ML20239A038
| ML20239A038 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 08/18/1987 |
| From: | Croley B, Humenansky D, Prince K SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | |
| Shared Package | |
| ML20238F564 | List: |
| References | |
| NUDOCS 8709170040 | |
| Download: ML20239A038 (32) | |
Text
EXPANDED AUGHENTED SYSTEM REVIEW AND TEST PROGRAM (EXPANDED ASRTP)
EVALUATION OF THE DECAY HEAT REMOVAL l
SYSTEM SUBMITTED BY:
[d d
DATE: 06-/T-M KElTH PRINCE TEAM LEADER CONCURRENCE:
YMid t&> 4 DATE: _.99 Y7 1
[DAVIDHUMENXNSKY (f
EXPANDED ASRTP PROGRAM MANAGER S
CONCURRENCE:
DATE: '
/
BOB CROLEY -
,/
DIRECTOR, NUCLEAR TECHM, CAL SERVICES i
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TABLE OF CONTENTS l
Pace Number
1.0 INTRODUCTION
3 2.0 PURPOSE 4
1 3.0 SCOPE 5
4.0 OVERALL RESULTS AND CONCLUSIONS 6
5.0 SPECIFIC CONCERNS 8
5.1 Acknowledged (Valid) Concerns 5.2 Open (Potential) Concerns 6.0 ATTACHMENTS 9
6.1 List of Documents Reviewed 10 6.2 Status of RIs 12 i
6.3 Detailed Observations - Requests for Information 13 l
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EXPANDED AUGMENTED SYSTEM REVIEW AND TEST PROGRAM EVALUATION OF THE DECAY HEAT REMOVAL SYSTEM
1.0 INTRODUCTION
The Rancho Seco Expanded Augmented System Review and Test Program (ASRTP] evaluation effort involves an assessment of the effectiveness of the System Review and Test Program [SRTP).and an analysis of the adequacy of ongoing programs to ensure that systems will continue to function properly after restart.
The Expanded ASRTP is P. detailed system by system review of the SRTP as implemented on 33 selected systems and an in-deph review of the engineering, modification, maintenance, operations, surveillance, inservice testing, and quality programs.
It also conducts a review, on a sampling basis, of many of the numerous ongoing verification and review programs at Rancho Seco.
Six multi-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 members with consultant, architect engineer and vendor backgrounds are joined with the specific plant knowledge of SMUD team members.
Each 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 dccuments as needed.
Team synergism and communication is emphasizea curing the process in order to enhance the evaluation.
Each team prepares a report for l
each completed selected system evaluated. This report is for the Decay Heat Removal system.
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2.0 PURPOSE The objectives of the Expanded ASRTP evaluation are to (1) assess the adequacy of activities and systerns in support of restart and (2) evaluate the effectiveness of established programs for ensuring safety during plant operation after restart.
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3.0 SCOPE To accomplish the first objective, the Reactor Plant System team evalukted the Decay Heat Removal system to determine whether:
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1.
The system was capable of performing the safety functions required by its design bases.
2.
Testing was adequate to demonstrate that the system l
would perform all of the safety 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 accompitsh the second objective,- 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 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 primary source of leads for the team were the problems identified in the Decay Heat Removal System Status Report. Various source documents such as the USAR and Technical Specifications and available design bases docuraents were reviewed as needed to augment the information needed by the team.
The evaluation of the Decay Heat Removal system included a review of pertinent portions of support systems that must be functional in order for the Decay Heat Removal system to meet its design objectives. 1
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4.0 OVERALL RESULTS AND CONCLUSIONS The more significant issues identified pertaining to the adequacy of the SRTP and the effectiveness of programs to ensure continued safe operations after restart are summarized below.
The summary focuses on the weaknesses identified during the evaluation..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 summ ry or concern refer to the corresponding RIs in Attachment 6.3.
4.1 Summarv of Significant Findinas 4.1.1 Discussion Listed below is a summary of significant findings in the Expanded Augmented System Review and Test program (EASRTP) evaluation of the Decay Heat Removal (DHR) System. Many documents (see Attachment 1) were reviewed during this evaluation.
The initiating document for the evaluation, however, was the System Status Report (SSR).
Thi rty-five (35) problems were identified in the DHR System SSR.
Each of the identified problems were reviewed and pursued until avenues of proue were exhausted or until additional concerns were identified. Only five (5) additional concerns were identified as a result of SSR pursuit which indicates that the SSR was generally effective.
4.2 Concern Assessment No one concern on its merit alone appears to challenge the intended functionality of the DHR System.
There are a numoer of' concerns, newever, wnicn wnen considered :cgether, led the team to consider that the total DHR System may not 1
be as dependable as it should be.
