ML20237J387
| ML20237J387 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 08/04/1987 |
| From: | Croley B, Humenansky D, Prince K SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | |
| Shared Package | |
| ML20237J375 | List: |
| References | |
| NUDOCS 8708180156 | |
| Download: ML20237J387 (31) | |
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h EXPANDED AUGMENTED SYSTEM REVIEH AND. TEST PROGRAM (EXPANDED ASRTP) 3 1 I EVALUATION OF THE DECAY HEAT REMOVAL
. SYSTEM
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SUBMITTED BY: /1- ~ M DATE: OB -c5 -87 EITH PRINCE
'[TEAMLEADER CONCURRENCE: wiA w [ DATE: "3'67 '
/ DAVID HUMENANSKY j.
[ EXPANDED ASRTP PROGRAM MANAGER CONCURRENCE: / DATE:
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/ B08 'CROLEY \
DIRECTOR, NUCLEAR CHNICAL SERVICES _1_ khB180156870810 g ADOCK 05000312 PDR k -_ _ - - -- . - . - - . _ _ _ _ _ ~ _ - -
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'i 4 TABLE OF CONTENTS Pace Number-
1.0 INTRODUCTION
3 2.0 PURPOSE 4
;3.0 SCOPE 5 ,
- 4.0 '0VERALL RESULTS AND CONCLUSIONS ~ 6 r.
5.0 DETAILED OBSERVATIONS - REQUESTS FOR INFORMATION 9 6.0 ATTACHMENTS' 28 6.1 List of Documents Reviewed 29 6.2 Status of RIs 31 y
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, EXPANDED AUGMENTED SYSTEM REVIEH AND TEST PROGRAM
. EVALUATION OF THE DECAY HEAT REMOVAL . STEM l
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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 a detailed system by system review of the SRTP as implemented on 33 selected systems and an in-depth 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.
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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 documents as needed. Team synergism and communication is emphasized during the process in order to enhance the evaluation. Each team prepares a final report for each completed selected system evaluated. This report is for the Decay Heat Removal system. a
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-... :2.0 ~ PURPOSE' The objectives of the Expanded ASRTP evaluation'are'to (1) atsess the adequacy of activities and systems. 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 evaluated the Decay Heat Removal system to determine whether:
- 1. The systein was capable of performing the safety functions required by its design bases.
- 2. Testing was adequate to demonstrate that the system 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 systen.
- 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 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. Cuality Assurance
- 6. Engineerin;; 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 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 i available design bases documents 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. _ _ . _ . - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _a
V 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 fccuses on the weaknesses identified during the evaluation. Section 0.0 provides detailed findings by providing the Request for Information l (RI) forms that are used by the Expanded ASRTP teams to identify potential concerns during the evaluation. The numbers in brackets after each individual summary refer to the corresponding RIs in Section 5.0. 4.1 Summarv of Significant Findinas 4.1.1 Discussion Listed below is a summary of significant findings in t he Expanded Augmented System Review and Test program (EASRTP) evaluation of the Decay Heat Removal (DHR) System. Ma ny documents (see Attachment 1) were reviegi during this evaluation. The initiating document for the evaluaticn, however, was the System Status Report (SSR). Thirty-five (35) problems were identified in the DHR System SSR. Each of the identified problems were reviewed and pursued until avenues of probe were exhausted or until additional co1cerns were identified. Only five (5) additional concerns we e identified as a result of SSR pursuit which indichtes 1: hat the SSR was generally effective.
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4.2 Concern Assessment No one concern on its merit alone appears to challenge the intended functionality of the DHR System. There are a number of concerns, however, which when considered toge:her, led the team to consider that the total DHR System may not be as dependable as it should be. These concerns include:
- Possibility for system overpressurization
. 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 Sys tem 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-54) (RI-67) (RI-76) i l j
t OVERALL RESULTS AND CONCLUSIONS (Continued) 4.3 No Jacketina on Emeraency Sumo Recirc Pioina (Source: Halkdown) ; System walkdown identified the absence of protective jackets around the Emergency Sump Recirc Piping in the DH Pump rooms. Also, the Emergency Sump Isolation valves are jacketed (partially), but are not leak tight. The as-built design does not conform to B&W Design Basis Document, original QA Hanual 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 and a significant safety issue. (RI-23) 4.5 Egtential for Floodina DH Pumo Rooms (Source: USAR) DH Pump rooms are not isolated by a water tight door as stated in USAR. In case of line break, DH Pump room will g contain approximately 60,000 gallons of water before spilling into "B" DH Pump room. Action to mitigate consequences of a pipe break must be taken after receipt of level alarm indication. Level switches and sump pumps are neither Safety-Related nor EQ Qualified. Level indicators , are powered by Non-Q "E" Bus. (RI-44) 4.6 Testina and Calculations 4.6.1 The team identified several tests or calculations required, but not in place, to determine system capability at specified conditions or to ensure system operability.
