ML20128C448
| ML20128C448 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 11/19/1992 |
| From: | Rood H Office of Nuclear Reactor Regulation |
| To: | Horn G NEBRASKA PUBLIC POWER DISTRICT |
| References | |
| TAC-M84168, NUDOCS 9212040420 | |
| Download: ML20128C448 (176) | |
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UNI 1ED STATES NUCLEAR REGULATORY COMMISSION a
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November 19, 1992 g
Docket No. 50-298 Mr. Guy R. Horn Nuclear Power Group Mana ?r Nebraska Public F ser District Post Office Box 499 Columbus, Nebraska 68602-0499
Dear Mr. Horn:
SUBJECT:
AUDIT OF SAFETY-RELATED STRUCTURES AND CIVIL ENGINEERING FEATURES AT COOPER NUCLEAR STATION (TAC NO. M84168)
On August 17-20, 1992, the NRC staff conducted a cite audit at Cooper Station to assess the condition of the plant's safety-related strLetures and civil engineering features. The staff's report describing the audit is enclosed.
During the course of the audit, the NRC staff id:ntified a number of areas where the plant meets its oesign basis as described in the updated safety analysis report (USAR), but does not ineet present staff criteria for newer plants. While the staff does not require that these items be upgraded at Cooper, it does recommend that the Nebraska Public Power District (NPPD) cor. sider appropriate upgrading during (1) the ongoing design basis reconstitution work, (2) the resolution of Unresolved Safety issue (USI) A-46, or (3) during the individual plant examination of externally initiated events (IPEEE) margin evaluation.
in particular, three areas appear to warrant closer attention.
These are:
1.
Corrosion inside and cutside the torus - NPPD has identified spot pitting corrosion of the inside of the torus, and is monitoring the corrosion
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during each refuelina outagt.
The NRC staff also observed corrosion stains (streaks) along the outside periphery of the torus at one of the support columns.
NPPD attributed this te minor leakage from an overhead pipe.
The staff recommends that this condition be evaluated in conjunction with the corrosion inside the torus.
2.
he intake structure pump room for servi.:e water supply showed eroded concrete floor coating, corrosion on pipe supports, pumps and valves, and general moisture / water damage.
NPPD plans to perform some inodifications of the equipment and proper housekeeping of the room.
Based on its observations in this room, the NRC staff recommends that the supporting concrete structure, which is normally inaccessible, should be inspected du.ing an outage to assess its condition.
3.
At elevation 904 ft. in the control building, the residual heat renoval (RHR) service water booster pump gland water system was found to be in a 3
bad state of corrosion and degradation.
NPPD has been aware of the cond(tton for a number of years (since 1988 or earlier).
In the upcoming NA s
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9212040420 921119 PDR ADOCK 03000298
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Mr. Guy R. Horn 1993 outage, NPPD plans to remove the system.- The-staff recommends that NPPD develop and implement administrative procedures to assure that conditions adverse to quality of this type do not occur in the future.
The enclosed audit report includes a trip report with six attachments.
The attachments cover (1) a description of audit observations, (2) lists of attendees at entrance and exit meetings, (3) material presented by NPPD, (4) the NRC staff's walkdown routing and schedule, (4) a walkdown log, and (6) photographs of significant observatiuns.
This concludes the staff's audit of structures and civil engineering features at Cooper Nuclear Station and closes TAC Number M84168.
If you have any questions regarding this matter, please contact me.
Sincerely, ORIGINAL SIGNED BY:
Harry Rood, Senior Project Manager Project Directorate IV-1 Division of Reactor Projects - lil/IV/V Office of Nuclear Reactor Regulation
Enclosure:
As stated cc w/o enclosure:
See next page
,DISTRLB_UTION w/o enclosure:
(*w/ enclosure)
B E00cket Fils*
NRC & LPDRs*
PD4-1 r/f M. Virgilio J. Larkins P. Noonan ACRS (10) (P-315)
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A. B. Beach, RIV*
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6tFICIAL RECORD COPY Document Name:
C0084168.LTR
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Mr. Guy R. Horn.
1993 outage, NPPD plans to remove the system. The staff recommends that e
NPPD develop and implement administrative procedures to assure that conditions adverse to quality of this type do not occur in the future.
The enclosed audit report includes a trip report with six attachments. The attachments cover (1) 7. description of audit observations, (2) lists of attendees at entrance and exit meetings, (3) material presented by NPPD, (4) the NRC staff's walkdown routing and schedule, (4) a walkdown log, and (6) photographs of significant observations.
This concludes the staff's audit of structures and civil engineering features at Cooper Nuclear Station and closes TAC Number M84168.
If you 1. ave any questions regarding this matter, please contact me.
Sincerely, LRIGINAL SIGNED BY:
Harry Rood, Senior Project Manager Project Directorate IV-1 Division of Reactor Projects - III/IV/V Office of Nuclear Reactor Regulation
Enclosure:
As stated cc w/o enclosure:
See next page a
DISTRIBUTION w/o enclosure:
(*w/ enclosure)
Docket File
- NRC & LPDRs*
PD4-1 r/f M. Virgilio 1 Larkins P. Noonan ACRS (10) (P-315)
OGC (15B18)
A. B. Beach, RIV*
PD4-1 p/f H. Rood
- J. Roe H. Ashar D. Jeng H. Runyan OFC LA:PD4-l%I PM:PD4 D:PD4-l[N NAME PNoona'n}
HRood h JLarkinsb DATE II/6/92 l}/IS/92 11/IT 92
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0FFICIAL RECORD COPY Document Nan.e:
C0084168.LTR
l Mr. Guy R. Horn
.2-1993 outage, NPPD plans to remove the system. The staff recommends that NPPD develop and implement administrative procedures to assure that conditions adverse to quality of this type do not occur in the future.
The enclosed audit report incit. des a trip report with six attachments. -The attachments cover (1) a description of audit observations, (2) lists of attendees at entrance and exit meetings, (3) material presented by NPPD, (4) the NRC staff's walkdown routing and schedule, (4) a walkdown log, and (6) photographs of significant observations.
This concludes the staff's audit of structures and civil engineering features at Cooper Nuclear Station and closes TAC Num' er MS4168.
If you have any v
questions regarding this matter, please contact me.
Sincerely, x
Harry Rood, Senior Project Manager Project Directorate IV-1 Division of Reactor Projects - Ill/IV/V Office of Nuclear Reactor Regulation
Enclosure:
As stated cc w/o enclosure:
See next page d
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Mr. Guy R. Horn Nuclear Power, Group Manager Cooper Nuclear Station i,^
CC:
Mr. G. D. Watson, General Counsel Nebraska Public Power District P. O. Box 499 Columbus, Nebraska 68602-0499 Cooper Nuclear Station ATTN: Mr. Joh,n M. Meacham Site Manager P. O. Box 98 Brownville, Nebraska 68321 Randolph Wood, Director Nebraska Department of Environmental Control P. O. Box 98922 Lincoln, Nebraska 68509-8922 Mr. Richard Moody, Chairman Nemaha County Board of Commissioners Nemaha County Courthouse 1824 N Street Auburn, Nebraska 68305 Senior Resident inspector U.S. Nuclear Regulatory Commission P. O. Box 218 Brownville, Nebraska 68321 Regional Administrator, Region IV U.S. Nuclaar Regulatory Commission 611 Ryan Plaza Drive, Suite 1000 Arlington, Texas 76011 Mr. Harold Borchert, Director Division of Radiological Health Nebraska Department of Health 301 Centennial Mall, South P. O. Box 95007 Lincoln, Nebraska 68509-5007 mui ri i -
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EF_Q19SURE r
COOPER-NUCLEAR STATION TRIP ESPORT
Purpose:
Audit of Structures and Civil I:ngineering Features Location:
Cooper Nuclear Station Brownville, Nebraska Date:
August 17-21, 1992 Personnel:
D. Jeng (NRC), H. Ashar (NRC), M. Runyan (NRC),
R. Morante (BNL),
J. Bravermaa (BNL)
Backaround:
The objective of the plart visit was to obtain information about the performance of structures at operating plants and to draw some generic conclusions based on the information obtained from this and other plant visits.
To achieve this objective, an assessment of the existing condition and past performance of structures and civil engineering features at Cooper Nucletir Station (CNS) was performed.
Any failures, degradations, maintenance, surveillance, modification / repairs of safety related structures were of interest.
Structures reviewed include buildinga, tanks, cable tray and conduit
- supports, anchorages, underground structures, and the water intake structure.
Insnection Sunmary This section of the trip report describes the various audit-activities and the structures that were examined during the walkdown.
The section following this inspection summary presents the results and major observations noted during the audit.
A more complete list and detailed description of the observations are contained in Attachment 1.
August 17, 1992:
In the afternoon, the audit team met with W.L.
The NRC and BNL team was given instructions and tested in the use of protective clothing (anti-C's) and dosimetry equipment.
Then the BNL members of the team were given site-specific safety training.
The NRC members of the team already had prior site access training. Whole body counting was also performed.
A brief discussion was held regarding the audit schedule, meeting area, required walkdown supplies, and necessary camera passes.
1
l Auaust 18, 1992 An entrance meting was held with NPPD personnel and the-NRC/BNL represent tives.
This was. followed-by a
formal' presentation'by NPFi personnel.
The attendee list is contained in Attachment 2.
A c py of the agenda and viewgraphs used in the presentation is included in Attachment 3. _ The major. topics covered include design
- critoria, site geology / foundation conditions, settlement / structural boundaries, structural surveillance programs, Mark -I - torus, intake structure, spent fuel pool / racks, tanks, containment, and support anchorages.
Aucust 19, 1992 The audit team remained together during the plant walkdown activities this day.
NPPD personnel were assigned to guide us through the plant and to answer questite as they came up.
If questions could not be immediately answet they were recorded for followup by the plant staff.
The walkdown path, for the most part, followed the sequence described in the NRC Civil / Structural Walkdown list contained in.
The list describes the various structural components and areas to be walked down along with figures showing the location of each item.
This list, which was prepared by.NPPD prior to the arrival of the audit team, is based-on the list of structures and civil engineering features requested by the NRC.
NPPD was requested to prepare this walkdown path to. enable the audit team to review the many components and areas in an efficient and timely manner.
Additional areas were identified for examination which were not included on the walkdown list such as the Radwaste Building and Turbine Building.
The first area visited was the Reactor Building outsido containment.
Access inside containment was not possible since the plant was operating at the time.
The walkdown began at the top elevation-refueling floor and proceeded down to the lower elevations ending up at the torus. enclosure.
The building structural components examined include the concrete walls, floors, and ceilings and structural steel at elevations 1001 f t., 904 f t.,
976 ft.,
958 ft.,
931 ft.,
881 ft.,
and 859 ft.
Included in this review was examination of the drywell hatch, tanks, masonry walls, penetrations, conduit / supports, piping
- supports,
' equipment supports, anchorages, and the torus / supports.
Next, the team examined the roof and interior portions of the Control Building at elevations 948 ft., 932 ft'.,
918 ft., 904 ft.,
882 ft.,
and 877 ft.
The structural components reviewed include the concrete walls, floors, and ceilings; building structural steel; conduit / supports; cable tray / supports; equipment supports; pipe supports; suppor_t anchorages; and tanks.
Specific areas reviewed include the control corridor, roof, cable spreading room, 2
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I control room, -auxiliary relay room, battery room, DC switchgear room, and emergency condensate storaga tank.
While on the control Building roof, the exterior walls of the-Reactor Building woro -also examined.
l houst 20, 1992 To ensure completion of the walkdown in the allotted time, the audit team was split into two groups; A and B.
Team A, which consisted of H. Ashar and R. Morante, examined the interior of the Radwaste Building at elevation 875 ft. and basemat and the interior of the Turbine Building at elevations 932 ft.,
875 ft.,
and 903 ft.
NPPD had informed the audit team that the only safety related area in the Radwaste Building is at the lowest level (basemat) and the Turbine Building has no safety related arca/ components.
Team A also reviewed the interior and exterior portions of the Intake Structure.
While Team A was performing the walkdown described above, Team B,
which consisted of D.
- Jeng, M.
- Runyan, and J.
Braverman, examined the diesel generator rooms, exterior walls of the Control Building and Reactor Building, the buried diesel oil tank area, and the elevated release point tower.
Dr.
B.
D.
Liaw, Deputy Dizoctor DET/NRR, joined Team B in these walkdowns.
Team B also examined the exterior and interior portions of the Intake Structure.
Both teams examined the exterior and interior, portions of the Intake Structure at the grade level (operating floor elevation).
Access below the operating floor down to the water level could not be readily obtained.
- However, visual examination from the operating floor of all the bays in front of the traveling screens down to the water level were made.
In addition, walkdowns were performed in the service water pump room.