These concerns include:
1 Possibility for system over pressurization Hater hammer problems Reported excessive pipe radial movement Possible to violate successive start limits of DHP Possible vortexing problems when DHP takes suction from BHST Circuitry design affects system reliability Consideration did not lead to determination that DHR System was less than dependable.
It is apparent, however, that the DHR System has problems and potential problems that require immediate attention and follow through to correction.
(RI-23) (RI-52) (RI-67) I
OVERALL RESULTS AND CONCLUSIONS (Continued) 1 4.3
_N_g Jacketino on Emercency Sumo Recire Pioina (Source:
i Halkdown)
System walkdown identified the absence of protective jackets around the Emergency Sump Recirc Piping in the DH Pump rooms. Also, the Emergency Sump Isolat1on valves are jacketed (partially), but are not leak tight.
The as-built design does not conform to B&W Design Basis Document, criginal QA Manual Code application, nor ANSI N-271 Requirements For Containment Isolation.
(RI-18) 4.4 Potential for Water Hammer (Source:
SSR Problem 20)
High point vents are not installed in "A" DHR System.
This prevents system from being properly vented and allows potential for " water hammer" and associated damage upon each system initiation.
Continuing problems with water hammer in a system is indicative of a design deficiency.
(RI-23) 4.5 Testina DHR System SSR does not provide for testing DHR Pump to Pressurizer Auxiliary Spray.
The function is provided in DHR System SSR, Rev.1. Section 2.2.1.4.
(RI-03).
5.0 SPECIFIC CONCERNS A list of the specific concerns the Expanded ASRTP team believes are new concerns not previously identified for resolution follow:
5.1 Acknowledged (valid) Concerns 5.1.1 There is no test identified to determine functionality of pressurizer cooldown with use of pressurizer auxiliary spray I
and DHR pump.
(RI-03) 5.1.2 A review of Nuclear Operations Procedures for the Decay Heat Removal System indicates that some limits for the same operations conflict from procedure to procedure.
(RI-12) l 5.1.3 Plant Cooldown Procedure B.4 requires operation of one Reactor Coolant pump per loop whereas current B&W recommendation is to cooldown with two RCP's in one loop.
(RI-16) 1 5.1.4 The present method for determining Decay Heat Cooler outlet temperature in RCB pressure analysis may be inaccurate.
(RI-17) 5.1.5 Water hammer, with potential for damage, occurs each time "A" Dd Fump is started af ter initial fill because of lack of high point vents in the "A" DH System Discharge Piping.
(RI-23) 5.1.6 Decay Heat System (DHS) relief valves may lift when DHS is placed in service because of instrument inaccuracies.
(RI-67) 5.1.7 The Decay Heat Cooler Bypass line is not oroperly supported ano vibrates excessively at times wnen "S" : rain of JHR System is in operation.
The concern is that the vibration is sufficient to cause fatigue failure.
(RI-77)
Ocen (o tentiali Concerns 5.2 o
5.2.1 The suction line from the Emergency Sump to the DH Pump is not jacketed in the DH Pump Room as required in the original i
Design Bases Document.
(RI-18)
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I 6.0 ATTACHMENTS 6.1 List of Documents Reviewed 6.2 Status of RIs 6.3 Detailed Observations - Requests for Information i
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LIST OF DOCUMENTS REVIEWED 1.
System Design Bases for DHS, NEP 5419, Rev. 1 (draft) 2.
System Status Report "DHS", Rev. 1 3.
USAR Section 1.4, 5.2, 6.2, 9.E.2, 14.2.2 4.
Technical Specifications 3.3, 3.8, 4.5.1, 4.5.3 5.
Licensed Operator Training Program, OD 21 1 0800 6.
Licensed Operator Training Program, OD 21 I 0803 7.
Plant Operating Procedures A.8, DHS, Rev. 30 8.
Plant Operating Procedures B.2, Rev. 38 9.
Plant Operating Procedures B.4, Rev. 40
- 10. Plant Operating Procedures B.6, Rev. 26 11.
Plant Operating Procedures B.9, Rev. 12 12.
Process Standards AP.103, Rev. 12 i
- 13. Casualty Procedure C.12, Rev. 5
- 14. Casualty Procedure CP.101, Rev. 4
- 15. Casualty Procedure AP.103, Rev. 4 l
- 16. Administrative Procedure AP.23 j
- 17. Administrative Procedure AP.42, Rev. 5 l
- 18. Administrative Procedure AP.44 Rev.11 j
- 19. Administrative Procedure AP.48, Original
- 20. Administrative Procedure AP.49, Original 21.
Emergency Operating Procedures
- 22. Maintenance Administrative Procedure MAP.0006 23.
