonee &
and EIts Inmid=
SYSTEM Fegles)
W 18 Radiation Monitoring A
C A
C*
C C
C System 4
18-2 A
C A
C*
C C
C 18-3 A
C A
C*
C C
C 18-4 A
C A
C*
C C
C 2
18-5 A
C A
C#
C C
C 20
- Security System A
C A
C*
C C
21
- Fire Detection A
C A
C*
C C
C
(
22
- Fire Protection A
C A
C*
C C
C 36 Component Cooling Water A
C A
C*
C C
C 36-2 A
C A
C#
C C
C 36 Aax Component Cooling A
C A
C*
C A
C Water 39
- Bnergency Diesel Generator A
C A
C*
C A
C 40 Crane & Hoist FHB A
C A
C*
C C
C j
43A - RCB Containment Cooling A
C A
C*
C C
C i
The issue does have a potential effect on the system; the defined activity necessary to close out the concern has not been completed to date.
A-50 1
1
TABLE A-6 I
ISSUES No. 9 Wo. 10 No. 11 Wo. 12 Wo. 13 Wo. 14 No. 15 No. 16 Esteur h Inspector Caduelding Main Steam-Missing NCRa J.A. Jones Welding of Surveys and taiseme4 - qualifica-line Framing Speed "D" Level Exit Interviews tions (J.A.
Restraints tatters Material of QA Fersonnel Jones &
and EIRs Inside Fegles)
Containment-ST5TEMS 438 - Shield 51ds. Ventilation A
C A
C*
C A
C' 43E - Containment Vacuum Relief A
C A
C*
C C
C 465 - Control Room HVAC A
C A
C*
C A
C l
46D - RAB BVAC A
C A
C*
C A
C t
46E - RAB Chilled Water i
A-C A
C*
C A
C 46K - Fire Dampers A
C A
C*
C A
C 48
- LRT Containment vessel A
C A
C*
C C
C 49
- Process Analog Control A
C A
C*.
C A
C 503 - Misc. Pa2els A
C A
C*
C A
C 52A - Reactor Coolant System A
C A
C*
C C
C 528 A
C A
C*
C C
C 52C A
C A
C*
C C
C The issue does have a potential effect on the system; the defined activity necessary to close out the concern has not been completed to date.
A-51
TABLE A-6 ISSUES No. 9 No. 10 No. 11 No. 12 No. 13 No. 14 No. 15 Es. 14 Walder Cer-Inspector Cadwelding Main Steam-Missing NCRs J.A. Jones Walding af Surveys and tification qualifica-line Framing Speed "D" Insel Brit Intervians tions (J.A.
Restrainto Letters ~
Material of QA Perseemel
. Jones &
and EIRa Inside SYSTEMS Fegles)
Cas mi -
e 53A - Charging & l.etdown A
C A
C*
C C
C 538 - Boric Acid Makeup A
C A
CA C
A C
54 Frimary Sampling A
C A
C*
C C
C
$5A - Caseous Weste Management A
C A
C*
C A
C a
555 - Liquid & Laundry Weste A
C A
C*
C C
Management 55E A
C A
C*
C A
C 58
- Safety Injection A
C A
C*
C C
C 60A A
C A
C*
C C
C a
60B A
C A
C*
C C
C I
j 60C A
C A
C*
C C
C 59
C A
C*
C C
C j
61
- Fuel Handling & Storage A
C A
C*
C C
C The issue does have a potential effect on the system; the defined i
activity necessary to close out the concern has not been completed to date.
l A-52 1
i e
[
?
. s i
TABLE A-6 ISSUES No. 9 Es. 10 Es. 11 No. 12 No. 13 No. 14 No. 15 No. 16 Us1 der Cer-Inspector Caduelding Main Steam-Missing NCRs J.A. Jones Welding of Surveys and tification f>=1e'see-line Framing Speed "D" Level Exit Interviews tiens (J.A.
Restraints Letters Material of QA Personnel James &
and EIts Inside SYSTEMS Pagiss)
Containment 66
- Plant Frotection System A
C A
C*
C A
C 63 A
C A
C*
C A
C 65A Excore Nuclear Inst.
A C
A C*
C C
C 65A-2 A
C A
C*
C C
C j
71B - Condensate Makeup A
C A
C*
C C
C 73
- Bsergency Feedvater A
C A
C*
C A
C 75
- Secondary Sampling A
C A
C*
C A
C 76
- Steam Generators & MSIY A
C C
C*
C C
C 88
- Turbine & Turbine Controls A
C-A C*
C A
C 91
- Seismic Supports A
C C
C*
C C
C 19 Whip Restraints A
C C
C*
C C
C 19 System Supports (Hangers)
A C ~
C Ce C
C C
- Seismic Structures C*
C C
C*
C C
C
- The issue does have a potential effect ce the systam; the defined activity necessary to close out the concern has not been coupleted to date.
A-53 l
e:
TABLE A-6 ISSUES No,' 17 No. 18 No. 19 No. 20 -
- No. 21 No. 22 No. 23 QC Verifi-Documen-Water in Construction LF&L QA Walder QA Program cation of tation of Basemat Meterials Construc-Qualifica-Breakdown Expansion Walkdowns on Instruments Testing tion System tions.
Between Ebesco Anchor Quar-Non-Safety (CMT)
Status and (Mercury) and Mercury acteristics Related Personnel Transfer and Filler Equipment qualifica-Reviews
- aterial tion Records STStat Control (Site Wide) 02A - 125v DC Safety A
C A
A A
C A
03
- Switching Station A
A A
A A
C A
04
- Startup Transformers A
A A
A A
C A
06A - 4.16kv Elec..
A C
A A
A C
A Distribution Safety 07A - 480v Elec.
A C
A A
A C
A Distribution Safety OSA - 208/120v Elec.
