ML18153C328
| ML18153C328 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 08/15/1990 |
| From: | Buckley B Office of Nuclear Reactor Regulation |
| To: | Stewart W Virginia Power (Virginia Electric & Power Co) |
| References | |
| CON-FIN-A-3871, GL-88-005, TAC-73008 NUDOCS 9008170058 | |
| Download: ML18153C328 (20) | |
Text
(
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 Docket Nos. 50-280 and 50-281 Mr. W. L. Stewart Senior Vice President - Nuclear Virginia Electric and Power Company 5000 Dominion Blvd.
Glen Allen, Virginia 23060
Dear Mr. Stewart:
August 15, 1990
SUBJECT:
PREVENTION OF BORIC ACID CORROSION AT SURRY POWER STATION, UNITS 1 AND 2 (GENERIC LETTER 88-05) (GENERIC TAC NO. 73008)
The purpose of this letter is to advise you that our audit of your boric acid corrosion prevention program has resulted in an acceptable finding and we now consider this issue to be closed.
On August 6-8, 1989, the NRC staff and our consultant visited the Surry Unit 1 and 2 plants to audit the program to prevent boric acid-related corrosion.
The audit team included K. Parczewski (NRC) and C. Czajkowski (consultant, Brookhaven National Laboratory). Boric acid corrosion prevention requirements were described in Generic Letter 88-05 which was issued on March 17, 1988, and requested the implementation of such a program by all licensees of operating PWR~ and holders of construction permits for PWRs.
In your letter dated June 3, 1988, you provided a description of, and a commitment to, a boric acid leakage monitoring and a corrosion prevention program for Surry, Units 1 and 2.
A copy of the trip report covering the results of the audit which was prepared by our consultant is enclosed. The staff has reviewed this repor,t and agrees with its findings and the conclusion.
On this basis and the observations made during the audit, we conclude that you have adequately implemented a program for monitoring small primary coolant leakage through carbon steel components caused by boric acid corrosion as described in your submittal dated June 3, 1988, for the Surry facility. The results of the audit will be used along with audit results from other plants in our overall determination of future actions to be taken regarding NRC's final resolution of this industry-wide generic issue.*
Mr. W. August 15, 1990 Our review of this issue for Surry, Units 1 and 2 is closed.
Enclosure:
Brookhaven National Laboratory Report cc w/enclosure:
See next page Distribution:
Docket File NRC & local PDRs PDII-2 R/F SVarga GLainas HBerkow DDiianni CMcCracken
~
PM:PDII-2 BBuckley g IJt'i9o OFFICIAL RECORD COPY Document Name:
SURRY TAC NO. 73008 Sincerely, Original signed by Bart C. Buckley, Senior Project Manager Project Directorate II-2 Division of Reactor Projects I/II Office of Nuclear Reactor Regulation BBuckley OGC EJordan ACRS (10)
FParczewski FWitt DMi ller MSinkule, RII
. *\\
Mr. W. L. Stewart Virginia Electric and Power Company cc:
M1chael W. Maupin, Esq.
Hunton and Williams Post Office Box 1535 Richmond, Virginia 23212 Mr. Michael R. Kansler, Manager Surry Power Station Post Office Box 315 Surry, Virginia 23883 Senior Resident Inspector Surry Power Station U.S. Nuclear Regulatory Commission Post Office Box 166, Route 1 Surry, Virginia 23883 Mr. Sherlock Holmes, Chairman Board of Supervisors of Surry County Surry County Courthouse Surry, Virginia 23683 Mr. W. T. Lough Virginia Corporation Commission Division of Energy Regulation Post Office Box 1197 Richmond, Virginia 23209 Regional Administrator, Region II U.S. Nuclear Regulatory Commission 101 Marietta Street N.W., Suite 2900 Atlanta, Georgia 30323 C. M. G. Buttery, M.D., M.P.H.