Structural components reviewed in the areas described above include concrete floors, walls, and ceilings; building structural steel; conduit / supports; cable tray / supports; piping supports; anchorages; seismic gaps between buildings; equipment supports; tanks; and masonry walls.
This concluded the walkdown activities.
During all the walkdowns, a log was maintained, as sho"n in,
in which the team recorded for each observation the building / area, elevation,
- location, component / item, aspect-reviewed, photograph number, and any comments.
Data were recorded for structural components where aging degradation effects were present as well as where they were not.
Photographs were taken for selected items to enhance the documentation; these were noted in the log.
In addition, measurements were taken when appropriato (such as crack size), to determine the severity of the degradation.-
3
In the afternoon, the audit team reviewed several LERs and 10 CFR 50.59 evaluation packages related to civil / structural items.
These were selected from a list provided by NPPD during their presentation.
Throughout the audit, NPPD personnel provided responses ar.d documents in an effort to address and resolve many of the questions and concerns raised by the audit team during the formal presentation session and during the walkdowns.
The audit team then reviewed and discussed the observations noted.
A list of the more meaningful observations, including those that would be of benefit to NPPD was compiled.
This list, which is discussed in the next section of this trip report, was conveyed verbally to NPPD at the exit meeting.
Results/ observations The exit meeting was held in the af ternoon of August 20, 1992.
A list of the attendees at this meeting is contained in Attachment 2.
At the start of this meeting, B.D. Liaw made some introductory Ftatements on the purpose of the NRC audit and the general condition of the plant.
He also stated that the lack of timely management action to resolve the degraded condition of the service water booster pump (SWBP) gland water system is of concern. Then, D. Jeng summarized the observations noted as a result of the formal presentation given by NPPD and the walkdowns performed by the audit team.
It should be noted that the observations presented to NPPD were for their information and did not represent requirements by the NRC.
Any action the licensee might take as a result of these observations is considered voluntary.
While most of the civil / structural plant features examined at CNS are in very good condition after 18 years of operation, there were some components which did show varying degrees of aging degradation.
Some of the observations are discussed below, with a more complete list and detailed description presented in Attachment 1.
A few items of concern which were not aging related were also brought to NPPD's attention (e.g.,
insufficient thread engagement or loose concrete anchor bolts, and the adequacy of anchors used with epoxy in high temperature areas).
Signs of corrosion were observed on the external surface of the torur shell.
Apparently a water leak from above the torus had run down the torus shell wall removing the reddish coating and initiating some surface corrosion.
As a result of internal inspections of the torus performed in 4 of the 16 bays, NPPD identified numerous corrosion pits.
The existing pits were analyzed considering a corrosion growth factor and then 150 pits were patched using underwater cured epoxy.
NPPD 4
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a indicated that. the - epoxy, when applied properly, will displace water and trapped air in-the repair zone.
The patches will be evaluated during the 1993 - scheduled refueling outage to confirm their adequacy and the use of the underwater applied epoxy method.
Examination of the exterior concrete walls'and basemat floors of most structures showed no signs-of water infi.'.tration.
- However, evidence of water infiltration was observed through the exterior concrete wall of th. Rea;cor Builoing when viewed from inside the torus enclosure.
Apparently, water leaks in at a
piping penetration and runs down the concrete wall onto some conduit.
NPPD was aware of this situation, indicatirJ the source of the water is from surface runoff during heavy rainfall.
However, it appears that no effort hac been made to stop the water infiltration.
At elevation 904 ft. in the Control Building, the SWBP gland water system is in a very bad statt of corrosion and degradation.
This includes the tanks, piping, supports, and anchorages.
NPPD was aware of this situation, indicating that they plan to remove the system completely during the 1993 refueling outage.
- However, the_ system has been allowed to continue to operate in this condition for a number of years.
At elevation 882 ft. in the Control Building, corrosion of piping, valves, and flanges / bolts was observed in components of the Service Water Booster Pump System.
Corrosion of the pipe supports and anchorages as well as degradation of the floor coating were observed in the service water pump room of the Intake Structure.
Outside the pump room, water was collecting on the operating floor; some of which was coming from the partially closed traveling screen side door.
Ultrasonic testing (UT) is being utilized to measure pipe wall thickness in the sc.rvice water piping for the diesel generator lube oil system.
This was done due to erosion of the piping at elbows caused by the silty Missouri River water.
NPPD indicated that UT of the service water and residual heat removal service water booster system piping, fittings, and valves is routinely performed as part of the Augmented Erosion-Corrosion Program.
The coating at one of the structural steel legs of the elevated release point tower has degraded and corrosion of the base material is evident.
Photographs for some of the key observations discussed above and listed in Attachment 1 are presented in-Attachment 6.
4 Although there is no formal ' inspection program for all structures, NPPD described their Surveillance Procedures 6.3.10.1 and 6.3.10.12 for inspection of the interior and exterior-drywell 5
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a and ' torus.
In additilon, NPPD developed CNS Procedure 0.11
- entitled,
" Station Inspections".
This proc. dure defines the requirements for conducting station inspections by station management.
These _ inspections cover the_ overall material condition, fire hazards, safety and industrial -hazards, design deficiencies, radiological
- controls, and general housekeeping coditions.
This procedure or others should be expanded to more specifically include items such as corrosion,
- leake, concrete
- cracks, coating failures, water infiltration and other aging related effects.
NPPD currently performs 10 CFR 50.59 evaluations in accordance with their CNS Engineering Procedure 3.3, Rev. 11.
This procedure provides the method for determining whether a proposed change, test, or experiment constitutes an unreviewed safety question or requires a Technical Specification change as defined in 10 CFR 50.59.
The procedure defines the responsibilities and requirements for the preparation of the safety evaluation.
Several Station Design Change packages were selected for review.
Each of the packages did contain a Safety Evaluation and the 10 CFR 50.59 Reportability Analysis as required by the CNS Procedure 3.3 Conclusion Considering that the plant has been operating for approximately 18 years, most civil / structural plant features have performed very well.
Some structures / components, however,- do show signs of varying levels of aging degradation.
The most severe cases relate to corrosion of the torus exterior surface,-pitting due to corrosion of the interior surface oL the torus, corrosion of the SWBP gland water system, and corrosion and degradation of piping, supports, anchorages, and floor in the service water pump room.
Better maintenance and housekeeping, particularly of the SFBP gland water system in the Control Building and the service water system in the Intake Structure are recommended.
In addition, inspections of Category I
structures should be performed periodically to identify any aging related effects steh as corrosion,
- failures, leaks,
- cracks, settlements,.and water infiltration.
One area that typically receives ?.ittle attention but which should be inspected is the Intake Structure; particularly in the areas that are not visible or readily accessible (i.e.,
below the floor at grade elevation and below the water line).
The use of the Missouri River silty water has resulted in internal erosion of piping.
NPPD has an Erosion-Corrosion Program and an Augmented Erosion-Corrosion Program to ultrasonically-test the piping for pipe wall loss.
It is important to continue this effort to identify any degradation which would develop over time.
6
Te address the concern of corrosion pitting on the inner surface of the torus shell, appropriate inspections, evaluatio;-t and repairs should be performed. -All bays should be checked for pitting; past epoxy patching should be evaluated for the success of
-retarding additional corrosion growth; and evaluations should be made to. determine how to-prevent new pits / corrosion from developing.
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ATTACHMENT 1 COOPER NUCLEAR STATION DESCRIPTION OF CIVIL / STRUCTURAL OBSERVATIONS r
1.
Signs of corrosion were observed on.the external surface of the torus shell, near torus support #7.
The normally reddish color vf the torus shell now had dark streaks -on it.
Apparently, a leak from above the torus has been running down the torus shell wall, on both the inner and outer side.
This has removed the reddish coating and has ir.itiated surface corrosion of the shell.
2.
Evidence of water ' leakage through a piping penetration in-the exterior concrete wall of the Reactor Building was observed from inside the torus enclosure.
The water apparently ran down alcng the wall onto some conduit adjacent to the wall.
NPPD was aware of tais situation, indicating the source of the water is from surface runoff during heavy rainfall.
- However, it appears that no effort has been made to stop the water infiltration.
3.
In the service water pump room of the Intake Structure,_
corrosion of piping and pipe supports as well as degradation of anchorages and floor coating were observed.
-Better housekeeping and maintenance of the equipment,
- piping, supports, anchorages, and floor is recommended.
In addition, in view of the condition in the service water pump room.and excessive water on the operating floor outside the pump room, inspections of the concrete and steel structure should be periodically performed; particularly in areas that are not visible or readily accessible.
4.
At elevation 904 ft, in the Control Building, the RHR service water booster pump (SWBP) gland water system is in a very bad state of corrosion and degradation.
This includes the tanks, piping, supports, and anchorages.
_NPPD was aware - of this condition, indicating that they plan to remove the system completely since it is-not required.
NPPD stated that they plan to remove it during the 1993 refueling outage.
- However, the system-has been allowed to continue to operate in this condition for a number-of years.
At the request of the NRC, NPPD prepared a_ chronology of the events and steps taken to address the condition of the RHR SWBP gland-system.
A copy of this chronology is presented _at the end of this Attachment.
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At elevation 882-ft, in the Control Building,. the service water booster-pump systen was observed to be in a' degraded condition.
Corrosion was identified on the piping, pump B, valve SW-85, and on several flange connections.
6.
At elevation-87?
ft.
in the Control-Building, several horizontal pipes near the ceiling-over the emergency storage tank appear to show signs of corrosion.
7.
Several instances of cracks in concrete floors and walls were observed.
These include the floor at elevation 882 f t. in the service water booster pump room and West wall (floor to ceiling); exterior wall of the diesel gone.vator structure on the North end and East end; floor _at elevation 958 ft. of the Reactor Building (MG pumy area);- and the crack / spalled concrete of the Reactor Building exterior wall at the junction of upper and lower levels (North East corner). Although these cracks are not severe, it would be prudent to monitor them to ensure they do not worsen and lead to degradation of the structure.
t 8.
A number of instances of potential connection / anchorage problems were observed.
These include:
missing bolt / nut at cable tray splice plate in the cable expansion room, (cable tray C227 near hanger. _ #65),
conduit _ support anchor angularity about 10 -15' (excessive) and a gap exists between the anchor - head-and plate in the. cable - expansion room, insufficient thread engagement of anchor nuts in a base plate--
for the service water line near valve SW-125 at elevation 931 f t. of the Reactor Buildina, and missing anchors at base angle attached to corrugated steel supports to a masonry Wall at elevation 931 ft. of the Reactor Building.
9.
At the base of the structural steel-leg of the elevated-release point tower (horth West corner) the coating has peeled off and corrosion of the base material is evident.
10.
If epoxy dipped anchors are used in high temperature areas, then evaluation for possible hardening and-brittleness of the epoxy and the capacity of the anchors should be-evalucted.
In addition, the letter to the NRC regarding-IE Bulletin 79-02 Resolution describes criteria and plans to resolve the IE Bulletin but does not give any schedule for its completion.
' 1.
As a result of past internal inspections _of-the torus,:NPPD identified -umerous cases of corrosion pits in 4 of the 16 bays examined.
The existing pits were analyzed with a corrosion growth factor which would conservatively allow for all pits to remain uncoated until the 1993 scheduled refueling outage.
- However, NPPD decided to patch the 150 pits identified in the 4 bays using underwater cured epor.
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cont'd The audit team expressed a. concern about pits in other bays that were not inspected and about the. ability of the underwater cured epoxy to arrest the problem.
NPPD indicated that the epoxy used was Brutem 15 and, when applied properly, will displace water and trapped air in the repair zone.
The patches will be evaluated during the 1993 scheduled refueling outsgo and this evaluation should provide justification for the use of underwater applied epoxy.
12.
NPPD indicated that ultrasonic testing (UT) of the service water and residual heat removal service water booster system piping, fittings, and valves is routinely performed as part of the Augmented Erosion-Corrosion Program.
This program is concerned with lower pressure systems that could be subjected to erosion caused b.v the silty Missouri River water.
The Augmented Erosion-Corrosion Program contains those items which are potentially subject to wall loss, but are not flagged by the critoria of the Erosion-Corrosion Program.
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4 RHR SWBP GLAND SYSTEM CHRONOLOGY May 1988 STP 87-011 " Post - LOCA Service Water System F. low Test - determined that RHR SWBP Gland Water System was difficult to operate.
Oct. 1988 EWR 88-278 placed on hold because it was presented af ter the cut-off date for the 1987 Scope List.
6/21/90 EWR 90-174 Rev. I submitted 10/25/90 STP 90-174 approved 11/30/90 STP performed.
Tested SWBP without mechanical seal injection.
Results were successful.
5/16/91 CDD 90-174A prepared (Conceptual Design Document) 8/15/91 DC 90-174A prepared 9/19/91 DC 90-174A SORC approved 1991 outage DC 90-174A partially installed August 1992 DC 90-174A, Amendment 1, prepared for review DC 90-174 A modification of pumps (one at a time) scheduled to begin
ATTACHMENT R COOPER NUCLEAR STATION ENTRANCE MEETING
+
August 18,1992 NMiE TITLE PHONE lim Raherts CNS Enrineering Afer.