A C
C A
A C
A Distribution Safety 09A - Inverters & '~
A C
A A
A C
A Distribution Safety 10
- Communications A
C C
A A
C A
13A Heat Trace Safety A
C C
A A
C A
16
- Environmental A
C A
A A
C A
Monitoring 17
- Seismic Monitoring A
C C
A A
C A
~ _, ""
A.-54
e TABLE A-6 ISSUES No. 17 No. 18 Bo. 19 Bo.- 20 No. 21 No. 22 No. 23 QC Verifi-Documen-Enter Su Cemetracties LP&L QA Welder QA Program cation of tation of Beaumst hterials Construc-Qualifica-Breakdown Expansion Walkdowns on h 1setius tion System tions Between Ebasco Anchor Char-Non-Safety (EEEE)
Status and (Mercury) and Marcury acteristics Related Perasemel Transfer and Filler Equipment EpaleNm Reviews Material tien Becords Control SYSTEM (Site Wide) 18 Badiation Monitoring C*
C C
3 C
A Systee 18-2 C*
C C
A B
C A
i 18-3 C*
C C
A B
C A
18-4 C*
q C
A B
C A
18-5 C*
C C
A B
C A
20
- Security System A-C C
A A
C A
21
- Fire Detection A
C A
A A
C A
22
- Fire Protection A
C A
A A
C A
36 Component Cooling Water C*
C C
A B
C A
36-2 C*
C C
A B
C A
36 Aux Component Cooling C*
C C
A B
C A
Water 39
- Emergency Diesel Cenerator C*
C A
A A
C A
40 Crane & Hoist FHB A
C A
A A
C A
43A - RCB Containment Cooling C*
C C
A A
C A
The issue does have a potential ef fect on the system; the defined activity netessary to close out the concern has not been completed to date.
A-55
~
s.
TABLE A-6 ISSUES No. 17 No. 18 No. 19 No. 20 No. 21 No. 22 No. 23-QC Verifi-Documen-Water in Construction LP&L QA Welder QA Program cation of tation of Basemat Materials Construc-Qualifica-Breakdown Expansion Usikdown on Instruments Testing tion System tions Between Ebasco Anchor Char-Non-Safety (CMT)
Status and (Mercury) and Mercury acteristics Related Personnel Transfer and Filler Equipment qualifica-Rev.iews Material j
tion Records Control SYSTEMS (Site Wide) 4B - Shield 81d3. Ventilation C#
C A
A
'B C
A 4E - Csetainment Vacuum Relief C*
C A
A A
C A
465 - Control Room HVAC C*
C A
A B** -
C A
%4 - RAB RVAC C*
C C
A A
C A
46E - RAB Chilled Water C*
C C
A B
C A
46K - Fire Dampers A
C A
A A,
C A
48
- LRT Containment Vessel A
C A
A A
C A
49
- Process Analog Control A
C A
A A
C A
508 - Misc. Fanels C*
C A
A A
C A
52A - Reactor Coolant System C*
C A
A A
C A
525' C*
C A
A A
C A
$2C C*
C A
A A
C A
- The issue does have a potential effect on the system; the defined activity necessary to close out the concern has not been completed to date.
- - This system was incorrectly identified as 4389 in this issue.
A-56 l
i
TABLE A-6 ISSUES-No. 17 No. 18 No. 19 Se. 20 No. 21 No. 22 No. 23 QC Verifi-Documen-Water in Cemetruction LF&L QA Welder QA Program cation of tation of Basemat Heterials Construc-Qualifica-Breakdown Expansion Walkdowns on Instruments Testing tion System tions Between Ebasco Anchor Char-Non-Safety (OEE)
Status and (Mercury) and Mercury acteristics Related Personeel Transfer and Filler Equipment M fica-Reviews Material time Records Control SYSTDtS (Site Wide) 3A - Charging & I.atdown C*
C C
A A
C A
38 - Boric Acid Makeup C*
C C
A A
C A
,4-9
- Primary Sampling A
C A
A A
C A
5A - Caseous Waste Management C*
C C
A B
C A
5B - Liquid & Laundry Waste A
C C
A A
C A
Management 55E A
C C
A A
C A
8,4 S;f ty Injection C*
.C C
B C
A 60A C*
C C
A B
C 605 C*
C C
A B
C A
60C C*
C C
A B
C A
D
- Containment Spray C*
C C
3 C
A
,1
- Fuel Handling & Storage C*
C A
A A
C A
e Tha 1: sue does have a potential ef fect on the system; the defined activity necessary to close out the concern has not been completed to absta.
A-57 e
~.
TABLE A-6 ISSUES
.s
.I i
No. 17 No. 18 No. 19 No. 20 No. 21 No. 22 No. 23,
QC Verifi-Documen -
Water in Construction LF&L QA Walder QA Program t
cation of tation of Basemat Materials.
Construc-Qualifica-Breakdown Expansion Walkdowns on Instruments Testing tion System tions Between Ebasco Anchor Char-Non-Safety (CNF)
Status and (Mercury) and Mercury
,j acteristics Related Personnel Transfer and Filler Equipment Qualifica-Reviews Material tion Records Control
{
SYSTEMS (Site Wide) 1
)
I CS
- Plant Protection System C*
C A
A A
C A
63 C*
C A
A A
C A
l 05A Excore Nuclear Inst.
A C
A A
A C
A 65A-2 A
C A
A A-C A
N1B - Condensate Makeup C*
C C
A C
C A
D1
- Energency Feedwater C*
C C
A A
C A
95
- Secondary Sampling A
C A
A-A C
A E6
- Stian Cenerators & MSIY C*
C A
A A
C A
[D
- Turbine & Turbine Controls A
A A
A A
C A
C1
- Seismic Supporte A
C A
A C
C A
D9 idhip Restraints A
C A
A A
C A
09 Sy: tee Supports (sansers)
Ca C
4 A
A C
A
- - Seismic Structures C*
C C
C*
A C
A
- The issue does have a potential ef fect on the system; the defined activity necessary to close out the concern has not been completed to date.
A-38
)
J 6
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EXAMPLES OF SAFETY REVIEWS PERFORMED UNDER LICENSING PROGRAM PLAN I
A-59 l
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ITEM NO.'
/.9
- /
S/U NO.
ISSUE NO.
1.
EFFECT/ RELATIONSHIP No_effect.
No work was performed on this system by Mercury.
2.
SAFETY REVIEW ( ATTACH CHECKLIST)
Review complete.
No. work was performed on this system by Mercury.
~3.
STATUS A.
Done?
yes
,.w-B.
Verified (if req'd).
Not irequired 4.
OUTSTANDING ACTION
{
None b a ///p TTACHME'NTS Issue Person 1). Safety Review Checklist System Pers hu (One per item no.)
K.C. or RPB Nlb Revision O Sept. 5, 1984 x
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Nuclear Safety Review Checklist (Safety Evaluation) 10CFR 50.59 1.
EVALUATION APPLICABLE TO:
Systen
/8-Y ARM /RMC/Pfi f kkk %.n%-i.$ Mk t
Issue No.
1 - Mercury E
2.
SAFETY EVALUATION:
A' written basis / justification for the answers in Section 2 must be provided i
i 2.1) True X
False The probability of an accident Previously evaluatedintheF5ARwillnotbeincreased.
