ML20236B005
| ML20236B005 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 10/07/1987 |
| From: | Akins M, Croley B, Humenansky D SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | |
| Shared Package | |
| ML20236A981 | List: |
| References | |
| NUDOCS 8710230291 | |
| Download: ML20236B005 (37) | |
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'I EXPANDED AUGHENTED SYSTEM REVIEH AND TEST PROGRAM (EXPANDED ASRTP)
EVALUATION OF THE COMPONENT COOLING HATER (CCH) AND .
TURBINE PLANT COOLING HATER (TCH)
SYSTEMS SUBMITTED BY:
M. J. AKINS DATE:
b TEAM LEADER j CONCURRENCE: Md WMe~ts IS DATE: 10 ' b ' S 7
/ DAVID HUMENANSKY
( EXPANDED ASRTP PROGRAM LNAGER CONCURRENCE:
[ DATE: '
[D/7/I7
~B0B CROLEY /
DIRECTOR, NUCLEAR TECHNICAL SERVICES'
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3.0 SCOPE J ' 75 c 1 4.0 OVERALL RESULTS'ARD CONCLUSIONS 6 i
5.0 SPECIFIC CONCERNS 8 L
'5.11 AcknowledgediValid) Concerns
.i 5.2 Opent(Potential) Concerns i:
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6.0 ATTACHMENTS; r ,
I 6.1 ListofdocumentsReviewed i
, 6.2 Status of RIs a.
Detailed Observations - Requests for Information 6.3 i
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EXPANDED AUGMENTED SYSTEM REVIEH AND TEST PROGRAM
[h/% , EVALUATIONOF'HyCOMPONENTCOOLINGHATER/TURBINEPLANTCOOLINGHATERSYSTEM w
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1.0 INTRODUCTION
1 g Wg,- The Ranche,Seco Expanded Augmented System Review and Test Program !
n , [ASRTR evaluation effort involves an assessment of the sM, p' effectiveness of the System Review and Test Program [SRTP] and an gg ~ , analysis of the adequacy of ongoing programs to ensure that systems e' will continue to function properly after restart. The Expanded J ASRIP:is a detailed system by system review of the SRTP as implemented on 33 selected systems and an in-depth review of the engineedng, modification, maintenance, operations, surveillance, I 4
inservice testing, and quality programs. It also conducts a review, o
.on a sampling basis, of many of the numerous ongoing verification and review programs at Rancho Seco.
Six culti-disciplined teams composed of knowledgeable and experienced personnel are tasked with performing the Expanded ASRTP. Each multi-disciplined team consists of dedicated personnel with appropriate backgrounds to evaluate the operations, maintenance . engineering, and design functional areas.
Independence, perspective, and industry standards provided by team '
mtsbers with consultants, architect engineer and vendor backgrounds l are joined with the specific plant knowledge of SMUD team members. '
FEach team performs an evaluation on a selected system using the same fundamental evaluation techniques employed by the NRC in the ASRTP inspection. System Status Reports are used as the primary source of leads for the teams. They are augmented with references to available source and design bases documents as needed. Team synergism and communication is emphasized during the procu s in order to enhance the evaluation. Each team prepares a report for
< each completed selected system evaluated. This report is for the CCH/TCH system.
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$).if, %% The objectivps.of the Expanded ASRTP evaluation 'are to (1). assess : !
t nq. .Pl? the adequacy of activities and systems in support of restart and (2)'
.,; > I '4 evaluate the effectiveness 'of established programs for ensuring -
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'3.0 SCOPE '
r To accomplish the first objective, the Auxiliary Systems. team evaluated the CCH/TCH system to determine whether: 1
- 1. The system was capable 'of pt.rforming the functions
. required by its design bases.
- 2. Testing was: adequate to demonstrate that the system would perform all of the functions required.
- 3. System maintenance (with emphasis on pumps and valves)-
was adequate to ensure system operability'under postulated accident conditions.
- 4. Operator and maintenance technician training was adequate.to ensure proper operations and maintenance of.
the system.
- 5. Human factors relative to the system and the system's
. supporting procedures were adequate to ensure proper.
system operations under normal and accident conditions. To accomplish the.second o'bjective, the team reviewed the programs-as implemented for the system in.the following functional areas:
- 1. Systems Design and Change Control-
- 2. Maintenance
- 3. Operations and Training
- 4. Surveillance and Inservice Testing l
- 5. Quality Assurance
- 6. Engineering Programs The team reviewed a number of documents in preparation for and ,
during the Expanded ASRTP evaluation. This list of documents is ' found in Attachment 6.1. ' The preliminary source of leads for the team were the problems identified in the CCH/TCH System Status Report. h.rious source documents such as the USAR and Technical Specifications and available design bases documents were reviewed as needed to augment the information needed by the team. The evaluation of the CCH/TCH system included a review of pertinent portions of support systems that must be functional in order for the system to meet its design objectives.
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# j 4.0..OVERALL RESULTS AND CONCLUSIONS 'I
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l The more significant issues identified pertaining to the adequacy of i
.the SRTP and the effectiveness of programs to ensure continued safe I operations after restart are summarized below. The summary focuses on the weaknesses identified during the evaluation. ' Attachment 6.3 provides detailed findings by providing the Request for Information (RI)- forms that are used by the Expanded ASRTP teams to identify potential concerns during the evaluation. Section 5.0 lists the specific concerns identified.by the teams.. The numbers in brackets after each individual summary or concern refer to the corresponding-RIs in Attachment'6.3.
The evaluation of the CCH/TCH system has ' identified.several' concerns.~ Some are system specific, identified in subsection 4.1, ,) and some are. generic, identified in subsection 4.2. . 4.1 'CCH/TCH System Specific 4.1.1 -The evaluation team have not been able to locate calculations.to verify the design flow rates for the CCH/TCH pumps. [RI 249] Several specific items were found indicating inadequate flow k rates to components. These items included: excessive conciete temperatures at the Main Steam and_ Feedwater
-penetrations, flow rates to the RCP jacket coolers less than manufacturer requirements, flow balancing using questionable acceptance criteria, and errors in calculations. ~[RI 242]
[RI 248] [RI 269] [RI 271] ] Modifications were made to CCH/TCH piping for the RCP jacket coolers which do not have overpressurization protection. This protection was provided prior to the modification. [RI 268] i 4.1.2 Operations
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A review of selected Chemistry Logs for the CCH/TCH system identified an inadequate control of Hater Chemistry which could lead to a deterioration of system piping and components. [RI 258] 1
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'0VERALL'RESULTS AND CONCLUSIONS- (Continued) 4.2 Generic Concerns
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4.2.1 . Failure of Cast Ir'on Discs in Butterfly Valves.