Code of Federal Regulations 10CFR50 Appendix A,B,K,J l
24.
SYSTEMS Training Manual 10, Decay Heat Cooling 25.
SYSTEMS Training Manual 27, Emergency Core Cooling 26.
SYSTEMS Training Manual 35 SFAS 27.
Inservice Inspection Plan ISI 50-312 28.
P&ID H.522, Sh 1 29.
P&ID H.521. Sh 1.2,3 i
30.
P&ID M.5a4 l
31.
P&ID M.545 l
32.
Isometric Drawings:
All for DHS 33.
ECN A-2921 l
A-2931 l
R-0498 34.
Temporary Changes to Operating Procedure A.8, from 01-01-85 to 06-01-87 1
35.
IDADS Computer Points for "DHS" Group Display 36.
IDADS Computer Points for "SFAS" Group Display 37.
IDADS Computer Plant Schematic for DHS Pumps and LPI 38.
Licensee Event Report 86-30 87-38 ATTACHMENT 6.1 - - - _ _ _ _
LIST OF DOCUMENTS REVIEHED (Continued) 39.
Surveillance Procedures:
18 19, Rev. 0 29A 29B 203.05A, Rev. 19 203.058, Rev. 19 203.06A, Rev. 8 l
203.06B, Rev. 10 203.06C/D, Rev. 11 203.09, Rev. 10 203.11, Rev. 7 204.03A, Rev. 20 200.02 204.03B, Rev. 23 213.01, Rev. 9 214.01, Rev 6 214.02, Rev. 2 214.03, Rev. 33 40.
NCR 4449 NCR 6797 NCR S-006 NCR 5-029
- 41. Work Request #134850 H/R 4776 W/R 67 H/R 12552 H/R 15609 H/R 52036 H/R 80511 H/R 105763 W/R 98585 H/R 102132 H/R 102251 H/R 102676 H/R 100299 i
H/R 117875 H/R 129742 H/R 130460 H/R 133457 H/R 114607 l
42.
Rancho Seco Administrative Procedure RSAP-0803 l
43, E,203 Series Drawings 44.
N15.07 Series Drawings i
ATTACHMENT 6.1
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STATUS OF RIs.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 S.O.
RI NUMBER
$"2TUS 02 CLOSED 03 ACKNOWLEDGED 12 ACKNOWLEDGED 16 ACKN0HLEDGED 17 ACKNOWLEDGED 18 OPEN 23 ACKN0HLEDGED 29 CLOSED 38 CLOSED 44 CLOSED 52 CLOSED 54 CLOSED 67 ACKNOWLEDGED 68 CLOSED 72 CLOSED 76 CLOSED 77 ACKNOWLEDGED ATTACMMENT 6.2.
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5.3 DETAILED OBSERVATIONS - RE00EST FOR INFORMATION 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 l
potential concern.
RIs are also used to request.information that the EASRTP team is having difficulty obtaining.
These RIs are considered drafts.throughout the entire evaluation until they become part of the report.
Responsible organizations can accept the potential concern as valid or they may disagree with the potential concern.
If they disagree, they can submit information that convinces the EASRTP team members that the potential concern is l
not valid, or they may redirect the EASRTP members to better focus the concern.
RIs developed during the system evaluation comprise this section of the report.
l ATTACHMENT 6.3 REQUEST FOR INFORMATION (RI)
RI NO:
002 SYSTEM CODE:
DHS ISSUE DATE:
07-22-87
SUBJECT:
TESTING OF BACKUP SPENT FUEL COOLING MODE USING DECAY HEAT PUMP DEPARTHENT:
OPERATIONS COORDINATOR:
RICH HACIAS TEAM LEADER:
KEITH PRINCE POTENTIAL CONCERN /0UESTION:
There is a possibility of meeting the test acceptance criteria without meeting design basis ficw requirement of 3000GPH from a single source (BHST/or spent fuel pool).
SP.203.05A and B, Revision 19, does not require recording the spent fuel pool level prior to and following the recirculation test for verification of suction flow being 3000 GPM from spent fuel pool.
There is a possibility of taking suction from another source which does not meet the intent of the surveillance procedure.
There is no suction flow indication in the DHS for either the BHST or the spent fuel pool.
The flow is indicated only at discharge of DH coolers by FI-26003, FI-26004, FI-26048A, and FI-26049A.
j This RI is closed. A procedure would have to be violated to take suction from more than one source, l
REQUEST FOR INFORMATION (RI)
RI NO:
003 SYSTEM CODE:
DHS ISSUE DATE:
07-18-87
SUBJECT:
DECAY HEAT SYSTEM AUXILIARY SPRAY TESTING DEPARTMENT:
NUC OPERATIONS COORDINATOR:
JOHN ITTNER (SYSTEMS ENGINEER)
TEAM LEADER:
KEITH PRINCE POTENTIAL CONCERN /0UESTION:
The function of auxiliary spray is provided in the DH Removal SSR, Revision 1, Section 2.2.1.4.