(402)825 5420 Afile Dean licensine Sunenisor X 5663 Dasid N. Afadsen Licensine Enrineer (402)S63 5005.llliam L Swant Project Afanarer (402)S63-5354 E. Af. Aface Sr. Afer. Site Survort X-5324 LAf. Afracham Site Afanarer X-5775 ~ R. L Wen:1 NED Site Afanarer X 5747 R. D. Aussurrer Exad Civil Enrineer (402)S63 52]p Roter Afoberiv Site Elec.li & C Enrineer (402)S25-5676 Afite Siedlik CirillStructural Supervisor (402)S63 S646 David C. lene Section Chief DET/NRR/NRC (3011504 2727 Joseph Rmuerman Research Enrineer (S16)282-2186 ilans Ashar Sr. Civil Enrineer - NRC (301)SO4-2851 Afichael Runvan Reactor insnector. NRC-RIV (817)S60-8142 Richard 1. Afomnte Sr. Research Enr. (Brookhaven Nat'lisb) ($16)282-S860 Rick Gardner Plant Afanacer (402)S25-S233 R. L. Yant-Senior Civil Enrineer X S713 Brent Aforlier Sr. Aftce. Technical Enrineer (402)825-5717
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AT,TACHMENT 3 CNS STRUCTURAL VERIFICATION AUDIT PRESENTATION OUTLINE 1) INTRODUCTION 3WANTZ 2) PIANT OVERVIEW GARDhTR General Description District Commitments 3) SCSIGN CRITERIA AUGSPURGER Design Codes / Standards used for Class IS F' uctures Seismic Design Criteria Wind / Tornado Design Criteria Flood and Dam Failure Criteria 4) SITE GEOLOGY / FOUNDATION CONDITIONS: YANT4 Discussion of Site Geology Excavation and Foundation Design Criteria Ground Water Level Water Proofing 5) SETTLEMENT ISSUES:/STRUC'IURAL BOUNDARIES YANTZ Foundation Design Structural Boundaries 6) HELB EVALUATION SIEDLIK Jet Impingement Loads Containr..ent Penetrations Pipe Whip Major Modifications within Containment 7) STRUCTURAL SURVEILLANCE PROGRAM MOELLER Procedure 63.10.1 Procedure 63.10.2 Current Status 1
). 8) MARK I TORUS HYDRODYNAMIC LOADS HISTORY AUGSPURGER Statemer.t of Original Problem Summary of Modifications Closure of the Issue Surveillance Requirements Protective Coatings Ob.:rved Torus Coating Degradation Remedial Actions IE 84-01, IN 88-82 Discussion 9) INTAKE STRUCTURE MOBEELY GL 8913 Discussion Diver Inspections 10) SPENT FUEL POOL AND RACKS YANTZ Original Construction High Density Rack Mods Cask Drop Accident Scenario including Crack leak Rate an.d Make-Up Capacity 11) HEAVY LOADS AND ".EFUELING FLOOR ACITVITIES MOELLER Procedure 7.0.3 Procedure 10.26 Refueling Floor Crane Operation 12) SAFETY RELATED STORAGE TANKS AUGSPURGFR Design Criteria List of Safety-Related Tanks Anchorage Descriptions Settlement Issues Included in A-46 Scope 13) TANK INSPECTION AND MAINTENANCE AUGSPL91GER Preventive Maintenance Past Tank Problems / Leakage Future Actions 2
14) BURIED PIPING YANTZ Design Criteria Contract 69-4 Specifications Cathodic Pmtection 15) CONTAINMENT TESTING MOBERLY 16) TESTING OF CONTAINMENT BELLOWS MOELLER Required by 10CFR50 Appendix J Testing and Monitoring 17) SUPPORT ANCHORAGES SIEDLIK Anchor Types Design Criteria IE Bulletin 79 02 Piping and Pipe Support Qualincation Project IE Information Notice 80 21 18) SEISMIC INSTRUMENTATION MOBERLY Equipment Description Maintenance and Testing 19) PLANT SAFETY PROCEDURES MOBERLY - EP 5.1.1 Earthquake EP 5.1.2 Tornado Watch EP 5.1.3 Flood EP 5.1.7 Emergency Classification 20) USI A-46 ACTIVITIES AUGSPURGER Brief History of the Issue NPPD Future Plans for Resolution 3
) 21) MASONRY WALLS YANTZ 22) DEGRADATION OF STEEL CONTAINMENT SIEDLIK Background - IN 86 99 & GL 87 05 Comparison of Oyster Creek and CNS CNS Surveillances/ Inspections 23) 10 CFR 50.59 EVALUATIONS AUGSPURGER Short Summan of Each of the 10 DCs Chosen by NRC 24) IPEEE ACTIVITIES AUGSPURGER Scope NPPD Response to GL 88-20 Supp. No. 4 Current Status 25) LER CORRECTIVE ACTIONS MADSEN LER 9120 LER 86-28 26) INSPECTION REPORT CORRECTIVE ACTIONS MADSEN SSFI 50 298/8710 27) DISTRICT PLANS FOR LICENSE RENDVAL MADSEN 4
I_OPIC 3 DESIGN CRITERIA Design Codes / Standards used for Class IS Structures Seismic Design Criteria Wind / Tornado Design Criteria Flood und Dam Failure Criteria l
DESIGN CRITERIA Design Codes / Standards for Class IS Structures Loading Condition Allowable Stress D+E Normal code allowable stress (AISC for steel, ACI 318-63 for concrete, ASME Section III for primary containment) D+W Normal code allowable stress with 1/3 increase D+R+E Normal code allowable stress. 1/3 increase not permitted. For jet impingement on-primary containment, 90% of yield at 300 deg. F. D+E+F Local membrane stress may exceed yield with margin against rupture D+T 90% of AISC Code allowable for steel and Ultimate Strength - Design of ACI 318-63 with load factor of 1.0 for concrete t7--e r w ++-=d w-- 'y=9y
DESIGN CRITERIA (cont'd) Loading Condition Allowable Stress (cont'd) D + R+ E' 90% of AISC Code allowable for
- steel, Ultimate Strength i
Design in ACI 318-63 with load factor of 1.0 for concrete. For jet impingement on primary containment, 90% of yield at 300 deg. F. D= Loads from normal operating conditions E= OBE loads E' = SSE Loads R= Loads resulting from jet forces and pressure and ' nts associated with rupture of temperature trarx. a single pipe within primary containment F= Loads due to flooding drywell 23' above drywell flange level W= Design wind loading conditions T= Loads due to the effects of a tornado
f DESIGN CRITERIA (cont'd) Seismic Design Criteria Located in UBC Seismic Zone 1 PGA 0.lg OBE, 0.2g SSE Vertical acceleration is 2/3 maximum horizontal ground acceleration Seismic forces applied simultaneously in vertical and worst case horizontal directions Spectra for Class IS structures generated using time-history method (July 21,1952 Taft, CA earthquake) e damping values (percent) Reinforced concrete 5.0 and 7.0 Steel frame structures 2.0 Welded assemblies 1.0 Bolted assemblies 2.0 Vital piping systems 0;5 I i',
a DESIGN CRITERIA (cont'd)- ) u -Wind Load All structures can-withstand 100 mph winds Wind forces calculated in accordance with ASCE Paper 3269 " Wind Forces on Structures" Tornado Wind Loads Tangential velocity of 300 mph Transverse velocity of 60 mph-3 psi pressure drop over 3 second time interval Tornado wind loadings-(missile excepted) are combined with functional loads h c ,-.m .-_.----,,-..L-~.....,m.._- .._.,,.,L,..;,. ,ym.,, y _.,._m. ,m., ,,.,y,_.
. DESIGN CRITERIA (cont'd) Tornado Generated Missiles Class I structures designed to provide protection from the fenowing missiles: 35' iong utility pole with 14" butt at 200 mph 1 ton missile at O mph over 25 square feet 2" extra heavy pipe,12' long Any other missile resulting from failure of a structure or one which has potential to be lifted from storage or working areas Missile loading considerations based on Bates and Swanson paper " Tornado Design Considerations for Nuclear Power Plants", presented at ANS Annual Meeting,1967 i e-Local crushing and opposite face spalling perraitted l Panels above 1001' level designed to blow off at 100 mph l l
DESIGN _ CRITERIA (cont'd) Flood Protection Station site grade level is 13' above natural grade of 890 feet MSL, e Grade floor elevation is 903'6" Flooding considered highly unlikely due to the combination of upstream flood control and high final site elevation Return Frequency Discharge El. 1,000 yr. 900' MSL 10,000 yr. 902' MSL 1,000,000 yr. 903' MSL Dam break not wnsidered probable Only time that river levels are postulated to exceed 903'6" is ailer unlikely failure of dam (nearest dam is 275 mi. from site). If this occurs, river levels of 905 to 906 ft. MSL are expected Emergency Procedure 5.1.3, " Flood" for ( sandbagging and wood barrier installation procedure
DYNAMIC MODEL ) .ELEVu 1047'-Cf O I ~ 1001'-0" g 2 drywell reactor bldg.% 97 d -0" O ag= 977'- 3 IM" J 96l'-10 3M" 0 958'-3" 04 o 950'-0" 11 0 93 8' - 7"- 12 9 31'- 6" ( ) 5 0 9 21'- 8" 13 O 907'-3" 903'-6" )6 g exterior wallb 0 BG 2'- 5" 15 y N \\ / L 87 9'- 0" W VN 0 7 885'- 3 t/8" 8 S 4N ..'.!*((@))7 854' - 9" Yv M Ky \\ NWvyg yf/gN4// AN N Nv/ h,N ' " ( Nebraska Public Power Distnet COOPER NUCIIAR STATIO? UPDATED SAFETY ANALYSIS REPO% JF i Reactor Bldg. Dynamic Analysis Me Figure C-2-1
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) t h h TOPIC 4-i SITE GEOLOGY / FOUNDATION CONDITIONS Discussion of Site Geology Excavation and Foundation Design Criteria. Ground Water Level Water Proofing t e v-*yvy--,e.v-,v,- m ew y -*,we +-y3 - w s r-W w -r-e - eg w,y ++--- y wv -t m y r y,syvrr-w w w i,yr w v M,- v-g my v w--,w- ,,-_.,.ra-r-
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SITE GEOLOGY / FOUNDATION CONDITIONS SUBSURFACE CONDITIONS CNS is located on the Forest City depositional basin. The Nemaha Anticline forms a border approximately 20 miles West of site. Humboldt Fault is within this transition. Pre-Cambrian granite underlies the basinal deposits. ,~,
) FOUNDATION CONDITIONS Alluvial deposits in the flood plain at the site vary in thickness from 62 to 71 feet. Material was excavated to within 7 to 8-1/2 feet of bedrock. Near surface clay pockets were then excavated and replaced with structural fill sand. In-situ compaction performed brought the relative density to at least 75%. Compacted structural fill placed to provide base for foundations of major structures. All major structures except Diesel Generator Building have concrete mat foundations. The Diesel Generator Building has spread footings for outer walls and inner columns. Design allowable soil bearing pressure is 18 ksf. ,,a
~ GROUNDWATER EFFECTS Recent soil borings show groundwater at 882 ft. Design level of groundwater is 895 ft. All major structures have external waterproofmg membranes under base mat extending to elev. 900 ft. Piping penetrations below 900 ft. are sealed by Neoprene rubber boots bonded to pipe and the cast-in-place wall sleeves. Underground electrical duct banks use PVC pipe encased in reinforced concrete. Recent heavy rainfall has resulted in small leaks into the Reactor Building Torus Area. All major structures have drains and sumps in the base slab to collect any water. 91YWI . ~,.,,.. _ _
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TOPIC 5 SETTLEMENT ISSUES / STRUCTURAL BOUNDARIES Foundation Design Structural Boundaries
SETTLEMENT ISSUES AND t STRUCTURAL BOUNDARIES Foundation Settlement
- Class I structures are founded on compacted fill, therefore, settlement was not a-concern.
- Expected settlement was small and mainly occurred during construction phase.
- Since completion of construction, no visible signs of abnormal or unexpected foundation settlements have been observed.
Structural Boundarie.s Structures are separated by 2 inch or 4 inch gaps to account for relative movements. Calculated movements in areas with 2 inch gaps range from 1.06 to 1.56 inches. Calculated movements around Controlled Corridor range from 3.56 to 3.95 inches. No formal observations or measurements of free spaces have been made. m- .-n.- y9.y.y. ,.m,,-w-- ,-17 9 ,-..p-4,aw,- y 4 ,g w 9 ue=--:
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- Rationale for not having such data:
No visible signs of differential settlement. t No prior seismic activity, i Any possible structure interaction le.ss severe than effects of tornado generated missiles already considered in design. Drywell Air Gap
- Drywell is separated from surrounding concrete by 2 inch free space above elev. 888 ft.