Department of Health 109 Governor Street Richmond, Virginia 23219 Surry Power Station Attorney General Supreme Court Building 101 North 8th Street Richmond, Virginia 23219 Mr. E. Wayne Harrell Vice President - Nuclear Operations Virginia Electric and Power Company 5000 Old Dominion Blvd.
Glen Allen, Virginia 23060 Mr. J.P. O'Hanlon Vice President - Nuclear Services Virginia Electric and Power Company 5000 Old Dominion Blvd.
Glen Allen, Virginia 23060 Mr. R. F. Saunders Manager - Nuclear Licensing Virginia Electric and Power Company 5000 Old Dominion Blvd.
Glen Allen, Virginia 23060
A.
Introduction BORIC ACID PREVENTION Trip Report e
FIN A-3871 TASK ASSIGNMENT NO. 4 ENCLOSURE On August 6-8, 1989, a USNRC audit team visited the Surry Nuclear Power Station Units 1, 2. The team was comprised of Messrs. K. Parczewski of the USNRC and Mr. C. Czajkowski of Brookhaven National laboratory (BNL}.
The purpose of the plant visit was to audit the licensee's implemented program for prevention of carbon steel corrosion by boric acid in the reactor pressure boundary of the plant.
The*verification of the program implementation took the form of an audit
.of the Units' written procedures, interviews with plant staff personnel and verifying that the techniques used by the utility were proper and performed by adequately trained/certified personnel.
The guidelines for the audit fell into four broad areas of concern which should encomP,ass the utilities' elicited responses to NRC Generic Letter 88-05.
B.
Determination of the principal locations where leaks of primary coolant below the specification limits could cause degradation of the reactor pressure boundary components.
The primary document that describes the utility's boric acid corrosion prevention program is "Reactor Containment Leakage Walkdown," (1-PT-10.1 - Unit 1/2 -PT-IO.I - Unit 2).
This procedure was written to provide a vehicle for identifying any "boric acid accumulation and/or unidentified leakage."
The walkdowns for this leakage inspection are done during cold shutdown of the units prior to containment decontamination.
Paragraph 5.2 of the procedures' instruction section requires a check for both source of leakage and potential damage to surrounding components:
"5.2 While performing walkdown, it is important to determine not only the source of the leakage, but also the surrounding components should be examined for any evidence of excessive accumulated boric acid."
Leakage is noted on a checklist devised for this purpose.
Although the procedure complies with the intent of Generic Letter 88-05, it could be expanded to provide a scope of equipment/components/systems to be inspected.
Currently, the procedure only requires: "5.1 Conduct a visual inspection of the Reactor Containment.... "
The 1 i censee al so performs inspect i ans for leaks during hot shutdown "Reactor Coolant Pressure Test,". at (l-PT-11), 2335 psig.
This procedure outlines details for performing leakage checks which include boric acid leakage detection in the Reactor Coolant System.
1
~ -
The utility also utilizes procedure, "Containment Checklist," (l-OP-18/Unit I, 2-0P-18/Unit 2) for leakage checks during cold shutdown and as a final check during hot shutdown.
Attachment I is a copy of the checklist of a 11 the equipment reviewed during these inspections.
A similar checklist is used for Unit 1.
C.
Procedures for locating small coolant leaks As attachments to the procedures "Reactor Containment Leakage Walkdown" and "Reactor Coolant Pressure Test," the licensee has included a procedure for visual examination for leakage.
111.0 This examination shall be conducted to locate evidence of leakage from pressure retaining components, or abnormal leakage from components with or without leakage co 11 ect ion systems.
2.0 The Visual Examination of noninsulated systems shall consist of the exposed external surfaces of pressure retaining components for evidence of leakage.
3.0 The visual examinations of insulated systems may be conducted without the removal of insulation.
Accessible joints and exposed surfaces of the insulation shall be examined for evidence of leakage.
For vertical components (greater than 60°C from horizontal), only the lowest elevation need be examined for leakage.