.. H A review of CCH/TCH Hork Requests revealed several butterfly j valves, which have cast' iron discs, were experiencing cracks )
and breaks.: The team checked several other butterfly valves. . of the.same. type and model in other systems and found the j same problems. -[RI 223) [RI 235] 4.2.2 - Engineering Action Request (EAR) :) l
' AP.78 exists for initiating engineering, action in response- '
to questions and concerns raised by plant personnel.
'However, the team found that some specified actions were not
. completed or identified for tracking prior to EAR closure, q In addition ~, a significant number'of EARS are not being '
reviewed in a timely manner by Engineering to determined if any NCR conditions exist... [RI 254]- 4.2.3- Post Modification / Maintenance Testing A significant number of Hork Request.were found where the pressure boundary was opened. This required a' leak test following closure of the boundary. Most of-the Work Requests identified that inservice leak tests were required. However, the leak testing, that was done, was not , performed in accordance with the. requirements of the. applicable sections of USAS B31.1 Code. [RI 275] ' l l
,.4 . m 5.0. SPECIFIC CONCERNS A list of the specific. concerns that the Expanded ASRTP team. ' believes are new. concerns, .not previously. identified for. resolution,- follows: 5.1 Acknowledged (Valid) Concerns 5.1.1 ' Based upon'the significant number of cast iron butterfly. valve discs that have' experienced cracking.and breaking, a
. generic problem may exist with cast. iron' discs used in-
- butterfly valves. [RI 235]
.5.1.2 Calculation.Z-CCH-M2226.was used to. justify a reduction of
-the CCH flow to the Spent Fuel Cooler. However, it has.
questionable design bases, and.it was not approved. [RI 242] 5.1.3 The Reactor-Building Penetration Coolers have not been able to maintain the penetration concrete-temperatures below the design. limits defined in the USAR.. [RI 248] ! 5.1.4 Calculation have not been found to demonstrate the design basis ~for the CCH pumps. [RI 249] :l 5.1.5 Some corrective actions are either not being completed or.- identified for tracking prior to closure of.the EAR, and possible nonconforming conditions are not identified-in a timely manner. [RI 254]
-5.1.6 The water chemistry in the CCH system is not being adequately controlled and potential deterioration of system components and. piping could have occurred. [RI 258]
5.1.7 There are no provisions for overpressure protection for the RCPjacketcoolers. [RI 268] 5.1.8 The Acceptance Criteria for the "CCW Flow Balance" STP.1053, have been derived from an uncontrolled document. [RI 271] 5.1.9 Based on a review of the "CCH Flow Balance" STP.1053, the RCP Coolers may not be' receiving its required cooling with a single CCW pump operating. [RI 269] 5.1.10 Portions of low pressure piping systems may not have been adequately tested following modifications / maintenance (Generic Issue). [RI 275] 5.2 Open (Potential) Concerns 5.2.1 The missing half of a 6" butterfly valve disc in the CCW system may cause damage to CCH system components. [RI 223]
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J:i .6: 6.0. ATTACHMENTS 6.1 . List' of Documents Reviewed 6.2 Status'of RIs 6.3' Detailed.0 observations - Requests for Information i m 'k l i i
LIST OF DOCUMENTS REVIEHED Calculations Z-CCH-M2226 (05-14-87) Z-CCH-H2240 Z-CCH-H0148 (04-07-70) Z-CCH-M2067 (12-02-86) Z-CCH-M0203 (03-26-74) Z-CCH-M2079 (01-15-87) Documents 10CFR50 Appendix A, B, 50.36 Technical Specifications System Design Basis (Draft) DB-5407 USAR Section 5, 9, Amendment 4 CCH/TCH System Status Report, Rev. 1 Chemistry Logs (1/6/86 to 9/21/87) \
"The Behavior of Concrete in Prestressed Concrete Pressure Vessels,"
Browne and Blundell, 03-06-72 HEL Audit Report #0-370 INP0/NSAC SOE Report #81-4, 81-5, 81-6, 81-7 CCH Training Manual, Chapter 7 Station Manual, Volume I Bechtel Design Manuals, Rev. 2, August 1970 , , Precursor Review Task Final Report, Volume 1. August 1986 Lesson Plan 00-21-I-1000 USAS B31.1, Chapter VI Technical Specification Audit Summary FSAR Section 5.2.3.2.2 Drawinas E-203-16 -1,-1A,-18,-1C,-16,-29,-36,-41,-56,-56A,-57.-82A -85A E-304-17 E-104-1,-4 E-105-12A,-17A,-16.-13 E-100-2 E-101 E-1011-5 H-543, Sh 1-4 H-302 H-206 H-204 H-542
- H-525 M-109 M-115 H-164 H-165 H-288 M-299 M-297 ATTACHMENT 6.1 l
4 LIST OF DOCUMENTS REVIEHED (Continued) Meetina Minutes PRC #179, 182, 219 MSRC #22, 29 Vendor Manuals M19.17.1-196 M19.17.1-207 H19.17.1-211 Bingham Installation, Operation, and Maintenance Instructions
-Allis-Chalmers Operating Instructions
. Pall Installation, Operations, and Maintenance Manual Byron-Jackson Instruction Manual Joy Maintenance and Operation Manual Procedures Annunciator Procedures Manual OP-A.8 OP-A.26, Rev. 19 OP-B.4 OP-C.20, Rev. 9 AP.2.00 AP.2.27 AP.44 AP.78 AP.81 AP.82 AP.100 through AP.167 AP.306 EH.106B EM.117A,C EH.126A,B EH.133A EM.154 j I.002 i I.009 l I,016 l I.022 I.022A I.026A I.036A I.038 !