- Transition zone at elev. 888 ft. is a sand filled area.
- The gap allows shell to act independently-yet provides backing for limiting deformation from jet-impingement loads.
- Gap formed during concrete pours by controlled process.
F
TOPIC 6 HELB. EVALUATION Jet Impingement Loads Containment Penetrations Pipe Whip Major Modifications within Cbntainment
HELB EVALUATION AND PROTECTION FOR THE STEEL CONTAINMENT j Possible liner damage mechanisms due to a high energy line break
- Direct impingement on the wall by the jet of fluid issuing from the broken pipe
- Reaction forces of the jet acting on containment penetrations
- Impact of a pipe that is moved by jet forces (pipe whipping)
JET IMPINGEMENT LOADS Interior Area Jet Force Subjected to 1 Maximtun Jet Force Location Spherical part of drywell 660 kips 3.7 sq ft l Cylindrical part of drywell 468 kips 2.6 sq ft I and transition to sphere l Closure Head 33 kips 0.18 sq ft 2.3a W _
REACTION FORCES ON _ CONTAINMENT PENETRATIONS
- Coaxial guardpipes and rings at the inside end of the guardpipe serve the function of jet deflectors
- The rupture loading (R) utilized to design the penetrations is a function of (MU) the ultimate bending moment of the process pipe and (P) a force equal to the process pipe pressure multiplied by the cross-sectional area of the pipe.
- Limit stops were utilized in the vicinity of the flued head fittings when required by analysis. The loading combination which controlled the design of the limit stops = DL + R
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PIPE WHIP The recirculation piping system has been provided with whip restraints. Tornado siding is installed at strategic locations on the inner wall of the drywell.
- Corrugated 12-gauge steel plate sandwich which 6
can plastically deform to absorb.93x10 ft-lbs of kinetic energy per square foot
- The intent of the criteria of Reg. Guide 1.46
" Protection Against Pipe Whip Inside Containment" was met in selecting the pipe break locations Physical separation is provided between redundant engineered safety systems. This spatial separation precludes concurrent damage to more than one redundant safety feature by a single postulated pipe failure. ASME Section XI in-service inspection of 100% of the circumferential welds at all terminal points where pipe breaks are postulated. 1 e
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t MAJQR MODIFICATIONS 1 WITHIN-CONTAINMENT
- Pipe Replacement Project - 1984 to 1985 9
Recire, Piping Core Spray Piping RWCU Piping RHR Piping (Small Portion)
- Large Bore Pipe Support Project - 1988 to 1993
- Reactor Recirculation Pump Upgrade - 1991 Major modifications performed in the drywell have not-changed the original piping configurations or HELB design criteria, therefore, the CNS licensing basis has not been changed.
7
i r TOPIC 7-1 STRUCTUR.AL SURVEILLANCE PROGRAM F Procedure 6.3.10.1 t Procedure 6.3.10.2 Current Status 4 i .,c.- -.--...- .2--..
~. Structural Surveillance Program Required By Tech Spec. Procedure 6.3.10.1, Drywell And Torus Surfaces And Structural Elements Inspection Procedure 6.3.10.12, Terus Exterior Surfaces And Supports Inspect. Current Evaluations In Progress. 1 ._..r
1 SlEllfilintl_S11EYrilht11te_Atid_Illspection Progrant Implementation Required By. Tech Spec 4.7.A.1.d; A visual inspection of the suppression chamber interior, including water line regions, shall be made at each mtsor refueling outage. t Tech Spec 4.7.A.2.h; The interior surfaces of the drywell and torus shall be visually inspected each operating cycle for evidence of torus corrosion or leakage.
- Tech Spec 4.7.A.I.c; Whenever there is indication of relief valve operation with suppression pool temperature 2: 160 'F and reactor pressure > 200 psig, an external visual examination of the suppression chamber shall be conducted before resuming power operation.
Surveillance Procedure 6.3.10.1, Drywell And Torus Surfaces And Structural Elements Inspection. Surveillance program automatically initiates the performance of the subject procedure during each refuelling outage. Procedure identifies items of inspection for the suppression pool and drywell internal surfaces, documents inspection fmdings, and initiates corrective action if degradation is determined. Surveillance Procedure 6.3.10.12, Torus Exterior Surfaces And Supports Inspection. Surveillance program automatically initiates the performance of the subject procedure on a six (6) month cycle. In addition, the subject procedure is performed after seismic events, relief valve blowdown,= or other events that may create damage. Procedure identifies items of inspection for the exterior of the torus, documents discrepancies, and initiates corrective actions. l l L
i Sinictural Surveillance And Inspection Program (continue) Current Evaluations In Progress. The suppression pool's internal surfaces have documentation of paint degradation and corrosion. Engineering evaluations are currently in progress to disposition the corrosion area and restore the paint degradation. An extensive inspection is to be performed during the Spring 1993 Refueling outage. t u_-_-.._.._..,_,-_.._._..~._,___.-.--.__
4 TOPIC 8 MARK. I TORUS HYDRODYNAMIC LOADS HISTORY Statement of Original Problem Summary of Modifications Closure of the Issue i Surveillance Requirements i Protective Coatings Observed Torus Coating Degradation Remedial Actions IE 84-01, IN 88 Discussion L [- ._....-..~..,2__._...
MARK _I_ TORUS HYDRODYNAMIC, LOADS ISSUE Since the establishment of the original design criteria, additional loading conditions associated with the pressure-suppression concept utilized in the Mark I containment system design were identified Additional loads result from dynamic effects of drywell air and steam being rapidly forced into the suppression pool during a LOCA and from suppression pool response to S/RV operation l
Table 1.t* Sutesa cf CrmtAtwwtti At _ nG Mrx!fif AT(C_H5 s t CDNPONENT fuM NAluFE OF *weiflCATION re!Mt2,10A0 Cs sysst>St CCrettitcw entt SIRUCitfRAL COMr94EN15 forus Snell aad S e ports focus Support Colaan Plate reinfortet to cotten web and flanges Increase column capacity Spring 77 Colisen Anchorage Installed anchor bolts, brackets and bos tras assemeties Re.lst LOCA aptift forces sprieg 77 . Cohen-to-Torus Connet t ien Additional f=II panetration weldment increase connettlen capacity Spring 7E forvs Sa414e ivlt sa&fles connecting torus swport columns leprove dynamic response 5 asmer El f Coleman Anchorage 3teinfertenant Reinfortement of ben beams and I:recket we%ts Resist LOCA estif t forces Serieg 82 Einvy Girder Wh stif feners; local reinforcement of weld to shell Feel drag loads; attachment leads fall 81 Vent Systeu vent steaJer/Dnwncnmer intersection Reinforted 80 p wtrations with st!ffener plates and pads Chegging & S.TF discharge leads Fall 81 i Downroeers Reduced downcomer submergence by truncation Mitigate CSA blowdown lead Spring 80 Downumer Ties Installed tie bar and ring assembly at each har pair CD and thegging lateral leads Spring 80 Vent N a&r Dette<ter lastalled d=fie(tor assembly in all torus bays ieel swell lepect lord Sprima 83 I Dent Itender Supports Removed esisting supports; res o ported free girder above Peel drag loads Spring 81 DW/W vacuus Breakers Reinforced 12 vecteen brester penetrattens Feel swell / froth impact Isads Fall 81 Miscellaneous lorus Internals Moeorait Installed sidbay stepe-ts in all teres bays Froth tapingement lead Scris,jp El 3 Service Platfa s Replaced esistirq sis, ports; added eev seports, bracing Pool swell (apact/ drag leads Socing 87 i i and 7 akkag kie* b 5 l 1 Drywell Steel f raming Reinforcement of beae seat connections and fraeing S/RY pipe s gpert leads Sprdeg 87 eead>ers %T-MISCtttAMEOUS SYSTEM PFX)lfICATIONS O-Q - Drywell,Netwell Pressere Differential' Installed Pump Around System Mitigate tCCA blewdown went clearieg. Spring 16 f orm temperatwee Monitoring Systee Installed esattoring system and ivrstrumentation Monitor pool temperature Sussear 82 C i 5/#V tow-to. Set logic Will Install control logic and instrumentation for safety Mitigate / eliminate 5/WV subsegweat Spring 81) ~ ..e relief valves actuation leads g. O M5tv frip Set rei.it ws:1 I wer set point to reacter te.el i med cc 5/tv Challanges Sprino 8: 1 i w
Table 1.8 (Cont *d) Moveen in (OhtalNMINI AND Pf flNG M{l( At IOFC, LUMPONltil NAP 11 f441t*f GT MLil ILAIIOff P1tlMARY LOAD 07 Pt*FOM COMrttil0N Gall PIPING SYSItMS 5/RW Discharge Pipinj wetwell Piping Rerouted with strin y r pipe; added 12 new surports Fool $= ell impact / drag loads Spring 80 T-Quencher Dis <.hae p Device lastalled T isuencher device on each 5/Ir# line feitigate water / air clearing loads Spring 84 T-Quencher Support Installed quencher support assembly in 8 bays sig. ort quencher device Spring 90 Quencher Support Bracing Installed quencher support bracing in S bays 06str bute esencter reactions $pring 80 ' Vacutan Breaners installed two. 10-inch wacenas breakers on eacts lie.e Pre =e st =ecessive eeflood in 16-= 5 ering IM Pipe Sa m rts and Restraints lastalled 89 new or undified supports in drywell 5/RV ble=+n.n thrust loads
- 80. til & el forus Attar 3.ed Piping targe Bore Supports Installed 151 new er o04:! led supports forus settons due to LOCA & 5/RV leads Swanee 82 small Bore Supports Instriled 54 new seeports Torus motions dae to LCCA & 5/RV loads Summer 82
). Small Bore Rerouting Rerouted 5 lines, forms motions due to 10CA & 5/RY loads Srring 82 Branch line Supports Installed 25 new or modified supports f orms motions due to LOCA & 5/RV loads Sesmewr 82 forms Pen;trations Reinforced three large bore torus penetrations Pipe reactions from 10CA & 5/RV leads-Spring 82 Valve Operator 5>spports Reinforced 13 tive yotts forees motions dese to LOCA & 5/RV leads WM 1 Pump Anchors Ndif ird ashorage of 4 RHR granps Pipe reactions at sier2Ies Sasumer 82 4 forus Internal Pipier2 l stPCI Turbine fahaust Reree:ted arad ves,$ ported tFCI sparger Pool drag loads Sall 81 J RCIC Tutbine inhaint Reroute.J and resupported PCIC sparger Pool drag leads f all El j - Core Spray, Return lest tine truncated test lies Pool drag leads W ing 82 j RifR Return Test iise Installed reducer, distharge ells, and new supywrts Poel thereal mining Sprines ft2 N Spray Header Reinforced eaisting sqpperts IhermaI icads Spring 82 4 O ~ Rerouted leves anel installed supperts Pool drag leads Spring 80 Vest Drain line } C) Q
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( DRYWELL& REACTOR f E L. 990'-7%" ,- m. L.. --... h A __...L-.. @ p REACTOR VESSEL F c - - DI A. 35'-7" 4 p BIO. SHIELD WALL - l I I l -- DRYWELL b ! i! CL DI A. 65'-0" ' (' STEEL FRAMING i I i II PEDESTAL E 6. 911'-0" [.b. / ( j) j? MAIN VENT - - l G WETWELL w.g ff il \\ WETWELL 4 l l ...J L... -.. / DI A, 28'-9" N x \\ (- x,s i ! s. j 5 L. 57E*-7' '/ ~ ~ [ ~ /'_.. i jd[ g s E L. 359'-9" _l I i .v o I T I I f r, -y I 50' '0%" M-FOUNDATION SLAB FIGURE 1.1 CROSS SECTION - COMPOSITE PLANT IAYOUT l l mnnpa -, 9 (bvJ.4 _~ l
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\\ / / - VENT HEADER - 'Ms \\ ' / SUPPRESSION j,, CHAMDER 4 - - -- DOWNCOMER 4 E L. 880'- 11" 'w-f h PENETRATION [ ~ ~ " g' / ~ PLATFO R M -- -'- -] j [T 'Cy / ~~ f, ? -t-E L. 876'-7%". HWL EL. 875'-2" // p PLAT FORM - - O ~iss-- f SUPPORT ll j / SUPPORT \\ ' COLUMN m N _A CHOR SADDLE hi b f l I + i 1 . ANCHOR BOLT { l i SADDLJ BEARil%G. >{
- 1. VENT HEADER DEFLECTOR
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- 2. VENT HEADER SUPPORT
- 6. T QUENCHER SUPPORT
- 3. SPRAY HEADER
'7. DOWNCOMER TIE
- 4. MONORAIL
- 8. INTERNAL PIPING 7004:1*":.