Justification:
Mercury performed no work on this system.
j 2.2) True X False ~
The consequences of an accident previously
+
evaluated in the FSAR will not be increased.
Justification:. Mercury performed no work on this system.
I 2.3) 'True X
False The possibility of an accident which is I
different than any already evaluated in,the,
.FSAR Will not be created.
Justification:
Mercury performed no work on this system.
1.
l a
U l
~
2.4) True X
False The probability of malfunction of equipment
{
important to safety previously evaluated in the FSAR will not be increased.
[
Justification:
Mercury performed no work on this system.
l-L r
l UNT-TEM-00 Revision 0 5.1 (1 of 2)
i systan
/6-V M M lk M C / W I.'
Icau2-Na.
1 - Marcury
/
Nuclear;SafetyiReview Checklist q
(Safety : Evaluation)
't l
1
)
2.5)' True X
False The consequences of a malfunction of equipment i
important to safety previously evaluated in the FSAR will not be increased.
Justification: Mercury performed no work on this system.
2.6) True X-False The possibility of a malfunction of equipment important to safety different than any already evaluated in the FSAR will not be created.
Justification: Mercury performed no work on this system.
2.7).True y
False
% margia of safety as tiefined in the basis to
'g -
t any Tachaical Specification will not be reduced.
Justification: Mercury performed no work'on this system.
1-s i
If the answer for 'any of the questions for Section 2 is " FALSE", an unreviewed '
-ssf-ty question may be involved.
- 3) REMARKS:
(ATTACH ADDITIONAL PAGES IF NECESSARY)
- 4) PREPARED BY-Mdmn-t DATE 9/w/M
.5) REVIEWED BY [ M d Y DATE M,/fu
- 6) PORC REVIEW M DATE Y-/.3--2 '6_
O mal'A0000 h"b",0*lAL DATE ok\\
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SYSTEM / ISSUE SAFETY RESOLUTION WORKSHEET kdic.k.
Ad 6m/ % 0.~
rem /GMl' ITEM NO. 19-l.19-2 eru 1 - Mercury S/U NO.
ISSUE NO.
1.
EFFECT/ RELATIONSHIP No effect.
No safety related work was performed by Mercury on this system.
2.
SAFETY REhIEW ( ATTACH "HECKLIST)
Review complete.
No safety related work was performed by Mercury,on this system.
3 STATUS ~
A.
Done ?
Yes, ',
B.
Verified (if req'd).
Not required 4.
CUTSTANDING ACTION None TTACHMENTS Issue Person
/4//f/
- 1) Safety Review Checklist System Perso [ N e w (One per item no.)
K.C. or RPB-M l
Revision 0, Sept. 5, 1984
s-Nuclear Safety Review Checklist (Safety Evaluation) 10CFR 50.59
- 1. ' EVALUATION APPLICABLE TO:
System /E-/ /b2 AA.<f/R.*K/P/M sh MW UO
% h O%
Issde No. 1 - Mercury 2.
SAFETY EVALUATION:
A written basis / justification for the answers in Section 2 must be provided 2'.1)
True x False The probability of an accident previously evaluatedintheFbARwillnotbeincreased.
Justification: No safety r.1sted work van nerformed by Mercurv on thin arne 2.2) True x False The consequences of an accident previously evaluated in the FSAR will not be increased.
Justificat. ion: wo..r.cy v.1.,.A work una nerformed by Mercurv e ch 4.
2.3) True r
False The possibility of an a'cident which is
~
c different than any already evaluated in$the.
FSAR Will not be created.
Justification:
wo nor,ev r i.e a work una nerformed hv Mercurv on thin system.
2.4) True r__ False The probability of malfunction of equipment important to safety previously evaluated in the FSAR will not be increased.
Justification:
wo enreev r.1ne a work vnn nerformed hv Mercurv nn chim syne.m.
UNT-TEM-00 Revision 0
!5.1 (1 of.2)
- 5yiten ']s.l_ 'jf.2 AEMlkis1C/PArt asu;2 No. l' - Marcury
' Nuclear Safety Review Checklist t
(Safety Evaluation) 10CFR 50.59 2.5) True X
False The consequences of a malfunction of equipment important to safety previously evaluated in the t-FSAR will not be increased Justification;~ No safety related work was performed by Mercury
_ n this system.
o 2.6) True I-False The possibility of a malfunction of equipment important to safety different than any already evaluated in the FSAR will not be created.
Justification: No safety related work was performed by Mercury on this system.
2.7) True r False The margin of safety, as defined in the basis to g..
any Technical Specification will not be reduced.
Justification: un
. r.cy v.1.c.a unev o..
n rn-a wy w,...,.
on thic nyer,m_
.If the answer for any of the questions for Section 2 is " FALSE", an unrdriewed safety question may be involved.
3') REMARKS:
(ATTACH ADDITIONAL PAGES IF NECESSARY) l
- 4) PREPARED BT
/ %2m DATE 9//o/p</
- 5) REVIEkT.D BY MI[b DATE M 4 n g
- 6) PORC REVIEW MhNhm DATE Ti 3-J/
/
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k' SYSTEM / ISSUE SAFETY RESOLUTION WORKSHEET kMd4 b. A..3 %Aq_
ITEM NO. /S-3 AR!!!RMC/PQ+1 1 - pggcyp y S/U NO.
ISSUE NO.
O
-1.
EFFECT/ RELATIONSHIP The installation of safety related instrumentation on this system was incpected by potentially unqualified inspectors.
2.
SAFETI REVIEW ( ATTACH CHECKLIST)
Inspection of Mercury safety related work on this system by potentially unqualified inspectors indicates that an unreviewed safety question may be involved.
3 STATUb A.
Done?
No B.
Verified (if req'd)
_. ~
s 4.
OUTSTANDING ACTION The quality of safety related instrumentation associated with this system shall be verified. Verification is being accomplished by reinspection of NI instrument loops.
Satisfactory completion of th'is iample of Mercury installations will be the basis for acceptance of the remaining installations.
ATTACHMENTS Issue Person Af///f
- 1) Safety Review Checklist System Perso [ D 6 +,e _
(One per item no.)
K.C. or RPB M/N Revision 0, Sept. 5, 1984
\\
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i SYSTEM / ISSUE SAFETY RESOLUTION '40RKbHEET
\\
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1 - Tompkins-Beckwith ITEM NO.
Pc'm
~S/U NO. /F-3 ISSUE NO.
k
- 1.
E FFECT/ REL ATIONSHIP Work performed on this system was inspected by potentially unqualified Tompkins-Beckwith inspectors.