i I.109E I.306 I.412 ATTACHMENT 6.1 l
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LIST OF DOCUMENTS REVIEHED (Continued) Procedures (Continued) I.605 I 653 I.656A,B I.657 H.006 M.013 H.100 H.103 H.111 ! M.114 l M.115 l M.116 j M.130 M.133 M.134, Rev. 2 ; M.159 j MI.004 : NEP-4109 l NEAP-4104 l NEAP-4112 MAP-006 MAP-009 < RSAP-0305 RSAP-0507. SP.205.04 (09-01-77 through 05-09-85) STP.041 , STP.770 l STP.1053 STP.1057A,B System Test Matrix TP.245.1 TP.600-32 IAfli QA 87-51 AD-67-001 through 3 CH 87-001 through 6 EM 87-001 through 63 ET 87-001 through 10 , IC 87-001 through 110 , MM 87-001 through 60 NE 87-001 through 21 NT 87-001 ; OP 87-001 through 59 ATTACHMENT 6.1
,80 & LIST OF DOCUMENTS REVIEHED (Continued)
EG1 A-1805 A-3710 -
'A-3795, Rev. 2 R-0859, Rev. 3 ER1 S-049,.S-722, S-724, S-797, S-532' S-758, S-812. S-5879 S-053, S-532, S-286, S-6389, S-005, S-305, S-2288,.S-733, S-1660, S-340, S-810, S-2113 Maintenance Insoection Data Reoort MM 83A MM.96 Letters Mattimoe (SMUD) to Giambusso (AEC), 12-17 I Rodriguez (SMUD)'to Colombo (SMUD), 11-05-74~
Gillis (SMUD) to Quality Assurance (SMUD), 01-12 . Goodman-(Bechtel) to McColligan (SMUD),- 12-16-80 1 Perkins (SMUD) to Hood (Bechtel), 02-10-87 Schomer (B&H) to Mattimoe (SMUD),- 12-03-70 Schomer (B&W) to Mattimoe (SMUD), 11-08-72 Kennedy (B&H) to Rodriguez (SMUD), 05-23-73 Kellman (SMUD) to Abbott (SMUD), 03-19-74, [ memo] Schomer (B&W) to Mattimoe (SMUD), 03-27-75
-Logan (B&H) to Rodriguez (SMUD), 04-18-74 Logan (B&H) to Rodriguez (SMUD), 05-31-74 {
1 Janis (B&H) to Rodriguez (SMUD), 08-13-75 l Janis (B&H).to Rodriguez (SMUD), 09-05-75 Janis (B&H) to Rodriguez (SMUD), 12-12-75 Janis (B&W) to Raach (SMUD), 06-09-76 Janis (B&H) to Rodriguez (SMUD), 10-19-76 Janis (B&H) to Rodriguez (SMUD), 02-18-77 Janis (B&H) to Rodriguez (SMUD), 10-25-77 Janis (B&W) to Rodriguez (SMUD), 09-07-78 Medina (SMUD) to QA (SMUD), 05-05-80, [ memo] Abbott (SMUD) to Heisberg (SMUD), 11-03-86, [ memo] Raasch (SMUD) to Rodriguez (SMUD), 06-03-81, [ memo] i Goodman (Bechtel) to McColligan (SMUD), 05-22-81 '
- l. Alvi (SMUD) to Rodriguez (SMUD), 05-05-78 l
ATTACHMENT 6.1 I \. .. .. . .
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LIST OF-DOCUMENTS REVIEHED (Continued)- Hai I i
'018705, 025284,.030913, 052332, 060341, 085014, 097454, 012975, j
.022639, 043267,.048908, 051950, 052327, 051170, 053902, 062708, i 111004, 085013, 090469 -102977, 107663, 108652, 111435, 105800, !
115845, 123121, 113167, 055466, 053897, 054507. 057913. 057897, 059946, 060122, 070366, 072664, 079209, 078360, 074116, 078211, 079519, 079888,.079894, 079923,~079970, 080259, 076227. 086233, 095808, 095481, 096606, 101186, 100287, 112677, 111002, 053899, 053900, 055464,.055465, 110909, 113167, 110911.L110912, 110640, 100771,'085311, 086304, 077128, 074229, 028647, 110887,=113160, 113178; 111084, 076335, 046823. 110907, 115304,'115306,'115312, 115313, 113081, 045566, 110910. 113082, 113092, 113137, 052821, 049839, 047633, 058711, 052712, 026516, 058010, 047634, 033015,- j 031865, 012531, 050901, 025260,'018734, 018736, 018737, 025296, ; 025297, 025298, 025299, 031741, 022250, 110848, 076082, 076090,' 109919, 062926, 076094, 093071, 018743, 103233, 069751, 088162, 040406, 040407, 043918, 045147, 049710, 049853, 060238, 060280 L 1 ATTACHMENT 6.1
h* STATUS OF RIs Attachment 6.2 provides RI status as of this report date. An'RI is considered closed if the Team Leader was convinced a potential concern was. not valid or not significant enough to be an RI. An RI.would also be closed if. requested information was provided. All other RIs are open. Acknowledged RIs are open RIs that have been accepted as valid.by.the responsible-organization and have been stated as concerns in Section 5.0. RI NUMBER STATUS 223 open 225 Closed 235 Acknowledged
-242 Acknowledged 248 Acknowledged 249 Acknowledged 253 Closed 254 Acknowledged i 258 Acknowledged 268 Acknowledged 269 Acknowledged 271 Acknowledged 275 Acknowledged l
i ATTACHMENT 6.2
DETAILED OBSERVATIONS - RE00EST FOR INFORMATION j During an~ evaluation,.all potential concerns are documented on Request for Information sheets (RIs) that are sent to the responsible organization to receive their input concerning the potential concern, i
,RIs are also used to request information that the EASRTP team is having j difficulty obtaining. 1 1
These RIs are considered drafts throughout the entire evaluation until I they become part of the report. Responsible organizations can accept the l potential concern as valid or they may disagree with the potential l concern. If they disagree, they can submit information that convinces j the EASRTP team members that the potential concern is not valid, or they. may redirect the EASRTP members to better focus the concern. RIs l developed during the system evaluation comprise this section of the ! report. .l l I f 1 mmmeur ATTACHMENT 6.3
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' REQUEST FOR INFORMATION (RI)'
l RI NO: _213__ SYSTEM CODE: CCH ISSUE DATE: 09-15-87
SUBJECT:
HALF OF 6" BUTTERFLY DISC LOST IN CCH SYSTEM i DEPARTHENT: MAINTENANCE C0ORDINATOR: J. DARKE TEAM LEADER: M.J. AKINS I
' POTENTIAL CONCERN /00ESTION:
The missing half of a 6" butterfly disc may cause damage to CCW system components. During a review of HR #113167 (refurbish CCH-004, a 6" Allis Chalmers butterfly valve.model 150R, miniflow bypass around P-462A) the following statement was-found: . " Removed valve and operator. Found disc on valve broken-half of disc missing."' Based on discussions with the responsible foreman, plant mechanics and. cognizant engineer, the following information was received: !
. A Boroscopic inspection was made to locate the broken half of the
' disc.