FIGURE 1.2 CROSS-SECTION..- NETWELL
TORUS COLUMN [, 13*- S 3/16" I RING ~ -4.. p GIRDER 1 i ? %),. Ar i // TORUS l ~ s' GIRDER SHELL I FLANGE / \\ OVERLAY 3 g,, 33. ~ \\ y, -/ ,\\ WELD a f I \\ 3/ WEB REINF. /- [ y! 1%" STIFF. / / s ) g \\\\ / f s A SUPPORT \\ h hl COLUMN i fj y' /, \\ / jf ij' R pN e C j5 s'. k i x/ 3[ A,g'# / '~" i ? i f g /h_ { 'O, n ['JL.Lhi L=.Ll., i .j !g..L iji. kff-E- .M4 m[". A' li Wi e 1 ~ A /A t l-- 7/8" p ANCHOR BOLIS %" SELF LUBR. PL BASE PL FLANGE PL SOLE PL g.' WEB PL - 1%" x 29" x 3*-O" 1 %" x 20" 3%" x 22" x 3'- 6" Sl O O Q b e. 'I FIGURE 1.4 F~L I TORUS SADDIE
a -r w-7 / ts Q,_ 1 WELD REINF. AT COL. TO SHELL CONN. -_/ RING 1 [ GIRDER TORUS SHELL I I ~ P j lI s l E ' '-W-Ill l N 'li i WEB \\ l! I REINF. PLATES y i lp i WEB REINF. PL .I Ji %" INSIDE COLS.
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l! %" OUTSIDE COLS. I M l ll SUPPORT I-COLUMN i Q Fq s W14" x 136" J p } 18 ll t d b '6 COL. R EINF. PL p fi 2-16" x %" lNSIDE COL. f iI .0 Ill II 2-16" x 1" OUTSIDE COL, IP% y rJ1 ww m = b [ $, ~~~~ l l g,,l, Ilg ANCHOR 11 <l i bolts n 1-2u" e aa_, I, ji f j n (
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a MARK I TORUS HYDRODYNAMIC LOADS ISSUE (cont'd) Closure Submitted PUAR to NRC April 29,1982 Received NRC SER on PUA.R January 20, 1984 stating that containment modifications made have restored the original design safety margin to the Mark I containment USAR changes Program history Short-term program description Long-term program description Design criteria e Modification summary Reference to PUAR for more detailed description Technical Specifications-changes Various nonstructural operational changes -(e.g. SRV low-low set, MSIV trip setpoint) e-e
i MARK I TORUS HYDRODYNAMIC LOADS ISSUE (cont'd) Surveillance requirements Technical Specifications Section 4.7.A Visual inspection of suppression chamber interior, including. water line regions, shall be made at each major refueling outage Interior surfaces of the drywell and torus shall be visually inspected each operating cycle for evidence of torus corrosion or leakage SRV operation and torus temperature above 160 deg. and primary coolant system pressure is above 200 psig, an external visual examination of torus shall be done before resuming power j operation CNS Surveillance Procedure 6.3.10.1, "Drywell and p Torus Surfaces and Structural Elements Inspection" fulfills internal inspection requirement l CNS Surveillance Procedure 6.3.10.12 " Torus Exterior Surfaces and Supports Inspection" fulfills l external inspection requirement l l
CONTAINMENT PROTECTIVE COATINGS Drywell - Prime coated with Carbo Zinc 11, finished with Carboline Phenoline 305 Torus - Carbo Zinc 11 Coatings vendor tests show coating systems will withstand temperatures and pressures of steam environment of a DBA LOCA Observed torus coating degradation 10/88 NRC IN 88-82 " Torus Shells with Corrosion and Degraded Coatings in BWR Containments" 12/88 IN 88-82 response - Torus periodic inspection conducted in accordance with CNS Proc.'s 6.3.10.1 and 6.3.10.12. No evidence of a generic problem. Engineering recommended no action at this time. 5/89 IN 88-82 Sup No.1 issued with update on use of underwater inspection techniques
CONTAINMENT PROTECTIVE COATINGS-(cont'd): 1 5/89 S. G. Pinney performed underwater desludging-and coating-inspection. Inspection results: Established 16 grids to-benchmark. coating degrad.ation and corrosion rates Most severe coating defects located near the-bottom of the torus 3/90 S G. Pinney performed underwater inspection of-16 grids established during 1989 outage. Result:- Recommendation to ' delay any-torus -recoat~ until at least 1993
- outage, based on projected coating degradation and: corrosion 59 s
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CONTAINMENT PROTECTIVE COATINGS (cont'd) 1991 outage inspection results Detailed inspection of 4 bays Pitting corrosion exceeded expected growth rate. Anion concentrations and the presence of iron oxidizing / depositing bacteria were noted to be contributors. About 150 pits repaired using underwater cured epoxy. Epoxy was applied where power tool cleaning was utilized or where significant localized corroded areas were encountered. Detailed review of Mark I torus stress analysis was done to ensure compliance with PUAR requirements. Conservative fatigue evaluation completed to verify ' compliance to ASME Section III 1977 as required per PUAR.
) CONTAINMENT PROTECTIVE COATINGS (cont'd) 1993 Outage Plans Desludge all bays to provide suitable work areas, reduce corrosion, reduce dose Inspection of all critical areas of the torus in all 16 bays Complete new calculation mapping Estresses in the torus Map pit sizes / stress areas to verify repairs are not required Provide repair program for all pits. Coating repair methodology similar to that currently utilized in the industry. 9
CONTAINMENT PROTECTIVE COATINGS (cont'd) 4 Future-Activities Continued inspection and coatmg repair program -- e requires-monitoring pit repair areas Possibly revise torus stress analysis using newer. e criteria and revise PUAR Possible torus recoat Continually monitoring and appropriately treating water 4 h 9 c- .a
1. TOPIC 9 ~ INTAKE STRUCTURE GL 89-13 Discussion l Diver Inspections I -~
INTAKE STRUCTURE GENERIC LE' ITER 89-13, " SERVICE WATER SYSTEM PROBLEMS AFFECTING SAFETY-RELATED EQUIPMENT". DOWNSTREAM HEAT EXCHANGERS HAVE BEEN INSPECTED AND DO NOT INDICATE DEGRADATION OF SERVICE WATER PIPE COATING. ESTABLISHED PREVENTATIVE MAINTENANCE ITEM FOR DIVERS TO INSPECT INTAKE BAYS FOR BIOLOGICAL FOULING. DIVERS REMOVE DEBRIS FROM INTAKE SCREEN AS REQUIRED. RUSTED SERVICE WATER PIPING BOLTS CURRENTLY BEING EVALUATED FOR STRUCTURAL INTEGRITY. NO STRUCTURAL PROBLEMS NOTED. I 4 i
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je e TOPIC 10 p_ . SPENT FUEL POOL AND RACKS Original Construction High Density-Rack Mods Cask Drop Accident Scenario. Including Crack Leak Rate-and Make-Up. Capacity .a m1 m ini-n g i i yisu _____,_L
r SPENT FUEL POOL & RACKS Original Construction Pool floor to be flat within 1/16 inch per lineal ft. Accomplished by leveling 6 inch wide flange sections and grouting prior to liner plate. installation. All wall butt joint welds backed by angle sections to act as drainage path. Wall construction coordinated with concrete pours. Walls have 1/2 inch x 51/8 inch long concrete studs-on 2 ft. grid pattern. High Density Rack Mods. DC 77-01 completed February 1979. Storage-capacity increased from 740 spent fuel cells to-2366-spent fuel cells. No anomalies of liner noted during implementation. ~
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P TOPIC 11 HEAVY LOADS AND REFUELING FLOOR ACTIVITIES Procedure-7.0.3 L 1 i Procedure 10.26 Refueling. Floor Crane Operation- ,r,- r,,..,, e, e- ,~,-,,nn,.. ~. ,.,-c. ++ - - - - <, ~. -+ -,+- - - -
1 1 Cask Drop Accident Analysis Pool floor evaluated for falling shipping cask. Mass used was 135,000 lbs. Free fall from height 5 ft. above water. Impact penetration less than half concrete slab depth. Remaining concrete section evaluated for crack size. o Leak rate postulated at 15.6 gallons /sec. Make-up capacity 16.7 gallons /sec. from volume of 450,000 gallons. l l l-4 ..--,.-.-__..._._....,-[
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o o o Heavy Loads & Refueling Floor Activities. 1 Required By NUREG 0612. And Administrative. Procedure:7.0.3,. Control Of Heavy Load Lifts. Procedure 10.26; Working Over. Or In Reactor Vessel Or Fueli ~ PoolLRequirements. 1 3 '[ a Refueling Floor 1 Building.. Crane Operation. Two Documented ' Cases df Dropped-Items In The. Fuel Pool. [ c 1
A Control Of Heavy Loads g Implernentation Required By. NUREG 0612, Control Of Heavy Loads At Nuclear Power Plants. Maintenance Procedure 7.0.3, Control Of IIcavy Load Lifts. All work activities are controlled by Maintenance Work Requests (MWRs). During the development of MWRs, Maintenance Planning brings to the attention of the System Engineer maintenance work activities that may be NUREG 0612 lifts. The System Engineer identifies if the maintenance work activities will require an NUREG 0612 lift and implement the necessary controls. Procedure is applicable to lifting loads of greater than 1000 lbs over or near irradiated fuel, or nuclear safety related equipment. h Procedures that develop the plant modification work packages flags the Design Engineer to incorporate NUREG 0612 controls in the installation procedme when applicable. Procedure is not applicable when existing procedures implement NUREG 0612 controls (refueling floor procedures). Refueling floor procedures which implement NUREG 0612 requirements (i.e., identify load paths). MP 7.4.1 / 7.4.1.1, Shield Plug Removal / Installation. MP 7.4.2 / 7.4.2.1, Drywell Head Removal / Installation. MP 7.4.4 / 7.4.4.1, Reactor Pressure Vessel Head Removal / Installation. MP 7.4.5 / 7.4.5.1, Reactor Vessel Steam Dryer Removal / Installation. MP 7.4.6 / 7.4.6.1, Reactor Vessel Steam Separator and Fuel Pool h Gate Removal / Installation.
O control or aerueling Floor Activities Implementation Required By. Administration; to avoid imposing the requirements of Tech' Spec 3/4.10, Core Alterations. Procedure 10.26, Working Over Or In Reactor Vessel Or Fuel Pool Requirements. Administratively implements controls during any activities over or near the reactor cavity or fuel pool. Logs and requires applicable items to be tied off in the controlled area.
- All load movements in the controlled area that are not identified in existing SORC approved documents are evaluated and require the approval of the Operations Manager and Plant Manager. s O
Refueling Floor Building Crane Operation. Crane is operated in the RESTRICTED mode while handling fuel casks. RESTRICTED mode only allows the crane-to be operated in a predetermined path. Dropped Items In The Fuel Pool. CRD guide tube was being lowered into storage in the fuel pool and free fell the last two (2) feet in October 1984. Fuel pool storage rack seismic support brace was dropped on the cask pad during reinstallation preceding spent _ fuel shipping activities in November of 1985. No visible damage was found during the inspection of the cask loading pad and seismic support brace.
i O TOPIC 12 SAFETY RELATED STORAGE TANKS Design Criteria g-List of Safety Related Tanks Anchorage Descriptions Settlement Issues Included in A-46 Scope O
e O SAFETY RELA'dD STORAGE TANKS Design Criteria No safety-related tanks outside except Diesel Oil Tanks (buried), settlement not an issue Seismic design was to FRS ZPA Flat bottom tanks designed using TID-7024 methodology DG Oil Tanks missile protected by concrete bunker e O Safety-Related Tanks DG Day Tanks 1 and 2 DG Storage Tanks A and B DG Starting Air Receivers (4 total) RHR SWBP Gland Water Tanks A & B SWP Gland Water A & B Nonsafety-Related Tanks Designed for Class IS Event ECST's 1 & 2 REC Surge Tank SLC Storage Tank O Various tanks in Radwaste Basement
SAFETY RELATED STORAGE TANKS (cont'd) Anchorage Flat bottom tanks anchored with cast-in-place anchors e Elevated dish bottom tanks anchored with expansion e anchors Tanks will be reevaluated in A-46 effort currently planned for 1994 Refueling Outage Preventative Maintenance DG Storage Tanks cleaned and internally inspected e for corrosion Past Tank Problems and Leakage DG Day Tank had 3/4" pipe failure in 9/77 due to e cycle fatigue. Piping for both tanks redesigned and replaced. O
? i O i i i L IOPIC 13 1 IANK INSPECTION AND MAINTENANCE h O Preventive Maintenance ? Past Tank ProblendLeakage I Future Actions t l + o LO - i v,,, v. ,-,,--na,. ,, ~......, -,. -,....,. -.. ... ~.. -, - -......,. ~., -,,. --....,,.Lai.. ,.. - + -.