3.1 Discoloration, staining, boric acid residue and other evidence of leakage on i nsul at ion surfaces and the surrounding area shall be given particular consideration as evidence of component leakage.
If evidence of leakage is found, removal of insulation, to determine the exact nature of the deficiency, may be accomplished with the component depressurized....
11 4.0 For components with external surfaces and joints not accessible for direct visual examination, the surrounding area (including the floor, equipment surfaces underneath the inaccessible component and other areas where leakage may be channeled) shall be examined for evidence of component leakage.
- 5. 0 A visual examination of components with leakage co 11 ect ion systems (Valve packing glands, pump seals, flange gaskets, etc.) shall be conducted. The leakage collection system shall be examined for leakage, or evidence of leakage, from the component to point of termination and for proper disposal of coll~cted fluid at the termination point."
2
e The personnel performing these inspections are trained and certified VT-2 qualified inspectors. During the audit, documentation of certification for the following personnel were checked:
Name Certification L. E. Carter Level. I II VT-2 C. L. Conner Level III VT-2
- s. P. Hami 11 Level I II VT-2 R. L. Adkins Level II VT-2 T. E. Aldridge Level II VT-2 K. L. Okleshen Level II VT-2 D. W. Wong Level II VT-2 D. Quackenbush Level II VT-2 A review was also made of the training module for VT-2 which was used to train and qualify the above.
No discrepancies were noted.
The attachments to the procedures coupled with the training/certification of inspectors complies with the intent of the Generic Letter.
D.
Procedures for evaluating boric acid induced corrosion of carbon steel components in the reactor pressure boundary.
The utility uses Engineering Procedure ENG-84, "Evaluation of Safety Related Components with Excessive Boric Acid," 12/10/87, for evaluating the possible effects of boric aGid induced corrosion on safety-related components, including long term effects of damage.
The procedure uses a checklist for recording the field observations. It is quite detailed and states in part:
"5.4 Assess or estimate the extent of boric acid coverage. Provide sketch in the space provided.
N/A this step if the acid was previously removed.
The use of a camera can preclude the sketch requirement.
If a camera is used, attach the photo-graph to this procedure and so note in the space provided.
Removal of insulation or other interferences may be required, if so, use appropriate procedures as required (Block 3}.
- 5. 5 Determine the source of the boric acid.
The source may be directly connected or removed, leaking or spraying the component.
Note source location in appropriate space (Block 4).
5.6 Determine if the boric acid is wet indicating active corrosion or if the deposit appears dry and inactive.
Note in space provided.
Mark unknown if previously removed (Block 5}.
5.7 Remove the boric acid as necessary to determine the extent of the general corrosion.
This is defined as the approximate percentage of uniform wastage of a surface of the component.
Refer to Table 1 for the effect of boric acid on various 3
materials. Note pitting or cracking visually observed. Note in space provided.
Note Table 2 provides information on material references (Block 6).
5.8 Determine when the general corrosion exceeds 10%, if the component can continue service or recommend repair or replacement.
If the component can continue service, recommended when another evaluation should be made.
N/A this in the space provided if less than or equal to 10%.
Note dry boric acid should not continue to act against the-component (Block 7)."
During the site audit, completed checklisti were reviewed for two evaluations, 5/31/88 and 6/10/88.
No discrepancies were noted. This procedure appears to comply with the intent of Generic Letter 88-05.
E.
Corrective actions taken by the licensee to prevent recurrence of similar tvpes of corrosion.
- 1)
The Surry plant is currently in the process of upgrading the packing in all valves by the use of A. W. Chesterton supplied die-formed graphite-foil rings and split carbon sleeve s*pacers. This is a positive step in reducing stem leakage.
- 2)
Another step in reducing leakage at the Surry plant is the incorporation of live load valve stem packing.
(EQR 88-193).
- 3)
An area which appeared to need fortification was in failure/root cause analysis and evaluation of corrective actions.