However, this alternate method of pressurizer cooldown is not mentioned in any testing section of the SSR.
How this function is verified is not addressed e.g., by taking credit I
from tests or by other means.
This function is not presently addressed in the DHR SSR testing section.
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t REQUEST FOR INFORMATION (RI)
RI NO:
012 SYSTEM CODE:
DHS ISSUE DATE:
07-20-87 SU8 JECT:
DISCREPANCY IN LIMITS AND PRECAUTIONS BETWEEN VARIOUS PROCEDURES DEPARTMENT: OPERATIONS COORDINATOR:
RICH HACIAS TEAM LEADER:
KEITH PRINCE POTENTIAL CONCERN /0UESTION:
J Review of Nuclear Operations Procedures for the Decay Heat System indicates that the limits for the same operation conflict from one procedure to the next.
Some examples are:
PROCEDURE NO. / PARAMETER
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1.
DHS Pump Operation from Emerg. Sump Haximum Flow:
E0P's/3000 gpm AP.103/3500 gpm i
Excessive flow rates on DHS pump j
while lined up to take suction from the emergency sump increases the potential for loss of NPSH and Vortex Formation 2.
Maximum successive starts i
for DH Pump Hotor @ Ambient Temp:
A.8/2 AP.23.10/3 3.
Maximum successive starts for DH Pump motor @ Het:
A.8/
AP.23.10/2 Several successive starts of a large motor can cause stator i
damage from heat buildup caused by hign starting current without suf ficient time for the motor-to Cool.
4 Maximum RCS Pressure on DHS Cooling:
A.8/255 psig AP.103/290 psig When indicated RCS Pressure is 290 psig the pressure at the suction of the D!i Pump is greater than the 300 psig design pressure of the s.uction piping..
REQUEST FOR INFORMATION (RI)
RI NO:
016 SYSTEM CODE:
RCS ISSUE DATE:
07-22-87
SUBJECT:
REACTOR COOLANT PUHP COMBINATIONS DURING COOLDOWN 1
DEPARTHENT:
NUCLEAR OPERATIONS COORDINATOR:
RICH MACIAS TEAM LEADER:
KEITH PRINCE POTENTIAL CONCERN /0UESTIQM:
Plant Cooldown Procedure B.4 requires operation of one Reactor Coolant Pump per loop whereas the current B&W recommendation is to cooldown with two RCPs in one loop. This is because with one pump per loop the pumps are operating near the runout condition where there is risk of cavitation damage, increased vibrations, and inadequate NPSH at low RC pressures.
The operating procedures at Davis-Besse and Crystal River have been revised to go into decay heat cooldown with 2/0 RC pump combination.
i Horeover, there are added limits and precautions to minimize running a l
single RC pump per loop, i.e.,1/1 or 1/0 pume combination (to less than l
i 5 minutes).,
REQUEST FOR INFORMATION (RI)
RI NO:
017 SYSTEM CODE: _ DHS
. ISSUE DATE:
07-30-87
SUBJECT:
DECAY HEAT REMOVAL COOLER CAPABILITIES DEPARTMENT:
SYSTEM ENGINEERING COORDINATOR: __ JOHN ITTNER TEAM LEADER:
KEITH PRINCE j
POTENTIAL CONCERN /00ESTION:
The present method for determining DHC outlet temperature in RCB pressure analysis may be inaccurate.
Design bases document, Section 4.1 (2) under licensing design bases requires each D.H. cooler be designed to cool the ECCS sump water.
The containment building pressure analysis is based on this cooler performance.
The SSR (D.H. Removal System) rev. 1, page 4-15, paragraph 7, requires calculation of heat transfer co-efficient.
It is more appropriate to use the actual cooler outlet water temperature as direct verification of the cooler performance and document the validity of the Decay Heat Removal Cooler Characteristics.
This could replace the analytical method which may not be accurate due to possible crud build up, blockage or degradation of the cooler performance.
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r REQUEST FOR INFORMATION (RI)
A RI NO:
018 SYSTEM CODE:
DHS ISSUE DATE:
07-22-87
SUBJECT:
JACKETING AROUND REACTOR BUILDING SUMP ISOLATION VALVES DEPARTMENT:
NUCLEAR ENGINEERING COORDINATOR:
RON LAWRENCE TEAM LEADER:
KEITH PRINCE POTENTIAL CONCERN /00ESTION:
Decay heat system design bases document Section 4.1(8) page 13 of 49 under Licensing Design Bases states, "A jacket shall be installed at the exterior of the Reactor Building.