O TOPIC 14 BURIED PIPING Design Criteria Contract 69-4 Specifications Cathodic Protection O
O BURIED PIPING Design Criteria Underground pipe to meet USAS B31.1.0 - 1967 design requirements for pressure thermal and seismic anchor movements. Underground piping to resist floatation effects of design flood level of 903 ft. - 6 in, by minimum backfill depth requirements. Class I piping to be placed within Class I O structural fill zones beneath building mats or within 30 ft. zone beside buildings. Installation Specifications Trench bottoms graded to provide uniform support for pipe on undisturbed soil or compacted backfill or lean concrete. Backfill over and around pipe compacted to 85% relative density. Minimum thickness specified to be placed prior to use of heavy equipment over pipe. " ~ "
O Pipe Coatings and Coverings Exterior surfaces: 1 coat Bitumastic 50 primer 2 coats hot bitumastic 70-B enamel to thickness of 3/32" 1 layer asbestos coal-tar saturated felt In addition to the above coatings, Class I piping has a layer of vinyl wrapping. O Interior surfaces of all underground pipe 4 inches and larger except AR-1, A.R-2 and PV-1 have: 1 coat Bitumastic 50 primer 2 coats hot bitumastic 70-B enamel to thickness of 3/32" Coatings applied in accordance with AWWA specification C203. Some pipes exiting Reactor South are encased in concrete for additional protection. O lt!Nwol
O Cathodic Protection Safety-related underground pipe and tanks O protected from corrosion by rectifiers and anode beds of Cathodic Protection System. Surveys conducted in 1982 and 1985. r Discuss survey results and recommendations.- O
-a I 6 ( e. 9 TOPIC 15 ( o CONTAINMENT TESTING O[ O
CONTAINMENT TESTING fO' PRIMARY AND SECONDARY CONTAINMENT TESTING REQUIRED IlY 10CFR50, APPENDIX J. REQUIREMENTS INCORPORATED 1NTO CNS TECIINICAL SPECIFICATIONS, SURVEILLANCE REQUIREMENTS, SECTION 4.7. INTEGRATED LEAK RATE TESTING PERFORh!ED TO VERIFY PRIMARY CONTAINMENT. PRIMARY CONTAINMENT IS CONFIRMED IF LEAKAGE RATE DOES NOT EXCEED EQUIVALENT OF 0.635 PERCENT OF PRIMARY CONTAINMENT VOLUME PER 24 IIOURS (vtG/24 hrs) AT 58 PSIG (Pa). TECIINICAL SPECIFICATION 4.7. A.2.a. TIIE MEASURED INTEGRATED LEAK RATE (Lam) SIIALL IlE LESS TIIAN 0.75 La (0.75 x 0.635 wt%/24 hrs). TECIINICAL SPECIFICATION 4.7. A.2.d. 5. EXEMPTIONS TO 10CFR50, APPENDIX J: LOCAL LEAK RATE TESTS FERFORMED ON MSIV'S AT 29 PSIG (Pt) RATHER 58 PSIG (TECHNICAL SPECIFICATION 4.7.A.2.f.3). MAIN STEAM LINE AND FEEDWATER LINE EXPANSION BELLOWS TESTED AT 5 PSIG (TECHNICAL SPECIFICATION 4.7. A.2.f.4). PERSONNEL AIRLOCK TESTING AT 3 PSIG RATHER THAN Pa WHEN NOT OPENED BETWEEN REFUELING OUTAGES OR WITHIN THREE DAYS OF OPENING DURING TIMES WHEN CONTAINMENT INTEGRITY IS REQUIRED (TECHNICAL SPECIFICATION 4.7. A.2.f.5). SURVEILLANCE PROCEDURES: S.P. 6.3.1.1, " PRIMARY CONTAINMENT LOCAL LEAK RATE TESTS". u. l
S.P. 6.3.1.3, "PRlhiARY CONTAINh1ENT INTEGRATED LEAR RATE TEST". S.P. 6.3.1.8, " ELECTRICAL PENETRATIONS LEAK CIIECK". S.P. 6.3.10.1, "DRYWELL AND TORUS SURFACE AND STRUCTURAL ELEhiENTS INSPECTION". S.P. 6.3.10.8, " SECONDARY CONTAINhiENT LEAK TEST". S.P. 6.3.10.12, " TORUS EXTERIOR SURFACES AND SUPPORTS INSPECTION". S.P. 6.3.10.17. " SECONDARY CONTAINh1ENT PENETRATION INSPECTION". INTEGRATED LEAK RATE TEST RESULTS. ~ Q {, a9 00 3 1 a at ^ I D$ p ~ -:1; } 4 N .g e, hg ---wee-.- 4
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t O F 3 ) i TOPIC 16 1 TESTING OF CONTAINMENT BELLOWS O Required by 10CFR50 Appendix J Testing and Monitoring i 1 O i
1 Testing Of Containment Bellows Three Containment Bellows. Required By 10CFR50 Appendix J. 1 Testing And Monitoring. No Past Problems Or Leakage. 4 O O
O I"Si*ctiat2"d 7"82"x r C""ta!""'""' """""" Containment Bellows. Reactor vessel to drywell vessel bellows. Drywell v:ssel to Secondary Containment bellows. Vent header from the torus to drywell bellows, eight (8) total, j Implementation Requis ed Hy. Pressure integrity as required by 10CFR50 App. J. j IE Bulletin 84-03, Refueling Cavity Water Seal. Reactor and drywell vessel bellows provide the sealing functions during reactor well floodup to support refueling activities, and are not subject to containment 4 pressures. These bellows are not required to be tested, however, they are required to be monitored for seal leakage. O '1 Testing And Monitoring. Vent header bellows arc pressure tested every three (3) years during the primary containment integrated leak rate test. . Reactor vessel bellows iias a leak detection system =that alarms in the Control Room with a leakage of 5 gpm and increasing. The flow switch is inspected and calibrated prior to refueling activities,_ and_ has a measured range of 0 to 20 pm. 3 Drywell vessel bellows has a leak detection system identical to that of the reactor vessel bellows leak detection system. Past Problems Or Leakage. No problems or documented leakage to date. O
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O-TOPIC 17 S_UPPORT ANCHORAGES Anchor Types O Design Criteria = i IE Bulletin 79-02 Piping and Pipe Support Qualification Project l IE Information Notice 80 '21 O I
O SUPPORT ANCHORAGES ANCHORS USED DURING ORIGINAL CONSTRUCTION Cast-In-Place "J" bolts or threaded rods used for-major l equipment; i.e., pumps, motors, air handling equipment, etc. A grid system of cast-in-place Richmond inserts is provided on the underside of Reactor Building floors and in walls for pipe supports. l Phillips " Read Head" self drilling shell anchors rangmg i from 1/4" to 7/8" diameter were used extensively in. addition to the cast-in-place inserts. Each self-drilling anchor is installed with epoxy under rigidly controlled procedure. ANCHORS CURRENTLY BEING USED Hilti "HDI" and K.wikbolt II are used predominately. 1 Drillco undercut anchors are used for large loads.- O i W1%s04 ' t ...,.a,-,.,;,-....n,-.._. ,-,...-....-..,,.-,.,.,.,-,.n.. .. _, -.~...,.,-,u,,. -,,,,- ..;:w-
O DESIGN CRITElllA Manufacturer's published test data and recommended o capacities are being used for design of anchors. Original safety factors for Red Head self-drilling anchors were 6 for tension loads and 5 for shear loads (OBE). Currently using a factor of safety of 5 for shell anchors and 4 for wedge anchors to be consistent with the requirements of IE Bulletin 79-02 (SSE). O Concrete strength of 5000 psi is used to determine anchor capacity. Prying action is accounted for by considering base plate flexibility. Finite element analysis is performed if necessary. 4/3 elliptical shear-tension interaction equation used for analyzing anchors. l O l - ~, b
4 O IE BULLETIN 79-02 Account for base plate flexibility and assure a minimum factor of safety of 4 for wedge anchors and 5 for shell anchors. Resolution: 1144 supports were reviewed and 169 required modification. O es ribe design requirements f r cy i 1 ads. Resolution: Bolt torquing loads were all above the maximum design loads including thermal and seismic loadings; therefore, state of stress of the bolts does not change. Use of epoxy prevented plug from working loose. l
i 4 IE BULLETIN 79-02_ (Continued) O Evaluation of small bore pipe supports (2" diameter and under). TORQUE INSPECTIQN_RESULTS No. Supports No. Anchor No. Anchor lasaccled Holts _Insnecteld D.oksJ)_efectixe 60 137 2 CONCLUSION: A DEFECT RATE OF LESS THAN 5% IS FREDICTED WITH A 95 % CONFIDENCE LEVEL FOR SMALL LINE ANCHOR BOLTS. IIANGER EVALUATION RESULTS FOR-g ANCHOR BOLT F.S. No. Supports No. Supports Enluated 5 5 SSE 60 60 CONCLUSION: DUE TO THE SM ALL LOADS, ANCHOR BOllr FACTORS OF SAFETY ARE GENERALLY VERY HIGH. FACTOR OF SAFETY OF FIVE PER IE BULLETIN 79-02 WITH SSE LOADINGS EITHER MET OR EXCEEDED. O kU5%uG4 4
O PIPING AND PIPE SUPPORT QUALIFICATION PROJEC'1 Complete reanalysis of all large bore class IS piping and pipe supports with the exception of Torus Attached Piping and SRV Discharge Piping. Included approximately 1500 pipe support analyses and 93 piping stress problems. Complete updating of pipe support drawings. Resulted in approximately 850 support modifications. The work started in 1988 and is scheduled to finish during the 1993 Refueling Outage when the remaining 80 to 90 modi 0 cations are implemented. O NPPD NED Engineers reviewed all calculations performed by EAS Energy Services. Calculations can be maintained by NPPD on an in-house IBM RISC 6000 workstation using ADLPIPE software. Complicated pipe supports can be analyzed using Images 3D STRUDL software on in-house PCs. Mr. Roger Reedy (Chairman of ASME Section III) of Reedy Associates performed an independent third party review of the whole project. In Reedy Associates' opinion, "the criteria, methods employed, and resulting designs are quite good and certainly go beyond the original requirements for the design and documentation of the plant." ,o A design criteria document has been completed and is being d used to ensure that the plant design btsis remains current.
O IE INFORMATION NOTICE 80-21 DeGeient anchorage of safety-related electrical equipment. Station Service Transformers DC to AC Inverter Battery Chargers Batteries DG Switchgear Cabinets Motor Control Center Cable Trays Control Room Panels Instrument Racks Recently Replaced =
- = Deficiencies noted at CNS and Corrected In addition to the past seismic reviews, most of the above items will be reviewed again under USI A-46 and IPEEE.
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O 1 TOPIC 18 SEISMIC INSTRUMENTATION O Equipment Description Maintenance and Testing f O
SEISMIC INSTREMENTATION KINEMETRICS STRONG MOTION ACCELEROGRAPII MODEL SMA-3 ENGDAHL PEAK ACCELEROGRAPHS MODEL PAR 400-3 RELOCATION OF SEISMIC INSTRUMENTATION MAINTENANCE OF SEISMIC INSTRUMENTATION 3
SEISMIC INSTRUMENTATION i O KINEMETRICS STRONG MOTION ACCELEROGRAPIIS (MODEL SMA-
- 3) INSTALLED AT CNS.
TilREE SENSING LOCATIONS: REACTOR BUILDING, NORTHWEST QUAD, ELEVATION 859'. REACTOR BUILDING, ELEVATION 1001', NORTHWEST CORMER. NORTH
- YARD, 140' NORTH AND 80' EAST OF DIESEL GENERATOR BUILDING NORTHEAST CORNER.
TIIREE PEAK ACCELEROGRAPIIS (ENGDAIIL MODEL PAR 400-3) INSTALLED AT CNS. REACTOR VESSEL SUPPORT SKIRT AT ELEVATION 915'. ON CORE SPRAY LINE 1 A ENTERING DRYWELI AT ELEVATION 946'. g SEISMIC CATEGORY I ISOLATION RELAY PANEL "C" IN CABLE SPREADING ROOM ELEVATION 932'. o RELOCATED SEISMIC INSTRUMENTS IN NORTII YARD DUE TO-CONSTRUCTION OF MULTI-PURPOSE FACILITY AND SPURIOUS ACTUATIONS DURING OPERATION OF DIESEL DRIVEN FIRE PUMP. -PEAK ACCELEROGRAPIIS RELOCATED DUE TO RECORDING ERRONEOUS DATA FROM- 'WATERIIAMMER AND NORMAL OPERATIONAL DISTURilANCES. MAINTENANCE OF SEISMIC INSTRUMENTS. PM 02010, FUNCTIONAL TEST OF SEISMIC MONITORING SYSTEM PER I&C PROCEDURE 14.3.9, " SEISMIC MONITORINO SYSTEM TESTING". PERFORMED MONTHLY. PM 02011, FUNCTIONAL TEST OF SEISMIC MONITORING SYSTEM WITH -KINEMETRICS REPRESENTATIVE. PERFORMED _ SEMI-ANNUALLY.