The inspection was through a check valve (CCH-007) into the CCW pump (P-4628) and through discharge valve CCH-013 of the heat exchanger (E-4608). However, the disc half was not found. Therefore, the disc half or its pieces which are still somewhere in the system, could cause significant damage to, or blockage of flow to system components. This includes the potential seizing of the pump impeller.
. No NCR had been w+' + n addressing this problem.
This RI is open because Maintenance believes: a) If the piece remained intact, it probably will not travel throughout the system, and b) It most likely got broken into smaller pieces, and since they have experienced no problems to date, that it is unlikely to experience a problem in the future. ATTACHMENT 6.3 j f
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j i c, :. REQUEST FOR INFORMATION (RI) RI N0:' 225 SYSTEM CODE: GENERIC ISSUE DATE: __09-17-87
SUBJECT:
USE'0F TEFLON TAPE ON TUBING INSTALLATIONS DEPARTMENT: MAINTENANCE COORDINATOR: J. DARKE l ! TEAM LEADER: D. HUMENANSKY j l . ,) l 1 POTENTIAL CONCERN /0UESTION: ! l
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Teflon tape is being used on instrument pipe threads at numerous l locations inside the plant. 1 The use of teflon tape has been banned at many nuclear sites in the'U.S. because it contains fluorides which could induce IGSCC and.because it ! does not disintegrate when it enters a system thus causing a blockage of ' small flow areas. Plugging has been experienced at several nuclear plants.
. Teflon tape was found on instrument air lines in the HFH System -
.during a walkdown of the system.
. In excess of 20 instruments and valves in the EDG had teflon tape on them.
. Several locations in the DHS system had teflon tape on pipe threads.
. HR #01349180, page 3, step 4 for the AFH systems explicitly states that teflon. tape is to be used.
. Letter SOM-39 from J. P. Kennedy of B&W to D.G. Raach of SMUD, dated 16 April 1973, expressed concern about the use of Teflon Tape in the Reactor Coolant and Auxiliary Systems. This letter state the reasons why Teflon tape should not be used.
This RI is closed because actions initiated by its identification have been implemented, and documents controlling the use of teflon tape have been generated. ATTACHMENT 6.3
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REQUEST FOR INFORMATION-(RI) RI NO: 235 . SYSTEM CODE: GENERIC ISSUE DATE: 09-21-87
SUBJECT:
CAST IRON DISC'S USED IN BUTTERFLY VALVES DEPARTMENT: NED' COORDINATOR: T. TELFORD q TEAM LEADER: M.J. AKINS
. POTENTIAL CONCERN /0UESTION:
Based 'upon the:significant number of butterfly valve discs that have experienced cracking and breaking, a generic problem may exist with cast iron discs used in butterfly valves. These failures could seriously affect the function of the systems in which they are used.- Base' d on a review of work requests specifically addressing butterfly q valves, and discussions with the-System Engineer, cognizant engineer, _ i maintenance foreman, and plant mechanics, the following information was y received: I
. CCH-004, A 6" Allis Chalmers butterfly valve, when being refurbished (HR #113167), was found to have its cast iron disc broken in half.
One half was missing _in the system. (Ref: RI 223)
. CCH-005. A 6" Allis Chalmers butterfly valve, when refurbished l (HR #112677) was found with its cast iron disc cracked almost in nalf.
. MCH-Oll, A 6" Allis Chalmers butterfly valve (HR #110909) was found ,
damaged by debris and its disc was replaced. '
. . MCH-012, A 6" Allis Chalmers butterfly valve, when refurbished (HR #110911) was found with its cast iron disc cracked. j
. MCM-053, (HR #85311), the problem / request section of HR #853111 for a 12" Allis Chalmers butterfly valve states in part " disc parts found downstream." During refurbishment of HCM-053 the disc was not regjaced.
- LV-35023A, A 4" Allis Chalmers butterfly valve, when refurbished on HR #86304, a metal chunk was found stuck in the valve.
. During the performance of STP.1053 (CCH Flow Balance) using a controlotron. flow measuring instrument, CCH-106, a 4" Allis Chalmers j butterfly valve, leaked by at a flow rate of 300 gpm when closed.
CCH-035, a 4" Allis Chalmers butterfly valve, leaked by at a flow rate of 80 gpm when closed. hrs #134171, 134170 were written, respectively. ATTACHMENT 6.3 75
- 5. , ( ... -
RI NO: 235' (Continued)- 4 It has been reported to the team that PCH-009 and PCH-034 (inlet and
, outlet to the E-460A Heat Exchanger) both 30" Allis Chalmers butterfly valves, leak when closed. The discharge of CCH flow from heat exchanger E-460A cannot be maintained above minimum temperature with P-462A running.- TV-46012A. a 20" Allis Chalmers butterfly valve which is supposed to control the temperature (located next to' PCH-034) apparently does not function properly.
HQIE: CCH-114,.An 8" Fisher butterfly valve when. throttled varied from 11 -to 18 turns to close position indicating lost motion or improper valve operation. HR #134904 was written. Butterfly valves with cast iron discs are used in several plant systems. i M l l ATTACHMENT 6.3
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h4 l l 1 L REQUEST FOR INFORMATION (RI) ; j RI NO: 242 SYSTEM CODE: CCH . ISSUE DATE: 09-19-87
SUBJECT:
SPENT FUEL POOL COOLER CCH FLOW REDUCTION DEPARTMENT: NED COORDINATOR:' T. TELFORD - TEAM LEADER: M.J.AKINS I i i' POTENTIAL CONCERN /0UESTION: Calculation Z-CCH-M2226 was used to justify a reduction of the CCH flow to the. Spent Fuel Pool Cooler for the purpose of " balancing" the CCH system.- However, the calculation was never approved based on the concerns presented below, yet'the flow value ni used for the "CCH Balancing." Based on Calculation Z-CCH-M2226, the Spent Fuel Pool. Cooler, CCH flow rate was reduced to 600 gpm from 1000 gpm. However, the. calculation did not demonstrate that the flow reduction was justified. , i
- The calculation used three data points.to correlate Flows vs Pool Temperature.
The first data point is given in Attachment 3 to the calculation with a 120.31*F Peak Pool Temperature, a full- core of.177 Assembly. and both Spent Fuel Cooler and DHR Heat Cooler operating. The calculation also claimed the flow'was 4000 gpm, which was not given in Attachment 3. There were no bases, assumptions or references given for the 4000 gpm flow, nor was the flow defined as the flow through the Speat Fuel Pool Cooler or the OHR Coolers. The second data point was given in Attachment 2 to the calculation, with a 132.27'F, Peak Pool Temperature, a full core discharge with only the DHR Cooler Operating. The calculation claimed a 3000 gpm flow with no justification given. Since the Spent Fuel Cooler is not operating one can only assume the 3000 gpm is not CCH flow, therefore not compatible with data point one.