Enclosed piping up to and including the stop valve in each DHS suction line from the Reactor building."
The USAR, Section 5.2.4, page 5.2-45 under Reactor Building Isolation states:
l "Each of the two emergency sump recirculation lines has only one isolation valve, encased in a secondary housing, and located outside the Reactor Building.
These valves are required to open under certain emergency conditions and can fail in two ways:
(a) the valve body ruptures or leaks and-(b) the valve fails to open.
The housing takes care of the failure under (a) and 100 percent redundancy, provided by two recirculation lines, takes care of the failure under (b)."
This requirement was provided by B&H in the 18K1 manual " Duke type PWR Nuclear Steam System" issued 05/15/68.
The jacket is orovidtd to prevent drainage of the reactor building emergency sump in the event of failure of the stop valve body or creakage of the suction line between the Reactor building penetration and the valves.
During walkdown of the DH suction line, it was noted that the jacket around the valve body (HV-26105) is not leak tight. There is no jacket around the piping from valve HV-26105 to the penetration.
Hithout a proper designed jacket around the piping and the valve, there is a possibility of draining the entire emergency sump inventory during moderate line crack to the auxiliary building decay heat system and loss of EPSH to the redundant decay he&t pump.
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REQUEST FOR INFORMATION (RI)
RI NO:
023 SYSTEM CODE:
DHS ISSUE DATE:
07-22-87
SUBJECT:
DECAY. HEAT SYSTEM LICENSING DESIGN BASES DEPARTHENT:
NUCLEAR ENGINEERING COORDINATOR:
R. LAWRENCE TEAM LEADER:
KEITH PRINCE POTENTIAL CONCERN /00ESTION:
Each time the "A" Decay Heat System is started, water hammer occurs and creates potential for serious damage to the system.
The water hammer occurs because the "A" train does not have high point vants to allow proper filling and venting.
The Design Bases document, Section 4.1 under Licensing Design Bases (Paragraph 13) requir s the DHS to be completely filled so that venting is not required at initiation.
This criteria is from " Duke type PHR Nuclear Steam System" 18K1 manual issue 05-15-68.
The "B" DHR train contains adequate high point vents and drains.
REQUEST FOR INFORMATION (RI)
RI NO:
029 SYSTEM CODE:
GENERIC ISSUE DATE:
07-23-07
SUBJECT:
COMMERCIAL GRADE PROCUREMENT HETHOD DEPARTMENT:
NUCLEAR ENGINEERING COORDINATOR:
R. LAWRENCE TEAM LEADER:
KEITH PRINCE POTENTIAL CON.CERN/0UESTION:
4 The commercial dedication process was reviewed with Procurement Engineering and the following potential problems are noted:
There is no commercial vendor's list.
Vendor catalogs are not maintained.
i No form, fit, and function evaluation is done on commercial parts.
No piece parts "Q" list exists.
The Procurement Engineering Group relies on the vendor to identify which parts are supplied with 10CFR21 applicability and which parts are provided commercial grade. There is no documentation which identifies evaluation of l
parts to their critical attributes (en engineering function).
Ne sampling plan, testing and inspection requirements, acceptance criteria nor review of bolt pattern, weight, orientation, etc. is performed. Commercial vendor's information is used to maintain equipment qualification.
Conditioning, special processes are not reviewed, nor documented.
l Source inspections cf commercial veaders are not performed by Nuclear Engineering.
No training of personnel on commercial dedication.
Nuclear Engineering needs to develop a piece parts 0-list identifying all parts, their procurement levels and inspection / acceptance criteria.
1 This RI is closed.
There is insufficient objective evidence to support an RI at this time. However, it has been turned over to the j
EASRTP team to evaluate as a potential generic issue.
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REQUEST FOR INFORMATION (RI) i RI NO:
038 SYSTEM CODE:
GENERIC ISSUE DATE:
07-27-87
SUBJECT:
CONTROL OF VENDOR TECHNICAL MANUALS j
i DEPARTMENT:
ADMINISTRATIVE COORDINATOR:
R. LAWRENCE TEAM LEADER:
KEITH PRINCE l
I POTENTIAL CONCERN /0UESTION:
j Revision 4 of AP.46 dated 03-30-87 " Control of Vendor Technical Manuals" l
is not being fully implemented; training to AP.46 requirements has not i
been provided to appropriate personnel.