.I PM 02012, REPLACE PEAK RECORDING ACCEL INSPECT AND
- CLEAN, MI-PAR-1, PERFORMED DURING REFUELING OUTAGES, I&C PROCEDURE 14.3.2, "DRYWELL SEISMIC ACCELEROMETER PLATE REPLACEMENT AND DATA REDUCTION".
PM 02013, REPLACE PEAK RECORDING ACCELEROMETER PLATE, INSPECT AND MI-PAR--2, MI PAR-3, PERFORMED ANNUALLY,I&C PROLiDURE 14.3.1," SEISMIC ACCELEROMETERS PLATE REPL : CEMENT AND DATA REDUCTION". SURVEILLANCE PROCEDURE 6.4.8.8, " SEISMIC SYSTEM CALIBRATION CHECK" PERFORMED QUARTERLY. I&C PROCEDURE 14.3.6, " PASSIVE SEISMIC ACCELEROMETER CALIBRATION CIIECK", I&C PROCEDURE 14.3.9, " SEISMIC MONITORING SYSTEM TESTING". NO PLANS TO UPGRADE. O
t e O i TOPIC _19 T PLANT SAFETY PROCEDURES Q EP 5.1.1 Earthqua,, j EP 5.1.2 Tornado Watch = EP 5.1.3 Flood EP 5.1.7 Emergency Classification c n%)
PLANT SAFETY PROC _EDLRES 'i Emergency Procedure 5.1.1, " EARTHQUAKE" Emergency Procedure 5.1.2, " OPERATION DUIIING TORNADO WATCH" Emergency Procedure 5.1.3, " FLOOD" Emergency Procedure 5.1.7, " EMERGENCY CLASSIFICATION" e
PLANT SAFETY PROCEDURES EM1'RGENCY PROCEDURE 5.1.1, "EARTIIQUA'rE". SEISMIC EVENT AT d 0.01 g. EMERGENCY SEISMIC HIGH LEVEL AT h 0.1 g. REFER TO PROCEDURE 5.1.7, " EMERGENCY CLASSIFICATION". EMERGENCY PROCEDURE 5.1.2, "OPEr .)N DURING TORNADO WATCH". COMMERCIAL REPORTS OF SEVERE WEATHER OR TORNADO WATCHES. LOCAL VISUAL SIGHTINGS. REFER TO PROCEDURE 5.1.7, " EMERGENCY CLASSIFICATION" TO DETERMINE IF EMERGENCY DECLARATION IS APPROPRIATE. EMERGENCY PROCEDURE 5.1.3, " FLOOD". 895' MSL (MEAN SEA LEVEL) MONITOR AND RECORD RIVER LEVEL HOURLY AT INTAKE STRUCTURE. 897' MSL ESTABLISH PRIMARY BARRICADES. REFER TO PROCEDURE 5.1.7, " EMERGENCY CLASSIFICATION". 902' MSL INITIATE NORMAL SHUTDOWN, VENT AND FLOOD REACTOR VESSEL. EMERUrbcY PROCEDURE 5.1.7, " EMERGENCY CLASSIFICATION". NOTIF! CATION OF UNUSUAL EVENT GROUND MOTION > C01 g. RIER LEVEL > 889' OR < 867'. TORNADO TOUCE WITHIN OWNER CONTROLLED AREA. SUSTAINED WIND SPEED > 74 MPH. ALEl" 10UND MOTION > 0.1 g. R 7ER LEVEL > 902' OR < 865'. TORNADO TOUCHES DOWN WITHIN PROTECTED AREA. ? SUSTAINED WIND SPEED > 95 MPH.
O TOPIC 20 USI A-46 ACTIVITIES O Brief History of the Issue NPPD Future Plans for Resolution l 1 s
O A-46 ACTIVITIES 1 i Brief History 9 NRC USI-A46 - Seismic capability of equipment in operating plants must be reassessed. Evolving design criteria and methods in seismic qualification raised questions regarding margin of safety in operating plants. Seismic Qualification Utilities Group formed to address A-46 issues. 'O Developed Generic Implementation Procedure (GIP) Safe Shutdown Equipment List Plant walkdown Relay evaluation Anchorage evaluation Outlier evaluation guidance Documentation L Seismic verification methods rely heavily upon L earthquake experience and generic testing L May 22, 1992 NRC Generic Letter 87-02 Q Supplement No.1 issued-
I A-46 ACTIVITIES (cont'd) O NPPD Plans Submit GL response by September 19, 1992 committing to use GIP-2 as supplemented by the SSER No. 2 with no exceptions anticipated Safe Shutdown Equipment List work has been started Perform plant walkdown during Fall,1994 refueling outage Utilize 2 walkdown teams Each team consists of 1 NPPD engineer, 1 expert consultant,1 technician O
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A 4- -.m .L.,d.u--- 4-m -4.1..-4+ - - 3 4 .J O 4 ? TOPIC 21 O MASONRY WALLS l O
MASONRY WALLS
- Masonry block is not used for structural purposes in Class I structures at CNS.
- Response to IE Bulletin No. 80-11 indicated that no re-evaluation was necessary.
- Response to Information Notice 87-67 indicated that no new actions were required.
- No Masonry block wall additions have been made within Class I structures.
- Masonry block is used for removable wall sections in seven areas of the Reactor Building - these are addressed in FSAR Questions 12.3 and 12.45.
Six areas accessible during walkdown. l O 19.2d kil
+. l(}~ TOPIC 2_2 DEGRADATION OF STEEL CONTAINMENT O Background -IN 86-99 & GL 87-05 Comparison of Oyster Creek and CNS CNS Surveillances/ Inspections O
O " DEGRADATION OF STEEL CONTAINMENTS" INFORMATION NOTICE 86-99 GENERIC LETTER 87-05 " REQUEST FOR ADDITIONAL INFORMATION - ASSESSMENT OF MEASURES TO MITIGATE AND/OR IDENTIFY POTENTIAL DEGRADATION OF MARK I $ DRYWELLS" INFORMATION NOTICE 86-99, SUPPLEMENT 1 DATED FEBRUARY 14, 1991 0
4 + %.) v l \\ l 'n \\ \\ [ FLOOD-UP ey' BELLOWS v [ 4 i i o / / / \\ \\ \\ l I 'i .J I i i cygge SpuSH ^ N j_ ~,cg;....., / ~ ,/
COMPARISON OF OYSTER CREEK AND CNS IIEh1 OYSTER CREEK CHS NSSS/AE G.E./B&R G.E./B&R Year Operational 1969 1974 Drywell Liner 1.115" 1.50" Thickness at Sand Drain Gap Material Firebar D Polyurethane Foam (Chlorides Leached From Removed During g Gap Material Creating Construction Corrosive Conditions) Fuel Pool Liner 1/8" SS 1/4" SS Sand Drain Details Gap Not Scaled, No Top Gap Not Sealed, No Top
- Drain, Drain, 4-4" Bottom Drains 8-4" Bottom Drains Drywell To Reactor Cover Plate w/ Gasket and Seal Welded Construction Well Seal 2" Drain With 4" and 2" Drains O
g-- -s g k_/ (j N I"O WEEP HOLES, s li 0C. IfJ CLOSURE R.- I %" CL. (2 ETHAFOAM SHEETS) 18Ga CLOSURE R. _ EL.888'-0" .s THl0KOL JOINT '.h l-e g]/4-) SEALER ^ 9 O [' "Y.,.. / QY EL. 884'- 10" (-lI - .? t' u-
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1 F Y r 1 DRYWELL3 r r ere NEACRE WELL DRAIN g 8 f-8 o o '"Y 4 [ 2"9 REACTOR WELLS P g SEAL DRAIN J7 m=m 2"0 DftAIN / d 4"8 SEAL RUPTURE \\ DRAIN NSULATION MATERIAL \\ SECTION A-A p ,,, g REACTOR WELL SEAL DRAIN o l JRYWELL r ', i I I I O lO' 1 l i 'l Lb -4 N x 4. x 2 4 b COOPER 1 cut our o NUCLEAR l STATION 3/4,, x 7,, x 2 7,, COVER PLATE SECTION B-B OYSTER CREEK NUCLEAR STATION O FLOOD-UP BELLOWS
O .w/ CNS SURVEILLANCES/ INSPECTIONS UT measurements were taken of drywell liner just above the sand cushion. All measurements were within the tolerances of new plate. Any leakage past the reactor well seal and/or reactor well liner is indicated by flow indicating switch FPC-FIS-64. Switch is observed by operators when the reactor well is flooded. It is periodically removed and calibrated. OV Sand drains are inspected daily when reactor well is flooded. Drywell shell hatch PC-PENT-SIP 5, (El. 962') was opened and the liner was inspected for evidence of corrosion. The inspection revealed no moisture or any abnormal corrosion of the drywell. An STP is being written to provide assurance that the sand drains are not blocked. A vacuum test is planned for implementation during the 1993 Refueling Outage. A mockup of a sand drain was constructed in early 1992 and a vacuum test was performed to confirm that.the proposed equipment and methods would prove drain operability.
O TOPIC 23 10 CFR 50.59 EVALUATIONS g Short Summary of Each of the 10 DCs Chosen by the NRC \\ O
e O 10_CER.10.59 EVALUATIONS DC 84-26 Multi-Purpose Facility Built to support pipe replacement outage e Founded on piles to bedrock for isolation from other structures 4" separation between MPF and nearby structures ~ DC 86-60-1 Modification of Control Room HVAC C Rivet CR Ventilation ductwork to duct supports Same Safety Evaluation as original DC e DC 86-129 Reactor Building Conduit Hanger Fix Degraded conduit hanger modification Condition discovered when hanger was analyzed for addition of conduit O
10 CFR 50.59 EVALUATIONS (cont'd) O DC 87-92 SSFI Structural Modifications Removed Unistrut member between instrument rack and SW Pump discharge pressure gage diaphram seals because of spacial interaction concern Added support for drain pipe in Battery Rooms Added support to light fixtures in Battery Rooms Added support to REC piping in Battery Room and ECST area e Modified support for FCU in Control Building Basement DC 87-197 Electrical Conduit Hangers Modified conduit hangers found to be qualitative outliers during a walkdown using acceptance criteria developed by SQUG for USI-A46 resolution l O
I O 10 CFR 50.59 EVALUATIONS (cont'd)L DC 88-24 Modification-of RWD and SW Piping Supports Modified existing and added new pipe supports to Service Water and River Well Discharge piping DC 88-53A Missile Barrier Structure for Control Building Essential Ventilation l Built reinforced concrete tc,rnado generated missil-barrier to protect new CB ventilation system and O Crit. SWGR. Room Moved interferences from the new structure location-DC 88-302B Pipe Support Modifications Final Design Change for pipe support modifications in the scope of the program previeusly discussed .O i
10 CFR 50.59 EVALUATIONS (cont'd) O DC 90-342 RCA Access Facility e Provided for single access to RCA Cut new door opening through Radwaste Building reinforced concrete wall Moved conduit and tubing interferences from the new doorway location DC 92-14 SW-MOT-SWPA, B, C, D Lower Endshield Support Segments This DC currently being implemented Addition of endshield supports to enhance endshield load carrying ability 9
O TOPIC 24 IPEEE ACTIVITIES O Scope NPPD Response to GL 88-20, Supp. No. 4 Current Status O
IPEEE ACTIVITIES O Scope (GL 88-20 Supplement No.4) Seismic Events Seismic PRA Seismic Margins Approach Internal Fires Fire PRA FIVE Methodogy 9 Other events High winds and tornadoes External Floods Transportation and nearby facility accidents O
NQ 4 O 1eens AcTivi'11es (cont c) NPPD Response to GL 88-20 Sup. No. 4 Seismic Margins Approach will be coordinated with A-46 activities FIVE methodology will be applied toward internal fire evaluation Other events will be evaluated utilizing the progressive screening approach outlined in NUREG 1407 O IPEEE results to be submitted by June 28, 1994 contingent upon issuance of final SQUG SSER and all GIP open issues are resolved before August 1992. O
IPEEE ACTIVITIES (cont'd) e Current Status Seismic Margins not yet started - to be coordinated with A-46 utilizing the same walkdown teams Fire vulnerabilities being evaluated using FIVE methodology EPM contracted to do evalmtion (App. R consultant) Work has been commencing with emphasis on participation from engineering, PRA group, and $ Operations Support
- Group, to develop an appreciation of severe accident behavior NPPD staff currently investigating into other events original design basis and 1975 SRP requirements O
O . TOPIC 25 LER CORRECTIVE ACTIONS O LER 91-20 LER 86-28 O
1. SEVEN LERs ISSUED FROM 1985 TO PRESENT REG ARDING STRUCTURAL ISSUES g LER NO. REPORT DATE TITLE
- 91-020 12/10/91 Failure of the Primary Containment Integrated Leak Rate Test due to Drywell Vent Monitor System and Containment Leakage 91-007 07/30/91 Fan Coil Unit for Diesel Generators Not Seismically Qualified due to Inadequacies in Original Design 88-016 05/17/88 Pipe support design deficiencies discovered during design change engineering activities g
86-033 12/16/86 RHR and Core Spray Pump Deficiencies Believed to be Caused by High Cycle Fatigue Loading which were Discovered During Inspection
- 86-028 10/21/86 Standby Gas Treatment System Seismic Deficiencies 86-027 10/15/86 Safety-Related Instrumentation Seismic Deficiencies 85-005 08/18/85 Excessive Primary Containment Local Leakage Rate O
Denotes LERs selected for closer examination during audit. ) DNuwi
.i i' t n r O d 6-LER 91-20 Failure of Primary Containment Integrated Leak Rate Test due to-l Drywell Vent Monitor System and Containment-Penetration, a Leakage. W DESCRIPTION OF EVENT O ILRT Failure was on December 16, 1991. Due to failure of the Drywell Vent Radiation Monitor r and;- Leaking Containment Isolation-Valve in RWCU... i.: Leaking Containment Isolation Valve -in ' Reacto'r Feedwater Sy_ stem..