. The calculation proceeded to construct a correlation using the two data points and assumed a straight line relation on log-log paper.
A minimum of three data points is normally required to define a correlation of variables. An infinite number of curves could exist between two points. A straight line function could exist between two points only if proper justifications were given. No such justifications were indicated in the calculation. ATTACHMENT 6.3 i
. :e .
l RI'NO: 242 (Continued)
.- A third data point was given in Attachment 1 to the calculation .ith -!
113.9'F Peak Pool Temperature. 1/3 core _and only_the Spent Fuel Cooler' operating. The flow was claimed to be.1000 gpm with'no basis..
. A second curve, with a straight line function, was constructed using the third data point. The new' curve was assumed to be parallel to
'the' straight line constructed 1rith the first two data points with a '
fullicore discharge in the pool; The straight-line.and parallel function were assumed without any' basis or: justification.
. The 600 gpm flow was_then. extrapolated at 132.37'F from;the new
.second curve.
.It appears that calculation Z-CCH-M2226, with only three data points with-incompatible bases, was used to produce a set of curves to forcibly, extrapolate at 600 gpm CCH flow. The flow from these data points has no basis, and no justifications were_given for the derivation of the curves. The calculation, does not provide an acceptable engineering .
analysis -to justify the reduction of CCH flow to 600 gpm for_ the Spent Fuel. Pool Cooler. I de ATTACHMENT 6.3 , i
1
- ;. ; v -.
REQUEST FOR INFORMATION (RI) RI NO: 248 SYSTEM CODE: CCW ISSUE DATE: 09-23-87
SUBJECT:
-CCW COOLING'FOR-MAIN FEEDHATER AND MAIN STEAM PENETRATIONS DEPARTMENT: NED COORDINATOR: T.-TELFORD TEAM LEADER: M.J. AKINS POTENTIAL CONCERN /0UESTION:
The Reactor Building Penetration' Coolers do not appear to be capable of maintaining the penetration temperatures below design limits as. defined in the USAR. Originally FSAR Section 5.2.3.2.2 stated, "Tne main high temperature piping. consists of two penetrations for feedwater and two penetrations , for main steam which have a maximum operating temperature range between 464*F and 595*F. ' Thermal' insulation is provided on the outside diameter of each line, and. cooling water is circulated.through a cooling sleeve in the air gap between insulation and penetration liner sleeve. Local reading temperature indicators are installed in each of the four cooled '
. penetrations for monitoring of cooler function. The combination of insulation ~and cooling water circulation is designed to restrict maximum )
temperature rise in the concrete to 150'F." On 10-07-74, Non Conforming Report #S-049 was written, stating that at 15% reactor power, with the penetration cooler isolation valves' full open, and with a CCW temperature of 79'F, the "A" and "B" main steam line penetration temperature indicators read 188'F and 195'F, respectively. On 10-08-74, Nonconforming Report #$-053 was written, stating that some of the penetrations without cooling coils were also experiencing l temperatures that approached or exceeded the limits established in the ! FSAR. ) On 12-17-74, a_ letter was written from SMUD to the AEC stating that during the power escalation test phase, concrete surface temperatures in excess of the FSAR limits had been measured. It stated that they (SMUD) were attempting to correct the problem by adjusting the penetration coolers as well as the flows to those coolers. SMUD also said that final resolution of the problem could not be achieved until the unit was at or near 100% power. SMUD concluded by stating that if they were unable to meet the 150*F limit, an amendment to the FSAR would be submitted to revise Section 5.2.3.2.2 in accordance with Subparagraph C 3430A, page 197 of the proposed ASME code for concrete containment vessels. (The proposed code would allow local hot spots up to 200'F.) Amendment 29 to the FSAR changed the design limit to 200'F. i ATTACHMENT 6.3 I
.n
.g.
'n RI-NO: 248 (Continued On 05-30-75, Non-Conforming' Report #S-340.was written, stating that with I an ambient temperature of 91*F, and a CCH temperature of 88'F, penetration temperatures were as follows: - A". "
Main Steam Line - 195'F, "B" Main. Steam Line:- 200*F, "A" Main Feedwaters--205'F ~and "B". Main Feedwater - 210*F. The 50.59 Safety Analysis that was generated in
- response to this NCR. stated that due to the. location of the temperature Vindicators within the penetrations, the: indicators would read at least
'15'F hiaher than the penetrations themselves. However, there were no calculations or data to support this claim.
L On 12-09-75, STP.041.was written to correlate the temperatures indicated on.the penetration temperature indicators with the temperature of. the ) containment concrete. Measurements were taken using thermocouple at several locations in the penetrations and then compared to those of the local temperature _ indicators. The period of. data collection spanned just I under a year so as to obtain a variety of ambient conditions. The STP
. concluded that on " hot" days the Main Steam Line penetrations.did in fact-exceed 200*F. A review of the data collected in STP.041 indicates that any day with an ambient temperature in excess of 88'F would be defined as
ho t . " . The data also indicates that the local temperature indicators actually tend to read slichtiv lower than the penetration temperatures.
! On-11-01-77, Non-Conforming Report #S-810 was written stating, "The data i taken during STP.041 (Main Steam and Feedwater Penetrations Temperatures) shows that the concrete temperature around the Main Steam line penetrations exceeds 200'F." The NCR directed that additional insulation be.added to the penetrations and that on a day when the ambient temperature exceeded 90*F, readings be taken from the local temperature Indicators to determine if the problem still existed. The results were that the "A" Main Steam penetration now read from 220*F to 240*F while . the "B" Main Steam penetration read 220*F. The added insulation was later removed. i On 11-14-80, Non-Conforming Report #S-2133 was writt6n stating, "FSAR Section 5.2.3.2.2 states that the Main Steam penetrations are designed to ) restrict maximum concrete temperatures to 200*F. Measurements which have been taken indicated there are instances when the temperatures rise slightly above this design value. . The maximum temperature measured to date is 208'F. The NCR S-810-R2 disposition does not resolve @ overtemperature effect on the Reactor Building concrete." The resolution NCR #S-2133 was to apparently misinterpret the FSAR as to the meaning of the phrase, "... restrict maximum temperature rise in the concrete to 200*F." SMUD read this to mean that the concrete temperature was allowed to reach 200*F above the ambient temperature. Since there was no record of penetration temperatures exceeding that level no further action was taken. ATTACHMENT 6.3
. j 1
1
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)
RI.NO: 248 (Continued) l On 08-03-81, a test was conducted to determine if the penetration coolers were performing properly. The initial starting penetration temperature 1 was 235'F. The cooling water to the cooler was then isolated and the I temperature rise recorded. Within 75 minutes the penetration temperature I reached 252*F. Cooling water to the cooler was reinstated and no further 1 action was taken, j A Technical Paper titled, "The Behavior of Concrete in Prestressed Concrete Pressure Vessels," points out that concrete temperatures in excess of 70*C (158'F) will directly affect the moisture content of the concrete which could in turn result in structural degradation and a possible loss of load carrying capabilities. No information was available as to the effects on concrete due to-temperatures cycling above and below design limits on a daily basis over extended periods of time. At present the CCH temperature is maintained approximately 7'F higher than the temperature during any of the previously mentioned dates, which-would result in even higher penetration _ temperatures. l l l ATTACHMENT 6.3 l L_____-___-
, e .