(e.g. NEDC)
The responsibilities imposed upon Nuclear Engineering Document Control (NEDC) as defined in Section 6.0 of AP.46 are actually being implemented by personnel in the Technical Library..1 of AP.46 " Technical Manual Review Sheet" is not being utilized as required by Section 6.2.
Although, it should be noted i
that the review sheet currently used, requires essentially the same information as Enclosure 8.1.
1 Distribution of technical manuals for review and approval (Ref. 6.3) is not by NEDC, but by the Technical Library.
j NEDC personnel, when queried of the requirements of AP.46 indicated that they have no knowledge of its contents.
Similarly, Nuclear l
Procurement was not aware of the requirements imposeo upon them (Ref. 6.10).
It was indicated that they had no knowledge of the issuance of this AP.
How does the Technical Library determine distribution of " approved manuals" (Ref. 6.5.3)?
(Plant has not responded.)
Have all Engineering / Design personnel (as applicable). or other personnel who would receive vendor technical documents, been made aware of AP.46 requirements to ensure vendor technical manuals are being updated? (Plant has not responded.)
Why doesn't AP.46 make reference to AP.42 " Maintenance Information l
l Management System" (HIMS) procedure? Contained in the MIMs program is the Master Equipment List (MEL).
The MEL, in part, provides reference to Manufacturer's Instruction Book No.'s (Ref. AP.42, Rev.
5, page 40). Who is responsible for ensuring that Manufacturer's (vendor's) instructions, either new or revised, are incorporated into the MEL? Shouldn't AP.46 address who has this responsibility etc.?
(Plant has not responded.)
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RI-038 (Continued)
AP.46 does not address cross referencing of vendor manuals where information is contained in one or more than one manual or where one manual covers multiple equipment.
This RI is closed.
There is insufficient objective evidence to support an RI at this time.
However, it has been turned over to the EASRTP team to evaluate as a potential generic issue.
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REQUEST FOR INFORMATION (RI)
RI NO:
044 SYSTEM CODE:
DHS ISSUE DATE:
07-27-87
SUBJECT:
FLOODING IN DECAY HEAT PUMP ROOMS DEPARTMENT:
NUCLEAR ENGINEERING COORDINATOR:
R. LAWRENCE l
TEAM LEADER:
KEITH PRINCE POTENTIAL CONCERN /0UESTION:
USAR Table 14.1-17, Item 22, indicates that a barrier is provided between the DH pump rooms to stop room-to-room flooding.
Walkdown of the DH pump rooms determined that the DH pump room "A has a dimension of approximately 16 ft. x 60 ft.
Thcre is a fire door at -39' elevation. As such, approximately 8000 CFT of water can be contained in the 'A' pump room before water spills over to the 'B' pump room.
There i
are apparently no water tight doors between decay heat pump rooms, rather, fire doors have been provided.
It appears that a potential deficiency exists in water tight separation between decay heat pump rooms as addressed in the USAR.
Flooding is isolated by operator action after receipt of DH pump room level alarms.
These levels switches LSH-66407 and LSH-66311 are located on the walls of DH pump rooms, 1 ft. above the floor.
Neither the level switches nor the sump pumps are, safety related or EQ qualified.
The level switches are powered from the 'E' bus which is non-Q.
There is a possibility of flooding both DH pump rooms before operator action is taken to mitigate it, as credit cannot be taken for non-Q equipment and the level switches may be damaged due to harsh environment.
There is no level indication other than annunciation alarm in the control room.
I This RI is closed. After much di':cussion, the team agrees that the probability of water entering 2nd Decay Heat Pump Room in pipe break event is remote.
The team continues to pursue information on the use of non-EQ components in potentially harsh environments.
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REQUEST FOR INFORMATION (RI) l RI NO:
052 SYSTEM CODE:
DHS ISSUE DATE:
07-27-87
SUBJECT:
DECAY HEAT PUMP TRIP / RESTART LOGIC l
l DEPARTHENT:
TRAINING COORDINATOR:
PAUL TURNER TEAM LEADER:
KEITH PRINCE POTENTIAL CONCERN /0UESTION:
It is possible to exceed successive start limits on the Decay Heat R;mps by attempting to start the pump with 62/TD0 closed, not realizing the
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breaker closed when the start button was pushed.
Training materials for DHS state that the decay heat pumps cannot be restarted for 3 miitutes if either of the suction valves (HV-20001 or HV-20002) leaves its open seat.
Per drawing E-203, Sheet 3, it appears that the pump can be restarted as soon as both suction valves are full open regardless of the 3 minute time delay because the 62A/ INST contact would open de-energizing the trip coil.
Also, pushing the start button will start the pump whether the trip relay is initially energized or not.
If this is done after 62A energizes and before 62A/TD0 opens, the pump breaker will retrip when the start button is released.