- O DNMAu0t
g LER 91-20 (Continued) Failure of Primary Containment Integrated Le c Rate Test due to Drywell Vent Monitor System and Containment Penetration Leakage. CAUSE OF FAILURE O Drywell Vent Radiation Monitor attributed to design. Threaded connections became degraded through numerous assemblies and disassemblies. RWCU Isolation Valve Leakage case was not determined. Reactor Feedwater (RF) System leakage was due to normal wear of soft seats on RF check valves. O
o) \\ LER 91-20 (Continued) Failure of Primary Containment Integrated Leak Rate Test due to Drywell Vent Monitor System and Containment Penetration Leakage. CORRECTIVE ACTIONS /9 Detector bolt holes retapped and bolted in place. V Evaluation of Drywell Vent Radiation Monitor design. New soft seats were installed on RF check valves. RWCU valves replaced. Successful completion of LLRTs. O v DNMM1
l LER 86-028 Standby Gas Treatment System Seismic Deficiencies. DESCRIPTION OF EVENT Problem discovered October 21,1986, during a review of SBGT against requirements of RG 1.52 and ANSI N509, by an independent engineering firm. Review of complete SGBT System was an outgrowth previous discrepancies noted by LER 85-001. Study identified 58 of 88 seismic class "IS" supports on SBGT suction ductwork did not meet class "IS" criteria. (Determined to be structurally inadequate during DBE). SBGT immediately declared inoperable. Plant was in cold shutdown for a scheduled refueling outage. O
O LER 86-028 (Continued) Standby Gas Treatment System Seismic Deficiencies. CFSE OF FAILURE Original design documentation was located and it indicated design was adequate. O As-built configuration did not match original design documentation. Portion of study completed October 21,1986. 58 our of 88 ductwork supports did not meet class "IS" criteria. Determined to be undocumented design changes made by contractor during original construction of system. O
O IER 86-028 (Continued) Standby Gas Treatment System Seismic Deficiencies. CORRECTIVE ACTIONS Performed prior to startup from 1986 refueling outage. a Implement modifications to upgrade deficient SBGT g suction supports. Analyze SBGT discharge ductwork. Review Control Room Class "IS" seismic ductwork installed by the contractor in question. Deficiencies were attributed to a single contractor; event determined to not have generic implications. l l O
O TOPIC 26 INSPECTION REPORT CORRECTIVE ACTIONS O SSFI 50-298/87-10 + C a ~ O
O PURPOSE OF PRESENTATION To discuss past disposition of structural issues identified in previous NRC Staff inspection reports. Majority of all NRC identified structural issues are addressed in Inspection Report No. 50-298/87-10, " Safety System Function Inspection (SSFI)." Discussion limited to SSFI identified concerns. O
GV NFC INSPECTION REPORT NO. 87-10 SSFI P_IlEPOSE OF INSPECTION Inspection performed over period of May -11 to June 19, Q 1987. Inspection was to assess operational readiness and functionality of emergency electrical system and auxiliary support systems. Attention directed towards details of modifications and design controls, maintenance, operations, and testing applicable to the systems in question. O , ~..
O NRC INSPECTION REPORT NO. 87-10 SSFl IDENTIFIED STRUCTURAL ISSUES (ITEMS 3.1.'.2 (a-e9 (a) Batn q 9.ws A and B - Lighting fixtures, conduit, drainpipe, and 3 inch line located in close proximity to batteries. (b) Emergency Diesel Generators -I&C tubing supported by air operators. g (c) SVV Motor Operated btrainers - 3" drain line with manual valve inadequately supported by rope. (d) Cooling Unit for Control Room Basement - Inadequate support when effects of empty bolt holes were considered. (e) Service Water Root Valves - Four valves located immediately downstream of SW pumps inadequately supported to gage standards.
NRC INSPECTION REPORT NO. 87-10. SSFI ISSUE SSFI Item No. 3.1.2 (5) (a) - Battery Rooms A & B Lighting fixtures, conduit, drainpipe, and 3 inch line located in close proximity to batteries. C_OJ1RECTIVE ACTIONS Calculation NEDC 87-073 performed to show lighting O fixtures would be adequately supported with the addition of new steel pipe clamps. The clamps were installed per MWR 87-2025. Calculation NEDC 87-071 performed to show 4 inch cast iroa sewer pipes, with additional supports, would not separate and fall on the emergency batteries. These supports were installed per MWR 87-2025. Calculation NEDC 87-076 was performed to show that the 3 inch REC would be adequately supported with the addition of two new supports. Supports were installed per MWR 87-2026. O DNM% sol -
O NRC INSPECTION REPORT NO. 87-10. SSFI ISSUE SSFI Item No. 3.1.2 (5) (b) - Emergency Diesel Generators I&C tubing supported by air operators. O COBRECTIVE ACTIONS Calculation NEDC 87-092 performed showing existing e trays are adequate with addition of several supports. Supports installed under Temporary Design Change TDC 87-013. Lines rerouted during 1988/1989 Outages. O
0 0 NRC INSPECTION _ REPORT NO. 87-10 SSFI iSELE SSFI Item No. 3.1.2 (5) (c) - Service Water Motor Operated Strainers O 3" drain iine with manuai vaive inadequateiy supported by rope. CDRRECTIVE ACTIONS The subject piping, associated with the gland water strainers, was no longer required and removed per DC 86-105. O
O NRC INSPECTION REPORT NO. 87-10. SSFI ISSUE SSFI Item No. 3.1.2 (5) (d) - Cooling Unit for Control Room Basement Inadequate support when effects of empty bolt holes were considered. g CORRECTIVE ACTIONS Calculation NEDC performed showing existing anchors will be adequate with addition of two new anchors. l Above described modification installed under l MWR 87-2140. O IWWeset
O NRC INSPECTION REPORT NO. 87-10. SSFI ISSUE SSFI Item No. 3.1.2 (5) (e) - Service Water Root Valves Three valves located immediately downstream of SW O P"mPS inadequately supported to gage standards. CORRECTIVE ACTIONS Supports evaluated per NEDC 87-070 and determined that supports could cause the line to break during event. Supports removed per MWR 87-2024. 1 (V3 ( L
0 NRCESPECTION REEDORT NO. 87-10._SSFI 4 Su!HHlilry. Specific seismic issued / concerns (SSFI Items 3.1.2.(5) (a-e)) were addressed per District correspondence to NRC dated July 24,1987. These items have been analyzed (per NED calculation) and subsequently corrected during design changes or MWRs. Majority of seismic issued identified during SSFI deal with seismic interaction. CNS will be reviewed for seismic systems interaction as part of the required implementation of USI A-46. O Letter from R. J. Bosnak of the Division of Safety Review and Oversight to R. D. Liaw of the Division of BWR Licensing has previously recommended that no specific actions are required at CNS pending implementation of A 46. Items at CNS which do not have supporting seismic calculations, do not fall with SQUG database, and are in close proximity to essential systems, will be analyzed and modified, if necessary, to ensure they are restrained to withstand a SSE event and not damage any essential component. l O t utM W j 1.
1 O TOPIC 27-O DlSIRLCT PLANS FOR LICENSE RENEWAL O pag e s.im men pi e m
[ ATTACHMENT 4 l NRC. CIVIL /STRDCTURAL WALKDOWN ITEM DRAWINQ DESCRIPTION 1. Fig 2 Dress into Greens. 2. Fig 2 Rx Bldg 903 North general area, drywell equipment hatch, equipment anchorage and condition. 3. Fig 2 Take elevator up to 1001 level. 4. Fig 5 Inspect Puel Pool. 5. Fig 5 Inspect SLC tanks, anchorage and l condition. 6. Fig 5 Take elevator down to 958 level. '7. Fig 5 Inspect containment wall. 8. Fig 5 MG Set Oil Pump area; equipment anchorage and condition. 9. Fig 5 RWCU Mixing and Precoat area; Fuel ~ Pool wall, equipment anchorage and condition. 10. Fig 5 SBGT Rm; equipment anchorage and condition. 11. Fig 5 -MG Set area; inspect Fuel Pool and Containment wall, equipment anchorage 1 and condition. 12. Fig 5 REC Surge tank. t 13. Fig 5 Take stairway down to 931 level.. O 14. Fig 4 MG Set Lube Oil Cooler area; equipment s anchorage and condition.- 15. Fig 4 Inspect the RWCU and RHR heat exchangers removable walls. 16. Fig 4-RHR Heat Exchanger Rm 1B;' equipment ar.chorage and condition. 17. Fig 4 REC Heat Exchanger anchorage and Containment wall inspection, i 18. Fig 4 Take stairway down to-903 level. 19. Fig 2 Suppression Pool access hatch; inspection. 20. Fig 2 Alternate Shutdown Rm structure. 21. Fig 2 RHR Heat Exchanger 1B'Rm; equipment anchorage and condition. -22. Fig 2 ' Inspection of the 903 Southwest drywell equipment hatch._ 23. Fig 1 Southwest 881 level; equipment anchorage and condition.
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Fig 1 RHR_ Pumps 1B and 1D Rm;_ bldg-structures, -penetrations, equipment anchorage and. condition. 25. -Fig 1 HPCI Rm; b1dg structures,_ penetrations,- equipment anchorage and-condition._ 26. Fig 2 Personnel Air. Lock area; inspection. 27. Fig 2 Go-through the PCMs and dress out into yellows.
MRC CERfETEDQTITRAL_RALEQQRM (CONTINUATICN) ITEM DRAWING DESCRIPTION 28. Fig i Northwest 801 level equipment anchorage i and condition. 29. Fig 1 RllR Pumps 1A and 1C Rm; scismic M accoloromotor, b1dg structures, penetrations, equipment anchorago and condition. 30. Fig 1 Supprecolon Pool area; bldg structures, panotrations, equipment anchorage and condition. 31. Fig 2 Go through PCMs and dress into street clothes. 32. Fig 3 Cable Expansion Rm; inspection of structures and penetrations. 33. Fig 2 Control and Office bldg roofs through north stairway in Office b1dg. 34. Fig 4 Inspect the Critical Switchgoar Rms. 35. Fig 4 Stop by the Main Control Rm. 36. Fig 3 Cable Spreading Rm; accoloromotor, inspection of structure and penetrations. 37. Fig 2 Auxiliary Relay Rm inspection. W 38. Fig 2 RPS Rm 1B inspection. 39. Fig 2 Battery Rm 1B inspection. 40. Fig 2 DC Switchgoar Rm 1B inspection. 41. Fig 2 SWGW tanks, equipment anchorage and condition. 42. Fig 1 Control b1dg basement structure, equipment anchorage and condition. 43. Fig 1 Emergency Condensato Storage Tank area; anchorage and condition, b1dg structuro. 44. Fig 2 DG-1 Rm; Day tank, DGSA receivors, bldg structure, equipment anchorage and condition. 45. Fig 2 DG-2 Rm; Day tank, DGSA receivers, b1dg structure, equipment anchorage and condition. 46. Fig 2, 6 Servico Water pump Rm in the Intako Structure; SWGW tanks, equipment anchorage & condition. 47. Fig 2, 6 Intake Structure; walk around scroons, ir pect b1dg structure, equipment anchorage and conditions. 48. F i.g 2 Seismic instrument station. 49. .ig 2 DGDO Storage Tanks. 50. Fig 2 ERP Tower. 4
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