For two work orders examined which requested that a failure analysis be performed -
both components had been scrapped prior to the evaluations being performed.
Since good failure analyses/root cause analyses can provide an early warning of potential generic problems, a more uniform approach to failure analyses appears warranted.
- 4)
A plant tour was provided of accessible areas of Unit 1. Work Order/Request were randomly recorded during the tour:
Work Request No.
WR 636339 WR 321655 WR 599654 WR 635331 WR 560287 WR 635360 WR 560252 WR 635163 WR 635152 4
12/30/86 (work finished 1/19/87) 2/28/89 (work completed) not found in computer 12/20/88 (work finished 12/23/88) 2/23/89 (work completed) 8/05/88 (work finished 9/9/88) not found in computer not found in computer
Safety Injection MOV 1842 was bagged, corrosion about the* studs (staining) was evident but no tag was in evidence. This item and WR 635331/635163/635152 (above) are currently being investigated by the utility (D. Wong).
F.
Canel us i ans
- 1)
Procedurally the licensee meets the intent of Generic Letter 88-05.
- 2)
Two areas in need of further evaluation by the utility are:
a) failure/root cause analysis; b) providing a scope of equipment to be inspected during plant walkdowns.
These conclusions were discussed with the utility personnel at the Exit Critique.
G.
Documents Reviewed
- 1.
Letter Response to USNRC Generic Letter 88-05, June 3, 1988, SN #88-170A.
- 2.
Periodic Test Procedure, l-PT-10.1, Unit 1, "Reactor Containment Leakage Walkdown," 5/30/89.
- 3.
Periodi~ Test Procedures, 2-PT-10.l, Unit 2, "Reactor Containment Leakage Walkdown," 1/3/89.
- 4.
Engineering Procedure, ENG-84, Units 1 and 2, "Evaluation of Safety Related Components with Excessive Boric Acid," 12/10/87.
- 5.
Work Orders and the associated Work Requests - Nos.:
536918 566269 566268 536911 566276 566283 566270 566259 536915 566275 566282 566271 566260 566250 566274 536927 566264 566261 566248 536926 566263 566257 536925 536919 566265 566262 566255 536920 566266 566256 566254 566273 566267 536910 566277 329616 544002 544023 541265 544058 544016 544001 544024 541266 541268 544051 544052 544019 541267 541273 544015 544033 544020 541269 329038 544009 544032 544021 541270 297667 544008 544030 544013 541272 167533 544007 544029
.544057 544025 418946 544004 544027
- 541263 541271 328668 544006 544028 541262 544010 409539 544003 544022 541264 544059 409538 409931 409932 409540 5
e
- 6.
Administrative Procedure, SUADM-LR-07, ADM-62, "Failure Trending and Analysis of Safety Related Equipment," 5/17/88.
- 7.
Administrative Procedure, SUADM-M-17, "Minor Maintenance," 4/14/89.
- 8.
Virginia Power Nuclear Operations Visual Training Manual, Section VT-2, Rev.- 0, 2/17/86.
- 9.
Administrative Procedure, SUADM-ENG-08, "System Engineer Failure and Root Cause Analysis," 8/15/88.
- 10.
Operating Procedure, 2-0P-18, "Containment Checklist," 9/15/88 (Unit 2), l-OP-18 for Unit 1.
- 11.
Quality Assurance Audit Reports (Internal): a) S89-19 (6/21/89) and b) S89-02 (5/18/89).
- 12.
Memorandum and associated report, "leakage and Contaminated Area Tracking Report," 4/3/89.
- 13.
Periodic Test Procedure, l-PT-11 (2-PT-11, Unit 2), "Reactor Coolant Pressure Test," 1/8/89.
- 14.
Administrative Procedure, SUADM-ENG-10, "Component Failure Analysis Program," 12/23/88.
- 15.