2'.
SAFETY REVII'd ( ATTACH CHECKLIST)
Inspection of safety related work by potentially unqualified Tompkins-Beckwith inspectors indicates that an unreviewed safety question may be involved.
6' 3.
STATUS A.
Done?
No 3.
Verified (if req'd) 4 OUTSTANDING ACTION a)- Verify the qualifications of the initial Tompkins-Beckwith inspectors b) Where the. initial qualifications in (a) above are inadequate, verify the qualifications of the inspectors performing any over inspection.
c) Where A and B are not met, reinspect or justify on a case-by-case basis.
' $[3, / f ATT ACHME lTS Issue Person
/
- 1) Safety Review Checklist System Person A lo 4u ce
( O ne per item no.)
K.C. cr RPS MLO Revision C, Sept. 5, 19eu l
.o.
i SYSTEM / ISSUE SAFETY RESOLUTION WOR EMEET
&cB cargiNMear ITEM NO.
CC S
- #6 1 - Tompkins - Beckwith S/ U NO. N#
ISSUE No.
1.
EFFECT'/ REL ATIONSHIP No.effect. Tompkins-Beckwith performed no safety related work on this system.
2.
SAFETI' REVIEW ( ATTACH CHECKLIST)
Review complete. Ne safety related work was performed by T94==-hchrith on this system.
~
3.
STATUS A.
Done?
Yes B.
Verified (if req'd)
Not required 4.
OUTSTANDING ACTION
~
None ATTACHMENTS Issue Person b[/4#4
- 1) _ Safety Review Checklist System Person b ' b "
(One per item no.)
K. C. or RPB bd Revision 0, Sept. 5, 1984 o
e
Nuclear Safety Review Checklist (Safety Evaluation)
~0CFR 50.59 1.-
EVALUATION APPLICABLE TO:
System cc 5' 93M Issue No. 1 - Tompkins-Beckwith 2.
SAFETY EVALUATION:
A written basis / justification.for the answers in Section 2 must be provided 2.1) True X False h probability of an accident previously evaluated in the FSAR will not be increased, Justification:
No safety related work was performed by Tompkins-Beckwith on this systaus.
er 2.2) True I False The consegnances of an accident previously evaluated in the FSAR will not be increased.
Justificat. ion: No safety reisted work was performed by Tompkins-Beckwith on this system.
2.3) True I False' The possibility of an accident which is different than any already evaluated in the FSAR Will not be created.
Justification: No safety related work was performed by Tomokins-Beckwith on this system.
2.4) True X False The probability of malfunction of equipment important to safety previously evaluated in the ISAR will not be increased.
Justification: No safety related work was performed by Tompkins-Beckwith on this system.
- Cten h S UA
! d' go, 1 - Tompkin2-Beckwith l
Nuclear Safety Review Checklist (Safety Evaluation) l 10CFR 50.59 2.5) True 1 Yalse The consequences of a malfunction of equipment important to safety previously evaluated in the FSAR will not be increased.
Justification. No safety related work was performed by Tompkins-Beckwith on this system.
I 2.6). Tree
'Falsa The possibility of a malfunction of equipment important to s.afety different than any already evaluated in the FSAR will not be created.
Justification: No safety related work was nerformed by Tomokin=-Beckwith on this system.
=
2.7) True X Yalse The margin of safety as defined in the basis to any Technical Specification will not be reduced.
Justification: No safety relatc.i work was perfo'rmed by Tompkins-Beckwith on this system.
I If the answer for any of the questions for Section 2 is " FALSE", an unrevic.'ad cciety question may be involved.
- 3) REMARIS:
(ATTACE ADDITIONAL PAGES II NECESSARY)
- 4) PREPARED If Wmm DATE 9/v/#V
- 5) REVIEWED IT O/M--
DATE EJ72,/f/t
- 6) PORC REVIEW
,,, I2 8M~..
DATE 9-/ V.W I
7 ) ". "-"? ~- " -""
- uau, lN ~ff A
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SYSTEM / ISSUE SAFETY RESOLUTION WORKSHEET ITEM -NO. LV A-ac?//26 v 7
S/U NO. k.OscT ISSUE NO.
i: t OSA i-1.
EFFECT/ REL ATIONSHIP The data for the 34 in-place decsity tests performed in the first 5.5 feet of class A fill in fill area #5 frca elevation -41.75 to -36.25 has been located and has been transmitted to the Ebasco QA records vault.
Therefore, this issue no longer has any effect on this system.
2.
SAFETY REVIEW (ATTACH CHECKLIST)
Done
~
3.
STATUS A.
Done? Completed 4...
B.
Verified (if req'd) 4.
OUTSTANDING ACTION None 4
@ k, & :!= 0 ATTACHMENTS Issue Person William c. Hubacek
- 1) Safety Review Checklist System Person
[l (One per item no.)
K.C. or RPB
(
8 v
Revision 0, Sept. 5, 1984 l
7 i
/
j a.
Nuclear Safety Review Checklist
{
(Safety Evaluation) i 10CFR 50.59 1.
EVALUATION APPLICABLE TO:
Systen LND ;,2cybpov Gu c.rmest hSr. (68 A\\
Issue No. 7 Backfill Soil Densities 2.
SAFETY EVALUATION:
A written basis / justification for the answers in Section 2 must be provided
~
2.1) True X
False The probability of an accident previously evaluatedintheF5ARwillnotteincreased.
Justification: Review and analysis of soil backfill in response to this concern indicates that Class A backfill soil densities are in accord-ance with* specification and FSAR requirements and will provide the required design structural capacity to the plant under seismic loading:
(FSAR SECT 2.5.4.5.3..Ebasco Spec. LOU 1564.482) 2.2) True X False The consequences of an accident previously
..,9,,,.
t evaluated in the FSAR will not be increased.
Justif,icat, ion: conseauene.. nrevionni, evalune.a u.v. un.a on korten11, which met specification and FASR r'equirements. The backfi1{
analysis indicates that the backfill meet specifications and FSAR requirements.
2.3) True x False The possibility of an accident which is different than any already evalusted'In'the FSAR Will not be created.
Justification: Backfill was placed in ac~ardance with requirements that were evaluated in the FSAR (Section 2.5.4.5.3); therefore, backfill will not contri'oute to accidents different from..
those evaluated in the FSA.R.
2.4), True X False The probability of malfunction of equipment important to, safety previously evaluated in the FSAR will not be increased.
Justification: Backfill meets spec and FSAR requirements: therefore. it does not contribute to the probability of equipment mal-function.