I REQUEST FOR INFORMATION (RI) RI NO: 249 SYSTEH CODE: CCH ISSUE DATE: 09-24-87 l
SUBJECT:
CCH PUMPS 462A & B DESIGN REQUIREMENTS DEPARTHENT: NED COORDINATOR: T.'TELFORD TEAH' LEADER: H.J. AKINS
.s POTENTIAL CONCERN /0UESTION:
Calculations'have not been found to demonstrate-the design basis for the CCH pumps. Based on the'NED calculations log, calculation Z-CCH-H2240 "CCH Pump Verification" that was to be issued by 06-07-87, does not exist.
. System Design Basis (draft) for the CCH/TCH System states: "The system configuration currently requires both CCH pumps.to operate for the system to fulfill its performance requirements. The criainal system desian required single CCH pump operation, however, modifications to the system were made that necessitated operating both CCH pumps." There is no documentation defining the nature of the modifications, nor any calculation to support the change from one pump to two pump operation, nor any information about which parameter was not being met by the one pump operation.
. The pump performance requirements were not defined.
. The gr_igipal systen desian requiring single CCH pump operation was not defined or justified by any calculation.
. Additionally, Calculation Z-CCH-H2079 indicated that the cooling <
jacket and pump coolers for the reactor coolant pump were repiped in parallel instead of in series. There was no calculation or documentation concerning this change. . It presently appears that the entire CCH system depends on the " flow test balance" to justify its performance or design requirement. The flow test balance is a good means to verify the original design flow requirements but cannot be used as the basis for the design of the system. The flow test shTuld support the original design. There were no original design calculations available to establish CCH pump flow and head, or to define the system requirement. ATTACHMENT 6.3 l I L-_ ___ _ _-- _ _ - _
L- i
- i D' REQUEST FOR INFORMATION (RI) u>
l RI NO: 253 SYSTEM CODE: CCH ISSUE DATE: 09-23-87 ,
SUBJECT:
POTENTIAL OFFSITE RELEASE VIA CCH RELIEF VALVE l l DEPARTHENT: NED COORDINATOR: T. TELFORD TEAM LEADER: __M.J. AKINS It EQTENTIAL CONCERN /00ESTION:
", Several of the relief valves associated with the CCH system discharge to storm drainage areas that would allow for a potential unmonitored release
. of radioactive materials offsite.
The relief valve on the CCH return lines from the Circulating Water Pumps discharge to the ground from the water would then flow to a storm drain. The majority of the CCH relief valves in the tank farm are directed to ! the ground, again in the vicinity of storm drains. i The CCH system was contaminated in the past and is still contaminated. Therefore, all leaks from the CCH system must be considered radioactive.
/
This RI was closed because this problem has been identified in the Radwaste System SIRS and is in the QTS system for followup.
]
-1 ATTACHMENT 6.3 1
__________ _ _ _ - 1
, e REQUEST FOR'INFORMATION (RI)
RI NO: 254 Rev 2 SYSTEM CODE: GENERIC ISSUE DATE: 10-05-87
SUBJECT:
- EAR PROCESS DEPARTMENT: 00ALITY COORDINATOR: N. SAMPSON' ,, T. TELFORD l J. McCOLLIGAN ! TEAM LEADER:~ M.J. AKINS
-POTENTIAL CONCERN /00ESTION:
Corrective actions are either not completed or identified-for tracking prior to closure of.the EAR. .Possible nonconforming conditions are not identified in -a timely manner.
- 1. AP.78, Engineering Action Requests, allows closure of the EAR by the :
responsible engineer if either:
- a. All required Engineering actions resulting from the request are completed.
- b. All corrective action documents necessary to complete the Engineering action have been identified for tracking or
- c. No Engineering action is required.
However, the following examples identify instances in which either AP.78 was not adhered to or that corrective actions outside of the scope of Engineering action were not tracked or completed.
- a. EAR IC-87-094 Nuclear Engineering committed to revise drawings to change the designations of pressure switches from high to low.
However, no corrective action document to track this work to completion was identified. No mention was made in the NE response of changes to plant labeling of the pressure switches nor procedure changes required by the designation change even though the EAR was closed.
- b. EAR MM 87-036 Nuclear Engineering identified the need to replace four train limit switches with two spare rotors. However, no corrective action document to accomplish this work was identified and the EAR was closed.
ATTACHMENT 6.3 y ,q
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1
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RI NO: 254 Rev. 1 (Continued) J
.c. EAR EM 87-50 l . Nuclear Engineering recommended that all:8LPB retaining clips showing signs of paint deterioration be repainted with j
=an insulating: erosion resistant-paint. However, no_ j corrective action document to accomplish this repainting was "
l initiated and the EAR was closed.
- d. EARS.MM 87-013 through 019 System ' Engineering requested that 20x20x6 HVS filt'ers be replaced with the: design specified 20x20x2 filters.
However, no corrective action documents to accomplish the. replacement were identified and the EARS were closed. e.. EAR OP 87-054
. Nuclear Engineering recommended the insertion of a table a
.into Operating Procedure C.13B to assist in determining DHS j flow which is necessary to safely cooldown.the plant from- {
outside.the Control Room. Nuclear Engineering also - l, suggested'the permanent fix of installing a flow meter. However, no corrective action document was identified to. ] ensure that this would be completed and the EAR was closed.