Engineering Work Requests (EWRs) for:
a)
Live Load Packing Justification (Units 1 and 2),88-193 b)
Evaluation Acceptability for Chesterton Valve Packing Materials (Units 1 and 2),87-390 c)
Evaluate MI Packing and Bushings,89-153 d)
Replacement of CH Valve (l-CH-366),87-033 H.
Personnel Interviewed
- 1.
The following personnel were present at the entrance meeting:
K. Parczewski NRC/NRR C. Czajkowski BNL G. D. Miller VEPCO Licensing J. A. Price VEPCO D. Rogers VEPCO R. MacManus VEPCO D. Wong VEPCO
- 2.
The following personnel were interviewed during the audit:
D. Hart D. Wong VEPCO - QA Supr. - Auditors VEPCO - ISi Engr..
6
G. D. Miller VEPCO - Licensing K. Okleshen VEPCO - Operator R. A. Shaw VEPCO - H. P. Technician
- 3.
The following personnel were present at the Exit Critique:
K. Parczewski NRC/NRR C. Czajkowski BNL D. S. Hart VEPCO - QA Supr.
- w. R. Benthall VEPCO - Licensing Supr.
G. D. Miller VEPCO - Licensing D. Wong VEPCO - ISi Engr.
R. MacManus VEPCO - Engr. Supr.
T. Sowers VEPCO - Engr. Supt.
7
IN1TIALS 5.0 2-0P-IB Page 5 of 13 SEP 1 5 1988 Procedure AlTACHHENT..1..
PART I -.. cou, SHUTDOWN 5.1.1 Sump screens clean, clear of debris, all screens and grating installed.
5.1.2c Containment basement clear of debris.
- 5. 1. 3
- Loop Room lA clear.
5.1.4 Loop Room lB clear.
5.1.5 Loop Room lC clear.
5.1.6 Pressurizer cubicle clear.
5.1.7 Incore instrumentation room clear.
5.1.8 5.1.9 5.1.10 5.1.11 5.1.12 5.1. i3 5.1.14 5.1.15 Operating deck (47'4" level) clear.
Manipulator crane parked (Dillion Readout Removed).
Polar crane parked and hoists up, ensure large hooks not above reactor vessel and crane is de-energized at 2Bl-264 (upper cable vault).
Reactor Cavity*and Transfer canal clear.
Equipment properly stored (scaffolding, tools, welder, etc.)
- .*:*'"'-*****. :*****,-*-*:*~-~:-****.*****>
- -*.-... *~*--**.,*,:
-1.....
--~-.~---
Required shield blocks installed IAW Attachment 1.
Primary or secondary system leaks noted, documented and corrected.
(2)
Oil levels normal (check when completed).
1-RC-P-lA 1-RC-P-lB 1-RC-P-lC
Sec. Officer SRO NOTE:
Verified 5.1.16 5.1.17 f
Verify RCP oil collection tanks are empty, inspect collection system for integrity.
- a.
"A" Loop
- b. "B" Loop
- c.
"C" Loop Equipment hatch integrity checked.
2-0P-lB Page 6 of 13 SEP l 5 1989 an~ visually 5.1.18 Containment floors dry.
5.1.19 Personnel air lock integrity checked.
5.1.20 Lock open all Aux.
Feedwater MOV manual isolation valves.
2-FW-92 2-FW-61 2-FW-30
- 2-FW-93 2-FW-62 2-FW-31 5.1.21 Verify 2-RC-107 shut.
5.1. 22.
Conduct a security valkdown of containment by Security personnel and supervisor.
5.1. 23 Ensure jib crane restraint installed.
Stipulate on tagging record that restraint must be removed prior to clearing red tag.
- 5.-1. 23. l
- ** Red tag jib crane power supply (2Cl-112B) _,._.;.-;:-::,-7~,.., *.,.,*;,
4 to Shift Supervisor prior to leaving cold shutdown.
,)
2-0P-lB Page 7 of 13 SE? 1 5 1988 5.1.24 Verify the following valves on their Backseat.