In:Tc,TrM.nn n.-4.4..
n R
sys u n s
c Nuclear Safety Review Checklist (Safety Evaluation) 10CFR 50.59 2.5) Truel False The consequences of a malfunction of equipment important to safety previously evaluated in the FSAR will not be increased.
Justification: Backfill meets spec and FSAR requirements; therefore, it does not contribute to the consecuences of equipment mal-function.
2.6) True I False The possibility of a. malfunction of equipment important to safety different than any already evaluated in the FSAR will not be created.
Justification: Backfill meets spec and FSAR requirements; therefore, it does not contribute to the nossibility of a malfunction different than evaluated.
2.7) True X False The margin of safety as defined in the basis tog..
any Technical Specification will not be reduced.
Justification: In as much as the backfill meets spec and FSAR reauirements, it does not contribute to reduction of the margin of safety as defined in tech spec.
If the answer for any of the questions for Section 2 is " FALSE", an unreviewed safety question may be involved.
- 3) REMARKS:
(ATTACH ADDITIONAL PAGES IF NECESSARY)
~
- 4) PREPARED bye 9!Ph h DA*.E 9/2/#9
- 5) REVIEWED BY t @ l f d
- DATE9/7/d
- 6) PORC REVIEW hh,_
DATE 9 /o.4P M
- 7)."E.'.CEMEhT dFRYv'AI, DATE
~
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....s.
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-ATTACHMENT B SAFETY REVIEWS FOR PLANT SYSTEMS REQUIRED BY TECHNICAL SPECIFICATIONS FOR CRITICALITY, LOW POWER TESTING AND FULL POWER OPERATION t
1
LICENSING PLAN FOR CRITICALITY, LOW POWER TESTING AND FULL POWER OPERATION The program discussed in Attachment C and applied to Fuel load and Precriticality Post Core-Load Hot Functional Testing in Attachment A is being applied to those systems required for Criticality, Low Power Testing and Full Power Operation. These systems are listed in Table B-1.
This l
process has been initiated and is expected to be completed shortly, although*the issuance of the initial license is not considered to be dependent upon completion.
Upon completion of this review, summaries will be prepared (as described in Attachment A, Table A-4) and full documentation will be filed in the Waterford 3 On-Site Licensing Unit offices for inspection and review by the NRC staff.
1 B-1
TABLE B-1 PLANT SYSTEMS REQUIRED FOR CRITICALITY AND LOW POWER TESTING TO FIVE PERCENT, AND FULL POWER OPERATION OPERABILITY ACRONYM SYS. NO-DESCRIPTION REQUIRED PMC 15 PLANT MONITORING COMPUTER MODE 1 ( 20%)
FP 22-3 FIRE PROTECTION - HALON MODE 1 ( 20%)
HRA 43H RCB HYDROGEN RECOMBINERS/
MODE 1-2 ANALYZER CEC 64 CONTROL ELEMENT ASSY.
MODE 1-2 CALCULATOR INI 65B INCORE NUCLEAR INSTRUMENTATION MODE 1 ( 20%)
MNI 65C MOVABLE.INCORE NUCLEAR INSTR.
MODE 1 ( 20%)
VLP 69 VIBRATION & LOOSE PARTS MODE 1-2 MONITOR 1
B-2
1 e
ATTACHMENT C LICENSING PROGRAM PLAN SAFETY REVIEW PROCESS DESCRIPTION me Y
O
/
/
I
LICENSING PROGRAM PLAN SAFETY REVIEW PROCESS DESCRIPTION The process used to perform the safety reviews is shown on the attached flow chart (Figure C-1).
Each of the principal activities are described in more detail-in Table C-1.
The transmittal letter also provides a general discussion of the process employed.
The two work instructions / procedures used to control the safety review process are included in this attachment.
They are UNT-TEM-006, which is an Administrative Procedure entitled,."FSAR-NUCLEAR SAFETY REVIEW", and ISEG WORK INSTRUCTION 84-1 entitled, " Licensing Program Plan Audit Plan".
An additional enclosure is a copy of the SYSTEM / ISSUE SAFETY RESOLUTION WORKSHEET, used to summarize the results and basis for safety review, the basis for not being able to complete a safety review and document any corrective actions required to be able to finalize a safety review.
I
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TABLE C-1 LICENSING PROGRAM PLAN FLOW CHART DETAILS Step No.
Description 1
. Develop standard format for matrix and input sheets.
-2 Establish order of priority for review of twenty-tnree
_(23) issues.
3 Prioritize startup systems by Technical Specification prerequisites for Fuel Load, Initial Criticality to 5% of full power, and full power operation.
4 Develop Work Instructions for performance of audit by the Independent Safety Engineering Group -(ISEG) to include a verification ~of consistency with validated responses and auditing of the process to the Work Instructions.
5 Develop Work Instruction for performance of Safety
. Reviews including development of checklist.
~6 Assign project representative for each issue.
7 Assign plant representative for each issue.
8 Establish Safety Review Teams.
9 Review and approve the process and Work Instructions.
10-Initiate Safety Reviews per approved Work Instructions.
11~
Review and approve Content and Consistency of Safety Reviews.
12 Verify consistency of Safety Reviews with validated responses per Work Instructions.
13 Coordinate PORC review of Safety Reviews and content of the process.
14 Plant Manager approval of the PORC review.
15 Perform audit of.overall process per Work Instructions.
.16 Coordinate SRC Subcommittee review for logic of process /
Safety Review.
17 Prepare report in final form and schedule review meeting with upper management.
18 Prepare transmittal letter, obtain signacures and transmit to NRC.
19 Extract information from NTOL report, including CAT items, SCDS and Inspection Report Open Items.
C-3 s
WATERFORD 3 SES PLANT OPERATING MANDAL Louisiama POWER & LIGHT YnBain PCM VOLingg 1
UNT-TEM-006 PON SECTI0lg 10 REVISION 1
ADMINISTRATIVE PROCEDURE FSAR - NUCLEAR SAFETY REVIEW PORC Meeting No.
fY-fd; Reviewed:."
^=
ore N ruan Approved:
/2 8 b
f 8/[k Piant Manag p uclear Approval'Date.
WIWsv Effective Date l
TEMPORARY CONDITION Thta procedure shall remain ineffect until Commerical Operation.