- 2. QAP 17, Nonconforming Material Control, Attachment C requires an EAR to be initiated if discrepancies are found between the "as-built" and drawing configuration. Engineering is to issue a drawing. change if the drawing is wrong or an NCR if the as-built is incorrect.
1 However, Policy Section XV of the QAPs requires a nonconforming
. report to be generated when the actual condition is not in f !
accordance with the drawing / design configuration. l Also, the following is a partial list of EARS that were generated per the QAP 17 requirement. All of these EARS have been assigned as a Priority 2 or 3 with no resolution from Engineering. Therefore, the. ' possible nonconforming conditions have not yet been identified. l EM87-013, 016, 028, 034, 039, 044, 046, 049, 051, 052, ! 054, 056, 057, 059 ET87-009, 011, 012, 014 IC87-004, 027, 029, 032, 034, 044, 048, 060, 062, 065, 066, 069, 070, 076, 077, 081, 082, 092, 098, 100, 112 ATTACHMENT 6.3 F~ ~ C .- . R<v g ff? i s !. ' * - ' *
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, . m s
. ... l_'
.W ,rn 8 a.'
LRI NO: 1254 Rev. 1- (Continusd) o , , 4 .y ( This EARiteview indicates' that"ineasures.'are not in place;.to assure k
.that ' conditions adverse to quality;are being promptly idrintified'and
-l cpi rected. $ n 4W 11 Tne identified- portions ofo the EAR process 'may not meet the.10CFR50,.
Appendix B,-Criterion XVI.which requires measures to be established' to. assure conditions adverse to quality,Lincluding nonconformances, are oromotiv identified and corrected. HQII: The Quality' Department audit of CorNdtive Action, 87-024,- , , ,
. J.identifiedithat significant conditions. adverse to quality >
" are being included in EARS ins':end of being controlled in-
- appropriate corrective. action' mthods and' procedures . '1 (0I 87-024-04). The auditor specifically. identified 7.. EARS iniwhic$ the request identified a'. condition adverse-to ' /%e 2 quality. >The response to'the'open item indicated that the s " '!
EAR procedure will be'reviewe( arid, revised to ensure that- @ EARS are not misapplied.. . Item 1 of this, RI further clarifles the concern by
, identifying.the need to include not.only the request but - 1 also the' resolution of the EAR in determining conditions; adverse.to quality. ' A review of.'all previously written. Eusp i also appears necessary, considering the significant number *' '
of discrepancies identified in both the QA audit and this,RI# { j
-( ,
I
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rh
^
m . 1 l l ATTACHMENT 6.3 ') i l 1 I
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REQUEST FOR INFORMATION (RI) RI NO: 258 SYSTEM CODE: CCW ISSUE DATE: 09-25-87
SUBJECT:
DETERIORATION OF SYSTEM MATERIALS DUE TO HATER CHEMISTRY DEPARTMENT: OPERATIONS COORDINATOR: R. MACIAS TEAM LEADER: M.J. AKINS EOTENTIAL CONCERN /0UESTION: Potential deterioration due to the corrosion of components, equipment, and' piping may have taken place in the CCH system due to inadequate control of water chemistry. . Several factors which substantiate this concern are as follows:
- 1) A conflict exists between the water chemistry limits that are stated in the " System Design Base" (08-5407), and the " Chemistry and Radiochemistry Procedures Manual" (AP.306 III-03).
HQII: No direction was found for selecting between hydrazine and amerizine (each chemical's optimum temperature range for effectiveness is different).
- 2) A SMUD office memorandum, dated 12-22-77 includes several recoramundations for controlling water chemistry. Two of these are:
a) The addition of benotriazol as a copper inhibitor (based on review of surveillance data reports), and
.b) The installation of several system coupon monitoring stations to study deterioration trends.
- 3) A review of the Component Cooling Hater chemistry surveillance data sheets (01-06-86 through 09-21-87, a 79-week period), reveals the following: 1 a) For 57 of the 79 weeks, data was recorded showing the water !
chemistry to be outside designated limits (72% of period). b) For a period of 46 consecutive weeks, water chemistry j adjustments were required. There was no interim testing q recorded to establish the effect of the adjustments (58% of period).
]
yl l I ATTACHMENT 6.3 l l 1
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6:' "
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.n 1,))o 70, .u
- ,; ;L y p;r y
ft . '<"T1RirNO: y 4 #'
, 258' ,Montinued)-
e$s)- (p ; cc) / Fo'r'38of[the79 weeks,the~waterchemistrywasadjustedby ,/fr O the " Feed ~and Bleed"' method (this' accounted for'67% of the )( v .,n, f. . ,.
,j; adju,stments~made.during.the period).
. .y.
w Q .j.' y ' d ). l For 30 of the 79 weeks, an out-of-designated limits for 9 , copper content was recorded (38% of period). 4 n 4) A NHT Corporation letter, dated.03-01-82,-and a procedure for-at ' utilization of the coupon monitoring stations contains instructions ,
, , for. the use 'of various furnished coupons. This procedure has not !
been implemented _or used.- i
- 5) The CCH system includes a coupon monitoring station, but at no time.
were coupons found. installed. Also no individual could be found who had knowledge of any coupons ever being installed. i
- 6) During system drain-down (such as system outages, LLRT, or j maintenance) no consideration was found to have been given to using 4 an inert gas (such as nitrogen, as have been used in other systems i such as RCS, MSS) to reduce corrosion.
l i ed>ms ATTACHMENT 6.3 i l m l
.f A ggy ;,
REQUEST;FOR:INFORMATION (RI)
'RI N0': '268 ' SYSTEM CODE: CCW . ISSUE DATE:- 09-2S-87
SUBJECT:
OVERPRESSURE PROTECTION FOR REACTOR COOLANT PUMPS JACKET COOLERS
' DEPARTMENT: ~NED COORDINATOR: T. TELFORD-TEAM LEADER:- M.J. AKINS.
POTENTIAL CONCERN /00ESTION: Potential-damage"to the Reactor Coolant pumps (P-210A,~B,'C, 0) could occur due to no thermal overpressure protection for the RCP cooling jackets and.its associated CCW piping.
- 1) Vendor Manual-(N6'02 IM01) shows the CCH piped-through the inlet of l the tube-in-tube heat exchanger, through the' 4" relief valve,' then through the cooling jacket to an isolation valve back' to a return header. The 4" relief valve'provided overpressure protection to both coolers.
<2) The'P&ID M543-1 does not show the existing 4" relief valve'for the inlet of,the tube-in-tube heat'exchangers, nor any overpressure protection for the cooling jacket.