(1)
Loop Rooms are~ to be locked at this time.
(2). List the W.R. No. of those leaks noted:
Valve No.
c Pzr. Cubicle 2-RC-133 2-RC-132 2-RC-130 2-RC-126 Rx. Head 2-RC-163 "2-RC-178 2-RC-179 18' Level "C" Loop 2-RC-128 2-RC-127 2-RC-129 2-RC-131 2-RC-87 2-RC-88 2-BD-21 2-BD-24 2-RC-93 2-RC-94 2-RC-95 2-RC-165 (locked (locked (locked open) open) open)
Description Vapor Sample LT 460 isol.
LT 459 Isol.
LT 461 Isol.
Rx Head Vent Isol.
Rx Head Vent Isol.
Rx Head Vent Isol.
Liquid Sample LT 461 Isol.
LT 459 Isol.
- ~,*
LT 460 Isol.
TH RTD Return TC RTD Return "C" S/G Root Isol. Valve "C" S/G Root Isol. Valve RTD Return Flow FIC 492 RTD Return Flow FIC 492 RTD Return Flow
.... ;.* "**.* **:..,... -~**
RVLIS "C" Loop Hot Leg Isol. RTD Mainifold
,)
"B" Loop 2-RC-61.
2-RC-62 2-RC-63 2-RC-55 2-RC-56 2-BD-11 2-BD-14 2-RC-166 18' Level "A" LooE 2-RC-16 2-RC-17 2-BD-l 2-BD-4 2-RC-22 2-RC-23 2-RC-24 2-RC-167
. -*** *' ~
2-RC-164 3' Level "A" Loo:e 2-CH-314 2-RC-25 2-RC~26 l-RC-27 2-RC-28 f
e RTD Return Flow FIC 491 RTD Return Flow FIC 491 RTD Return Flow RTD TH Return RTD TC Return "B".S/G Blowdown Root Valve "B" 5/G Blowdown Room Valve 2..:oP-lB
- Page 8 of 13 SE? 1 5 19S8 RVLIS "B" Loop Hot Leg Isol. RTD Manifold Description RTD TH Re turn RTD TC Return "A" 5/G Root Isol. Valve "A" S/G Root -Isol. Valve RTD Return Flow FIC 490 RTD Return Flow FIC 490 RTD Return Isol.
RVLIS Seal Table Isol. Incore Room RVLIS Located on Head Vent.. Line (Long Ladder Required)
Letdown Isol.
FIC-414 FIC-415 FIC-416 FIC Low Side
)
~
2-0P-lB Page 9 of 13 2-RC-29 TC Drain SEP 1 s 1sas 2-RC-30 TC Sample 2-RC-31 Fill Hdr.
2-RC-2 FIC 480B Isol.
2-RC-3 FIC 480B !sol.
2-RC-4 f
FIC 480A !sol.
2-RC-5 FIC 480A Isol.
2-RC-6 TC Bypass Flow Ref.
2-RC-10 TH Sample "B" Loop 2-RC-37 PT 405 Isol.
2-RC-38 PT 405 Isol.
2-RC-39 Sample TH RX Side 2-RC-41 FIC 481A Isol.
2-RC-42 FIC 481A !sol.
2-RC-43 FIC 481B !sol.
2-RC-44 FIC 481B Isol.
2-RC-45 Bypass Ref. Flow 2-RC-49 TH Sample Loop Side
. **.... *\\
. -=-:.. **.
.... ~-.-... ~,1-.:---~~:-,....... *;-...
- 2-RC-64 FIC 424 2-RC-65 FIC 425 2-RC-66 FIC 426 2-RC-67 FIC Low Side 2-RC-68 TC Drain 2-RC-69 TC Sample 2-RC-70 Fill Hdr.