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This RI is closed. Training has revised the Training Manual in question.
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1 REQUEST FOR INFORMATION (RI)
RI NO:
054 SYSTEM CODE:
DHS ISSUE DATE:
07-31-87
SUBJECT:
DHS SUCTION VALVE CIRCUITRY AFFECTS SYSTEM RELIABILITY DEPARTHENT:
NUCLEAR ENGINEERING COORDINATOR:
RON LAWRENCE 1
TEAM LEADER:
KEITH FRIt(CE POTENTIAL CONCERN /00ESTION:
j The auto-close feature of Decay Heat Suction Valves HV-20001 and HV.20002 is a source of system unreliability. Closure of either valve renders
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both DHS trains inoperable.
Several reportable occurrences resulted from inadvertent closure of the SE valves (LERs 82-15, 86-16, 86-24, 86-30, and 87-38).
The causes of the events were loss of SFAS A or B power; loss and subsequent restoration of power to HV-2000); and a lifted lead on the core flood tatnk valve HV-26514 interlock during maintenance on the CFT valve.
Refer to elementary diagram E-203, sheets 600 and 60E for the following concerns:
The way the circuitry is designed, the valves will auto-close on restoration of control power.
The corrective action section of LER 86-30 says:
"A review will be performed concerning the necessity for auto-close interlocks (re-energizing the valve motor's MCC bus and having the valve remain in its last position) on HV-20001 and HV-20002 in light of the AE00 report on decay removal problems by July 6, 1987 (Ref. AE00/C503 " Decay Heat Removal Problems at U.S. Pressurized Water Reactor's dated December 1985).
Any appropriate design changes deemed necessary oy this review will 'oe installed prior to the end of the next refueling outage."
What improvements are planned to solve this problem?
(Plant has responded with plans to evaluate if design change l
required.)
The amber light on the breaker cubicle is wired such that it will be lit if the pressure interlocks are met, but is not affected by the status of the CFT valw interlock. This is not consistent with a good human factors approach. The light, if it is to be useful, should indicate that all applicable interlocks have been met.
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The 62A/TDC contact that was installed by ECN A-2487 to override the l
I pressure interlocks after a loss of SFAS A or B power also overrides the CFT valve interlock.
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l RI-054 (Continued)
Other B&W plants do not have the CFT valve interlock.
Is this interlock required?
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What is the basis for having an open-enable pushbutton at the breaker cubicle?-
Can the auto-close feature be eliminated entirely by providing an
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e alternative means of overpressure protection?
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1 (Plant responded that installation of other means of protection such as larger relief valves allows other B&W plants not to have this interlock.)
This RI is closed.
This issue was identified prior to the EASRTP Evaluation.
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REQUEST FOR INFORMATION (RI) j RI NO:
067 SYSTEM CODE:
DHS ISSUE DATE:
07-29-87
SUBJECT:
LOW RANGE AND WIDE RANGE PRESSURE INSTRUMENT LOOPS DEPARTMENT:.___ NUCLEAR ENGINEERING COORDINATOR:
R. LAWRENCE and NUCLEAR OPERATIONS R. MACIAS TEAM LEADER:
KEITH PRINCE I
I POTENTIAL CONCERN /0UESTION:
A potential system pressure error could cause DHR system relief valves to lift allowing RCS coolant to spill in Auxiliary Building.
The low range pressure indicator, PI-21261, provides RC pressure indication in the control room for decay heat system initiation.
The pressure indicator loop has an accuracy of : 10 psi.
The wide range pressure transmitters PT-21092 and 21099 are used for interlock permissiveness of DH dropline valves HV-20001 and 20002. These pressure
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transmitter loops have an accuracy of about = 25 psi. As such, there can be a difference of indicated RC pressure vs valve interlock pressure up to as much as 35 psi.
This means the decay heat system may not be i
operable up to about 215 psi (250-35) which is less than RC pump NPSH requirement for 1/0 RC pump operation, (Process Standard Curve
.AP.101-28). DH System may be put in service when the RCS pressure is at about 2B5 psig (with potential 35 psig error) which violates the design pressure criteria of DH piping and equipment.
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I REQUEST FOR INFORMATION (RI)
RI NO:
068 SYSTEM CODE:
DHS ISSUE DATE:
07-28-87
SUBJECT:
CARBON STEEL BOLTING MATERIAL DEPARTMENT:
NUCLEAR ENGINEERING COORDINATOR:
R. LAWRENCE TEAM LEADER:
KEITH PRINCE POTENTIAL CONCERN /00ESIlQN:
Carbon steel bolting materials are used in components that carry borated j
water. At least 3 NRC IENs, #80-27, 82-02, and recently, IEN 86-108, were issued concerning corrosion of carbon steel materials due to boric acid leakage.