A h4=40trative Procedure UNT-TEM-006 FSAR-Nuclear Safety Review Revision 2 6
TABLE OF CONTENTS 1.0 PURPOSE 2.0 EDIRUGS 3.0 MFINITIONS 4.0 RESPONSIBILITIES 5.0 PROGDURE 5.1 Project / Plant Team 5.2 PORC Review 5.3 Priject/ Plant Management Approval 6.0 ATvamemTS 6.1 Reclear Safety Review Checklist.(3 pages)
..._____-.--__.s.,_p) M f*'N l
LIST OF ITFECTIVE PAGES Title Revistes 2 1-4 Revisies 2 J
t 0
l 1
Administr:tiva Pr:cedure UNT-TEM-006 FSAR-Nuclear Safety Review Revision 2 1.0 PURPOSE This procedure provides guidances for performing safety reviews for those systems required for fuel load and post fuel load testing.
l The safety review shall assure completion of those actions necessary to insure a system is constructed and functions according to the requirements of the FSAR in light of the 23 issues raised by the NRC.
2.0 REFERENCES
2.1 Waterford 3 FSAR 2.2 10 CFR 50.59 Safety Evaluation 2.3 UNT-1-004 Plant Operations Review Committee 3.0 IEFINITIONS 3.1 PORC Review - A review that is performed to ensure that a 10CFR50.59 Safety Evaluation is performed, when required; that an unreviewed safety question does not exist; that nuclear safety is not adversely affected; that the Technical Specifications are not violated; and that the admini'strative controls for procedures, and changes thereto, have been strictly adhered to as prescribed by this procedure.
3.2 Plant Manager - Nuclear Approval - Review of PORC's recommendations for approval of the safety review.
3.3 Safety Review, A review performed for all discrepancies / deviations to determine whether an unreviewed 10CFR50.59 Safety geustion needs to be addressed.
x t
t s
2'
.e Administr:tive Procedure IRfT-TEM-006 FSAR-Nuclear Safety Review Revision 2 3.4 10CFR50.59 Safety Evaluation - An evaluation of a system for discre-pancies/ deviations to determine whether the discrepancy / deviation involves an unreviewed safety question.
3.5 Unreviewed Safety Question - A discrepancy / deviation from design requirements as described in the FSAR shall be deemed to involve an unreviewed safety question if:
- 1) the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or 2) the possibility for an accident or malfunction of a different type than any previously evaluated in the safety analysis report may be created; or 3) the margin of safety as defiaed in the basis for say technical specification is reduced.
3.6 The Project Plant Team is a team consisting of a =4n4-of two people chosen by management based upon their knowledge of the 23 NRC issues and knowledge of technical specifications and plant systems to perform the safety reviews described in this procedure.
4.'0 RESPONSIBILITIES 4.1 Project / Plant Management is responsible for ensuring the development and implementation of this procedure.
Project / Plant Management also reviews and approves Safety Reviews for content and consistency.
4.2 Project / Plant teams are responsible for providing sys, tem safety reviews in accordance with the requirements of this procedure.
4.3 PORC is responsible for assuring completion of all actions necessary to insure a system is constructed and will functfon in accordance with FSAR requirements. PORC actions include a review of Plant and Project team safety reviews.
3
l
/
Administrative Procedure UNT-TEM-006
.-fFSAR-NoclearSafetyReview Revision 2
' 5.0,PROCEURE 5.1 Project / Plant teams shall perform safety review for their assigned systans using the Nuclear Safety Review Checklist, Attach==nt 6.1.
A written basis / justification for answers must be provided.
Answers checked False will require follow up eva nation and corrective actions.
5.2 PORC shall review the Project / Plant team safety reviews for content and adequacy.
PORC shall recommend approval of the safety review by Project / Plant Management.
5.3 Plant Management shall review and approve safety reviews.
5.4 The SRC shall review the actions of 5.2 and 5.3.
... _ < _ _ _ M fNU 5.5 (f. - '
. O f.fystems may be identified by the system acronyms or name and the system (s) startup (S/U) number.
6.0 ATTACEMDTS 6.1 Nuclear Safety Review Checklist (2 pages)
O
^^
-.....A._
Of A _ A _h.
LmL A
4
SYSTEM NO.
-cApi J
ISSUE-FO.
Nuclear Safety Review Checklist 4M.
je3-77 (Safety Evaluation) 10CFR 50.59 EVALUATION APPLICtJLE TO:
~
System Issue No.
2.
SAIITY EVALUATION:
A written basis / justification for the answers in Section 2 aust be provided 2.1) True False The probability of an accident previously evaluated in the FSAR will not be increased.
Justificatica:
2.2) True False The consequences of an accident previously evaluated in the FSAR will not be increased.
Justification:
2.3) True Falsa The possibility of an accident which is different than any already evaluated in the FSAR Will not be created.
Justifications l
l I
UNT-TEM-006, Revision 2 5.1 (1 of 3)
SYSTEM NO.
ISSUE NO.
Nuclear Safety Review Checklist (Safety Evaluation) 10CFR 50.59 l
2.4) True False The probability of malfunction of equipment important to safety previously evaluated in the FSAR will not be increased.
Justification:
2.5) True Felse The consequences of a es1 function of equipment tapon. sat to safety previously evaluated in the FSAR will not be increased.
Justification:
l i
2.6) True False The Possibility of a malfunction of equipment important to safety different than any already evaluated in the FSAR will not be created.
Justification:
I UNT-TEM-006, Revision 2 6.1 (2 of 3)
SYSTEM No.
1885 No.
Nuclear Safety Review Checklist (Safety Evaluation) 10CFR 50.59 2.7) True False The margin of safety as defined in the basis to any Technical Specification will not be reduced.
Justification:
If the answer for any of the questions for Section 2 is " FALSE", an unreviewed safety question may be involved.
- 3) REMARES:
(ATTACH ADDITIONAL PAGES IF NECESSARY)
- 4) PREPARED BY DATE
- 5) REVIEWED BY DATE
- 6) PORC REVIEW DATE
- ) _ : : e n :-- :A:.
DATE--
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iv3 5'I 18rt 75ti-006, Revteten 8 7.1 (3 of 3)
r SYSIIM/ISSE SAFETT RESOLUTION WORKSEET l
ITEM NO.
S/U No.
ISSE No.
1.
EffECT/REATIONSHIP l
2.
SAIITT REVIEW (ATTAG CHECELIST) 3.
STATUS
- 4. %f 5.
Vertised (ti reg'd) 4.
OUTSTANDING ACTION ATIACRENTS Issue Person
- 1) Safety Redew Checklist System Person (One per item no.)