)
3' Level "C" Loop 2-RC-71 2-RC-73 2-RC-74 2-RC-75 L:
2-RC-76 2-RC-77 2-RC-81 2-RC-96 2-RC-97 2-RC-98 2-RC-99 2-RC-100
- 2-RC-101 2-RC-102 Basement 2-RC-33 5.1. 25 e
2-0P-lB Page 10 of 13 SEP 1 5 1988 Description PT 402 FIC 482A !sol.
FIC 482A !sol.
FIC 482B !sol.
FIC 482B !sol.
Bypass Flow Ret.
.FIC 436 FIC Low Side TC Drain TC Sample Loop Fill Head Seal leakoff isol.
Put the following valves on 1/16 in. deflection on the backseat:
"A" Loop MOV-2590 MOV-2591 "B" Loop MOV-2592 MOV-2593 Hot Leg !sol.
Cold Leg !sol.
Hot Leg !sol.
Cold Leg !sol.
)
. ~-
Inst. Tech.
Inst. Tech.
Inst. Tech.
"C" Loop MOV-2594 MOV-2595 Hot Leg Isol.
Cold Leg Isol.
Steam Flow Transmitters e
2-0P-lB Page 11 of 13 SEP 1 5 1988 5.1. 26 Verify the following valves open and transmitters are operable.
"A" S/G Root Valves (OPEN) and Backseated *.
2-MS-98 2-MS-97 2-MS-99 2-MS-96 "A" S/G Transmitter Isolation Valves OPEN FT-2474 FT-2475 "B" S/G Root Valves (OPEN) and Backseated.
2-MS-131 2-MS-129 2-MS-132 2-MS-130 "B" S/G Transmitter Isolation Valves PPEN FT-2484 FT-2485 "C" S/G Root Valves (OPEN) and Backseated.
2-MS-170 2-MS-167 2-MS-168 2-MS-169 "C" S/G Transmitter Isolation Valves OPEN.
FT-2494 FT-2495
.... *- *:- **:. ** *.: ~..
-:........ -***.- *4:*.:..... _,..: *. *
- 1*, ~.. * ~-.,_ r.: _.. -.- **
Completed By: _____________________ _
Date: ---------------------
)
INITIALS
. * '1'*
2-0P-lB Page 12 of 13 SEP 1 5 1388 5.0
- Procedure PART II - HOT SHUTDOWN 5.2.1 5.2.2 5.2.3 5.2.4 Primary and secondary system leaks noted, documented corrective (1) actions initiated as required.
Especially the following areas:
(Notify SRO of any found leaking).
- a.
Loop Rooms
- b.
RHR flat
- c.
Penetration area Containment floors dry*.
Verify* the following valves are backseated to 1/16 in.
deflection:
"A" Loop MOV-2590 MOV-2591 "B" Loop MOV-2592 Hot Leg Isolation Cold Leg Isolation Hot Leg Isolation MOV-2593 Cold Leg Isolation
- ~!*.;"a;... ::'!:::****!-...* '" * * ***~ **.:..* *.t~;-; :t': ** :.:. :" :_~;.:.... '".,.... :,:.
- *;.~..,, -< *.-: :** r\\ ** :-t '.' *:.: *,-?r~:4->o, * ~... *.13*~ '-,.. *:,*.-..:.,:~ t -,*.~.:e:.*(!* *, * ;""_.--...,*.,*.
"C" Loop MOV-2594 MOV-2595 Hot Leg Isolation Cold Leg Isolation Make a visual inspection of the pipe hangers and associated accumulator piping, report signs of any damage to the shift supervisor.
)
2-0P-lB y
Page 13 of 13 SEP ~ 5 1988 5.2.5 List the W.R.# of those leaks noted: ---------
Completed By: -------------
Date: -------------
......... "'*\\*"
~........ ~-... *.. ;:* ! :.'
. ****. ;-**.:..-:~--
'I.
~
~. '-.
ATTACHMENT 1 UNIT 2 CAVITY
- 2-0P-lB Page 1 of 1 SEf' l s 198a
.......,,.,*.-*,'.:~,*
r=Removed