Small leakages over long periods of time can cause severe degradation causing fastener failure and breaching of the pressure boundary.
The problem was recognized at SHUD about 1979 and ECNs A-2921 and A-2931 were issued to replace the carbon steel bolting on some Velan and Anchor valves, respectively, with Grade 630 stainless.
However, not all affected valves received the replacement bolting.
In the decay heat system, bolting was replaced o,. the Anchor valves in safety-related service, but not on Velan valves.
This concern is applicable to RCS, SIM, PLS, CBS, BHS, SFC, RCO a7d RHS.
All of these systems have valves with A-193 Grade B carbon. steel bolting.
This RI is closed.
This issue was identified prior to the EASRTP Evaluation.
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REQUEST FOR,INFORMATION (RI)
RI NO:
072 SYSTEM CODE:
DHS ISSUE DATE:
07-29-87
SUBJECT:
TESTING OF HPI PIGGYBACK MODE DEPARTMENT:
$YSTEMS ENGINEERING COORDINATOR:
R. LAWRENCE I
TEAM LEADER:
KEITH PRINCE POTENTIAL CONCERN /0VESTION:
Testing of the decay heat cooler to HPI suction line (piggyback mode) did not cover the case of an HPI pump and makeup pump operating in parallel.
l The original test, TP 203-4, demonstrated the capability to deliver 528 gpm to the HPI pump suction with sufficient HPSH available to the pump.
However, with both an HPI pump and makeup pump operating, it is
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possible to get nearly twice as much flow and still be within the guidelines of the E0Ps (Rule 2).
It appears from the margin available in the original test that the capability exists, however, it has not been demonstrated by a test or calculation.
This RI is closed.
It will be re-issued with the SIM system, as the issue concerns SIM function..
REQUEST FOR INFORMATION (RI)
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RI NO: __ 076 SYSTEM CODE:
DHS ISSUE DATE:
07-29-87 1
SUBJECT:
VORTEX FORMING ON NPSH FOR DH PUMPS l
DEPARTMENT:
NUCLEAR ENGINEERING COORDINATOR:
R. LAHRENCE/3849 TEAM LEADER:
KEITH PRINCE 1
POTENTIAL CONCERN /0VESTION:
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The problem of vortexing when DH pumps take suction from the Emergency Sump has been addressed.
Similar analysis was not found, however, for DH Pumps taking suction from BHST at low levels.
There is no assurance that vortex formation will not occur when both DH Pumps operate simultaneously with Containment Spray Pumps prior to switch over to Emergency Sump suction. Also, there is no mention of importance of BHST level i
i instrumentation accuracy and alarm indication used for operator action for switch over, j
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I This RI is closed.
This issue was identified prior to the EASRTP Evaluation.
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REQUEST FOR INFORMATION (RI) l J
l RI NO: 077 Rev. 1 SYSTEM CODE:
DHS ISSUE DATE:
08-19-87 l
SUBJECT:
VIBRATION ON DECAY HEAT PIPING 1
DEPARTMENT:
NUCLEAR ENGINEERING COORDINATOR:
R. LAWRENCE /3849 j
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TEAM LEADER:
KEITH PRINCE j
POTENTIAL CONCERN /00ESTION:
l The "B" Decay Heat Cooler Bypass Line is not properly supported and vibrates excessively when "B" train of DHS is in operation.
This vibration causes concern of potential fatigue failure.
Because of the magnitude of radial movement observed, the team believes that results of additional analysis would warrant reconsideration of corrective action priori ty.
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The Decay Heat Ccaler 6" Bypass piping is not supported by a hanger or under support.
The Bypass piping is supported only by sections of adjacent 10" pipe and has been observed to vibrate excessively at times.
The team is aware that the vibration hac been previously reported and that an evaluation was made by Engineering which resulted in recommendation that pipe supports be installed at a later time, "for system betterment".
The team strongly feels that the vibrating bypass piping continues to be a concern because:
Evaluation of the vibrating Bypass piping made in 1986 concluded that the vibration was acceptable based on a measured 1/8", zero to peak displacement (conservative).
(Calculation Z-DHS-M0370) i Recent observed vibration is mucn greater than 1/8" and has been estimated by several operators to be greater than 2" displacement, l
peak to peak.
The throttle and temperature control MOV is installed in the Sypass Piping. Does position of this valve significantly affect the amount of vibration in the DHC Bypass line?
Heasured vibration of the DHC Bypass during evaluation in 1986 was 1/8" and recently observed vibration is much greater.
Is the vibration problem worse than when evaluated?
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