K.C. or RPB t
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ISEG WORK INSTRUCTION 84-1
" Licensing Program Plan Audit Plan" x
9 !7![b Approved by
//
Date
/7M Approved by Date s
Revision 1
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ISEG WORK INSTRUCTION 84-1
" Licensing Program Plan Audit Plan" r
l 1.0 PURPOSE This Work Instruction provides direction for performing and documenting a special audit / review of the Licensing Program Plan (i.e., the System /
Issue Safety Resolution Worksheet, Safety Reviews, etc.).
2.0. INSTRUCTIONS 2.1 Consistency Verification (Attachment 3.1)
~
2.1.1 Verify the package is complete.
2,,1.2 Verify the package has received review by management.
2.1.3 Venfy that the technical content meaning 'is unchanged between the System / Issue Safety Resolution Worksheet and the NRC Submittal.
2.2 Audit Overall Process (Attachment 3.2)
~
2.2.1 Verify the package is complete.
2.2.2 Verify Nuclear Safety Review done in accordance with Work Instruction.
2.2.3 Verify the package has received review by management.
3.0 ATTACHMENTS
~' 3.1 Consistency verification Checklist 3.2 Audit Overall Process Checklist I
v Revision 1. 9/7/84
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ATTACHMENT 3.1 CONSISTENCY VERIFICATION CHECKLIS'T ISSUE NO.
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-1.
Verify the package is complete:
' System / Issue Safety Resolution Worksheets Yes-No, Comment
' Nuclear Safety Review Checklists Yes No. Comment 2.
Verify the package has received review by management.
Yes No, Comment 3.
Verify that the technical content meaning is unchanged during wording for NRC response:
'Does wording accurately reflect Effect/ Relationship Yes No, Comment
'Doeu wording accurately reflect Nuclear Safety Review Results Yes No, Comment
'Does wording accurately reflect Status Yes No, Comment
'Does wording accurately reflect Outstanding Actions Yes
-No, Comment.
O Reviewer Date ISEG Work Instruction 84-1 Revision 1, 9/7/8'+
-..2.
.- ~..~..
ATTACHMENT 3.2 AUDIT OVERALL PROCESS CHECKLIST ISSUE NO.
1.
. Verify the package is complete:
' System / Issue Safety Resoluation Worksheets
- Yes No, Coeument
- Nuclear-Safety Review Check 11s'ts Yes
-No, Cosament
'STS System Safety Review includes all applicable startup numbers.
Ye,s No, Comment 2.
Verify Nuclear Safety Review done in accordance with Work Instruction.
Yes No. Comment 7
- 3.. Verify PORC/PM-N Review Yes No,- Comment t
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Reviewer
-Date I
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~ISEG Work Instruction 84-1
_3_
Revision L 9/7]84
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ATTACHMENT D REGIONAL INSPECTION REPORT OPEN ITEMS REQUIRED FOR FUEL LOAD l
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REGIONAL INSPECTION REPO'LT OPEN ITEMS The latest. update of the Region IV NTOL has been reviewed and pertinent t
exhibits have been extracted for presentation in this attachment. These items are being pursued and closed through the normal processes with the Resident Inspector and are not directly related to the 23 issues.
~
Table'D-1 provides a listing of open. items, resulting from the construction and'startup test programs, which are remaining to be closed prior to fuel load.
As. indicated on the table, the preoperational testing is complete and the systems have been accepted by the plant staff.
Table D-2 provides'a listing of open IE Bulletins,. Circulars, Information Notices and NRR Generic Letters required for fuel load. No major fuel load prerequisite actions remain as indicated on Table D-2.
The listing of open Significant Construction Deficiencies (SCDs) preeented in Table'D-3 include those SCDs requiring LP&L action" prior to fuel load.
Open licensing commitments required to be. closed by fuel load from the Supplements to the Safety Evaluation Report and letters to the NRC are listed in' Table D-4.
Table D-5 provides a listing of NRC Inspection Report Open Items which are either-explicitly required to be closed by fuel load or have not been explicitly relegated to resolution at a later phase of the licensing process.
A status of TMI related open items is presented in Table D-6.
i D-1
.,. _., ~. _ _ -. -. _
Table D-1 SYSTEMS REQUIRED BY TECHNICAL SPECIFICATIONS TO BE OPERABLE BY FUEL LOAD HAVING OPEN FUEL LOAD ITEMS Preop.
Open LP&L Testing Items Staff System Description
% Complete Remaining Acceptance 02A 125 VDC-Safety.
100 3
A 10-20-82 09A Inverters & Dis-Safety 100 4
A 08-28-83 10 Communications 100 1
A 05-02-84 13A-1 Heat Tracing Safety 100 2
A 05-02-84 18-1 Radiation Monitoring-FHB 100 5
A 08-10-83 18-2 Radiation Monitoring-RCB/RAB 100 1
A 12-05-83 18-3 Radiation Monitoring-Process Effluent 100 3
A 05-11-84 20 Security 100 10 F 09-30-84 E22 Fire Protection 100 6
A 05-03-84 36-1 Component Cooling Water 100 16 A 12-12-83 36-2/3 Component Cooling water 100 5
A 08-03-84 39 Emergency Diesel Generator 100 3
A 05-18-84 43A RCB Containment Cooling 100 3
A 09-08-83 43E
-RCB Vacuum Relief 100 2
A 09-01-83 46B8 RAB Control Room HVAC 100 2
A 05-18-84 46D RAB Control Ventilation 100 1
A 02-17-84 46E RAB Chilled Water 100 2
A 05-25-84 48
. Containment Vessel 100 8
A 09-08-83 49 Process Analog Control 100 2
A 02-29-84 52A Reactor Coolant 100 18 A 03-03-84 53A Charging & Letdown 100 5
A 03-25-84 54-9 Primary Sampling 100 5
A 12-28-83 SSB Liquid Waste Management 100 3
A 08-03-84 55E Laundry Waste Management 100 1
A 09-21-83
-58 Refueling. Water 100 16 A 12-22-83 59 Containment Spray 100 3
A 12-05-83 60A High Pressure Safety Injection 100 2
A 03-29-84 60B Low Pre.ssure Safety Injection 100 3
A 04-02-84 60C Safety Injection Tank 100 2
A 09-02-83 61 Fuel Hdadling Storage 100 2
A 02-08-84 63 ESFAS '
100 1
A 03-09-84 65A-1,2 Nuclear. Instrumentation 100 8
A 04-20-84 66 Plant Protection 100 14 A 05-27-84 73 Emergency Feedwater 100 2
A 04-19-84 76 Steam Generator (incl. MSIV) 100 16 A 05-18-84 N-D-2 A
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-OPEN IE' BULLETINS, CIRCULARS, INFORMATION NOTICES AND NRR GENERIC LETTERS i
REQUIRED FOR FUEL LOAD ITEM ID #
TITLE STATUS IEC's 77-04 Inadequate Lock Assemblies One week prior to fuel load (security constraint)
IEB's 79-14 Seismic Analyses for As-Built Awaiting J. Tapia NRC Safety-Related Piping System Inspection Report GL's 83-28 Required Actions Based on Short-term corrective-Generic' Implications on Salem actions complete via ATWS Events letter to NRR - Remaining actions to be completed by 5% power.