'3) A walkdown of the CCW system indicates a modification was made to
' Reactor Coolant pump's P-210A, B, C, and D. The following was found to be common to all RCPs:
CCH water is supplied from a common supply header that has thermal overpressure protection. This includes.all of their outlets, except the cooling' jacket, where there is an isolation valve on both sides of the pump cooling jackets (this is the only section'of CCH piping to the RCPs which i have inlet and outlet isolation valves). The 4" relief j valve is on the inlet side of the pump coolers (a tube-in-tube). ~The jacket coolers have no thermal overpressure protection between isolation valves.
- 4) The vendor's manual was not changed to shown this piping modi fication. i ATTACHMENT 6.3
[s n '
- r-A
.: e y ,
REQUEST FOR INFORMATION (RI) RI-NO:L 269 = SYSTEM CODE: CCW' ISSUE DATE: 09-26-87'
SUBJECT:
RCP SEALS CCW FLOW REQUIREMENTS H DEPARTMENT: SRTP COORDINATOR: J. ITTNER
! TEAM LEADER: M.J. AKINS i
. POTENTIAL CONCERN /00ESTION: l STP.1053'does not appear to provide confidence that a single CCH pump can supply.the'necessary cooling water flow rates to all of the components-served especially the Reactor Coolant Pump Coolers.
The objective of STP.1053 is stated as, "To measure and adjust the Component Cooling Hater (CCH) flowrate through each component to ensure the flowrate is in accordance with engineering requirements." STP.1053.is a multiple step surveillance where the CCH flow rates through individual components are first measured in the "As Found" condition. ; The second step.is to adjust the flows to those components listed in the acceptance criteria. ' All of the flow measurements were taken using' the Contro1otron Flowmeter. , A review of STP.1053 shows some-discrepancies. When the flows through individual components, from Enclosure 9.3, the "As Found" condition flow rates are added'together and then compared to the. total flow through the pump, the result is. a difference in the flow rates of 814.6 gpm. When the same comparison is made with the data for the " balanced" condition in Enclosure 9.4, the difference is 3152.79 gpm. The data sheets-attached to STP.1053 show discrepancies between the CCH ; flow rates obtained using the Controlotron and those recorded via the ' installed CCH flow indicating devices for the RCP pump coolers. The data reads as follows: RCE Controlotron Installed Flow Indicator A- 58.7 gpm 46.45 gpm B 63.4 gpm 51.77 gpm C 68.5 gpm 41.52 gpm D 64.8 gpm 77.04 gpm ATTACHMENT 6.3
y
,o o i ..
RI N0:- 269 (Continued) The original surveillance acceptance criterion for the RC pump coolers was 75. gpm. ..During the surveillance this criterion could not be met even with two CCH pumps running. It was then lowered to 52 gpm without any calculation to justify the change.
.The RCP installation, operation, and maintenance instruction manual states that the normal flow through these coolers shall be 70 gpm.
If the flow rates'-through the Reactor Coolant Pump coolers is as questionable as the above data indicates the ability of the CCH system to deliver the mandatory 70 gpm to the pump jacket coolers required during a loss of. seal injection must also be in doubt. The inability to provide this necessary cooling to the pump jacket coolers during a loss of seal injection could result in a breach of the RCS pressure boundary due.to a failure of the seals. , i i l 1 a i i l i 3 I I l i ATTACHMENT 6.3 ) l i
REQUEST FOR INFORMATION (RI) i RI-NO: 271 SYSTEM CODE: CCW ISSUE DATE: 09-25-87 ;
SUBJECT:
- DESIGN BASES FOR CCH FLOH BALANCE AND TESTING PROCEDURE I DEPARTMENT: NED COORDINATOR: T. TELFORD TEAM LEADER: M.J. AKINS POTENTIAL CONCERN /0UESTION: i The design bases for the CCH flow rates through individual components used in STP.1053 appears to have been derived from an uncontrolled source. The STP (1053) compared the measured CCH flow rates to the expected CCH flow rates. The expected CCH flow rates are the acceptance criteria for , the flow balance test. All expected flow rates through the components l are based on the data presented in calculation Z-CCH-M2079 which is ! referenced in the acceptance criteria of STP.1053.
. Calculation Z-CCH-M2079 tabulated all of the expected flows and tolerances as the CCH system design bases.
. The majority of the expected flow rates in calculation Z-CCH-M2079 were based on data listed in the " Station Manual, System Description, Volume I," Page M-7-5.
The " Station Manual" indicated that the flow data could be used as a guide. The " Station Manual" is an uncontrolled document.
. There are no source references given in the " Station Manual."
Since the " Station Manual" is an uncontrolled document, the acceptance criteria in STP.1053, for the expected flow rates, becomes questionable. Thus the objective of the flow balance, which states, "...to ensure the flow is in accordance with the engineering requirement," may not have been met- because the acceptance criteria (engineering requirement) presently appear to have no credible design basis. t
)
ATTACHMENT 6.3
o O,Is.' , REQUEST FOR INFORMATION (RI) RI NO: 275 SYSTEM CODE: GENERIC . ISSUE DATE: 10-01-87
SUBJECT:
POST MODIFICATION / MAINTENANCE LEAK TESTING OF LOW PRESSURE PIPING-DEPARTMENT: SRTP COORDINATOR: J. ITTNER TEAM LEADER: M.J. AKINS POTENTIAL CONCERN /0UESTION: Portions of low pressure piping may not have been adequately tested,
, following modification / maintenance, per the . requirements or USA Standard Code for Pressure Piping B31.1.
Chapter VI, Section 137 of B31.1 requires a leak test of-low pressure piping prior to operation by bringing.the. piping. system up to desian:
-pressure and inspection for leaks.
_] 1 However, per HR #093071, FCV-40255'was removed from,the-piping system and reinstalled'without testing. Portions of CCH piping identified in hrs j
-#025296, 025297, 025298, 025299, 040407, 045147, 049883, 076082, 062926.. '
076090 and'109919 were leak tested but the test pressure was.not I identified. Portions of.PLS piping, identified in hrs #114040,1140401, ! 114042, 114043,.114044 and 115320 were flushed and possibly leak tested i
-but.the test pressure was not identified.
. Planning personnel indicate they specify that leak tests are to be performed at system oneratino oressure. This may be significantly below design-pressure.
Therefore, it appears that low pressure piping systems are not being leak tested at design pressure as required by 831.1. ATTACHMENT 6.3 _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ _ _ _ - - _ - - _ - - - _ - _ i}}