NRC Resident Inspectors' consider actions taken on the above files to be adequate to support fuel load.
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Table D-3
'OPEN SIGNIFICANT CONSTRUCTION DEFICIENCIES TO BE COMPLETED PRIOR TO FUEL LOAD SCD/PRD #
DEFICIENCY DESCRIPTION STATUS ECD-DUE-DATE SCD-057 Inadequate I&C System Installation 10/05/84 and Turnover Documentation.
SCD-060 Turnover Documentation and LP&L action 10/22/84 Inadequate hanger weld critical to problems.
fuel load SCD-061 Linear. crack in SS Tubing.
LP&L action 10/05/84 critical to fuel load.
SCD-078 American Bridge RAB, FHB LP&L action 10/30/84 Bolting and Welding critical to deficiencies.
fuel load SCD-084 Tube track welding deficiencies.
LP&L action 10/05/84 critical to fuel load SCD-090 Electrical conduit overstressed.
LP&L action 10/26/84 critical to fuel load SCD-101 Traceability of Stainless Steel LP&L action 10/05/84 (SS) Instrumentation Tubing.
critical to fuel load SCD-105 Electrical Separation LP&L action 10/19/84
-Deficiencies.
critical to fuel load SCD-112' Design Changes Via Memoranda.
LP&L action 10/26/84 critical to fuel load SCD-114 Damage to Safety Related LP&L action 10/19/84 Equipment due to waterhammer.
critical to fuel load
. SCD-116 Failure of SUPS inverters.
LP&L action 9/30/84 critical to.
fuel load
- SCD-117 Limitorque Limit Switch and LP&L action 10/17/84 Motor Space Heaters.
critical.to fuel load i
D-4 L
Table D-4 OPEN LICENSING COMMITMENTS DUE BY FUEL LOAD COMMITMENT I.D.
COMMITMENT
SUMMARY
ECD WEEK OF SS5 1.0.1
. Address HED Findings B'5-F4/
10/15/84 and B7-F4 of DCRDR.
SS6 01.9B Establish and maintain in effect all Fuel Load provisions of the FRC approved physical security plan.
~W3B84-0475A Conditional Certifications for CE Purchase 09/30/84 Orders will be reviewed to determine if the operability of equipment was affected.
W3B84-0480AI Reinspection & corrective action for the 10/15/84 Steam Generator. Framing has been completed (ISSUE 12).
Coating work on
.the newly installed bolts is scheduled to'be completed by fuel load.
W3B84-0480A2 To assure. accurate scoping of SCDs, a 10/30/84
~
a review has been performed and results will be submitted as part of the SCD package (SCD-078).
W3B84-0480A5 A review for accountability of all 10/15/84 Mercury NCRs is in progress.
W3P84-0361A All design changes for the emergency In closure feedwater system identified to date cycle (02/16/84) will be fully implemented before W3 receives an Operating License.
W3PE4-1353/1.2 Preoperation testing of systems per 10/15/84 FSAR 14.2.12.2.9, 10, 14, 15 25, 52, 57, 58, 62, 63, and 78 prior to Fuel Load.
-W3P84-1412/1.8 Afil past audit deficiencies will be 10/15/84 reviewed and corrected prior to fuel load.
W3P84-1547, W
. hen reading the labeling.for the 10/15/84 LPSI-pump AMP indicator on CP-8, LP&L has agreed to band the instrument to aid in differentiating between normal a'nd abnormal indications, prior to fuel load.
D-5
Table D-5 NRC INSPECTION REPORT OPEN ITEMS DUE BY FUEL LOAD ITEM ID #
ITEM DESCRIPTION ECD-WEEK-OF 82-11-14 Item to be closed when various 10/15/84 Operations, Maintenance, Health Physics, Chemistry and plant Engineering Procedures are finalized and others are developed and completed 82-14-0A Failure to adequately control the 10/15/84 quality of safety related work.
83-08-009 Emergency Plan training will be 10/15/84 provided on a set frequency in the future (annually at a minimum).
84-05-03B NUREG 0737 (Item II.F.1, attach-10/15/84 ment 1 & 2) -Noble Gas Effluent Monitoring and Sampling.
84-20-04 Additional local Early Warning Fire 10/15/84 Detection is necessary around each auxiliary component cooling water pump because of high bay ceilings.
84-20-09 Emergency lighting.
10/15/84 84-20-15 Fire Protection / Preventive Program 10/15/84 84-20-17 Fire fighting equipment inventory.
10/15/84 84-20-21 Fire protection audit deficiencies.
10/15/84 34-01 Physical verifications for work 10/15/84 performed by Chicago Bridge and Iron.
.84-34-02A GEO Construction Testing - Compliance 10/15/84 of QA Program.
84-34-02B GEO Construction Testing - Conduct 10/15/84 review of supporting documentation for GEO corrective action.
84-34-02C' GEO Construction Testing - Reportability 10/15/84 of identified deficiencies under 10CFR50.55(e) or 10CFR21.
D-6 s
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Table D-5
'NRC INSPECTION REPORT OPEN ITEMS DUE BY FUEL LOAD ITEM ID #
ITEM DESCRIPTION ECD-WEEK-OF
~
84-34-03 CIWA Tracking.
10/15/84 84-34-04 Documentation findings of work performed 10/15/84 o-by American Bridge.
84-34-05 LP&L operations Q'A Transfer reviews.
10/15/84
'84-34-06 LP&L and Ebasco QA Programs for Plant 10/15/84 system'" status" and trnasfer reviews.
84-34-07 Nonconformance of 10CFR50, Appendix B, 10/15/84 Criteria V.
84-34-08 Evaluation and disposition was not 10/15/84 completed under NCR W3-5760, for undersized welds.
e D-7
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I T.tble D-6 TMI OPEN ITEMS REQUIRED FOR FUEL LOAD Item #
Item Description Status r
I.D.1
' Control Room Design ECD 10/15/84 This list identifies the TMI Open Item for which LP&L still has work to do before the.NRC can close this item and is required prior to fuel load.
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D-8