. SP.203.05A and B, as written, do not ensure that suction for the test comes only from the SF Pool.
(Single Source) (RI-02) l l
- DHR System SSR does not provide for testing DHR Pump to Pressurizer Auxiliary Spray. The function is provided l in DHR System SSR, Rev. 1. Section 2.2.1.4. (RI-03)
- TedtingDHPumptoHPI(piggybackmode)doesnotcover HPI and Makeup Pump running in parallel. Need to determine adequate NPSH available. (RI-72) l i ,
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. OVERALL RESULTS AND CONCLUSIONS (Continued) 1 4.6 (Continued) l l
. Calculations are required (currently being performed) l to determine if vortexing could occur when DHR Pump !
takes suction from BHST at low levels. Calculations ! are available for same potential problem with Emergency ! Sump. (RI-76) ) 4.7 Generic Concern 4.7.1 Velan valves in the DHR System were found to contain carbon steel bolts and are subject.to corrosion and failure. . Moreover, many valves in the following systems are known to have carbon steel bolts and are subject to the same boron corrosion failure: RCS PLS BHS RCD SIM CBS SFC -RHS The carbon steel / boric acid corrosion problem was recognized at Rancho Seco as early as 1979. Since that time, ECNs A-2921 and A-2931 were issued to replace carbon steel bolts with 630 S/S in a number of Anchor Darling and Velan valves. DrlR System Anchor valves had bolting replaced, Velan valves in DHR System had no replacement. A brief walkdown of the affected systems will indicate the magnitude of the problem. (RI-68) \ i l L - __-------__--------------_----------_m
..- 5.0 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 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 final report. Responsible organizations can accept the potential concern as. valid or they may-disagree with the potential concern. lIf they disagree,.they can submit.information that convinces the;EASRTP team members that the potential. concern is 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. Attachment 2 of the report 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. Approximately one week will tu! provided after the report is. issued to provide time for departments to address each RI for validity. A j revision to Attachment 2 will then be issued to reflect the status of RI's. All RIs not acknowledged at the end of this period will have.an "Open" status. RIs are then transferred into the Restart Scope List tracking system for resolution and corrective action implementation. I a
i-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 DEPARTMENT: OPERATIONS COORDINATOR: RICH MACIAS TEAM MEMBER: DS. DENNIS MARTIN. TEAM LEADER: KEITH PRINCE S. CARMICHAEL i POTENTIAL' CONCERN /0UESTION: There is a possibility of. meiting the test acceptance criteria without meeting design basis flow requirement of 3000PM 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 cool rs by FI-26003, FI-26004, FI-26048A, and FI-26049A.
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, REQUEST FOR INFORMATION (RI) {
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-RI N0: 003 SYSTEM CODE: DHS ISSUE DATE:. 07-18-87 -{
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SUBJECT:
DECAY HEAT SYSTEM AUXILIARY SPRAY TESTING DEPARTHENT: NUC OPERATIONS C0ORDINATOR: JOHN ITTNER ; (SYSTEMS ENGINEER) TEAM MEMBER: D. SATPATHY TEAM LEADER: KEITH PRINCE l S. CARMICHAEL POTENTIAL CONCERN /0UESTION: The fundtion 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 from tests or by other means. This function is not presently addressed in the DHR SSR testing section. 1
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.. REQUEST FOR INFORMATION (RI)
RI N0: 012 SYSTEM CODE: DHS ISSUE DATE: 07-20-87
SUBJECT:
DISCREPANCY IN LIMITS AND PRECAUTIONS BETHEEN VARIOUS PROCEDURES DEPARTMENT: OPERATIONS COORDINATOR: RICH MACIAS TEAM MEMBER: D. MARTIN ~ TEAM LEADER: KEITH PRINCE i l l POTENTIAL CONCERN /00ESTION: 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
- 1. DHS Pump Operation from Emerg. Sump Maximum Flow: E0P's/3000 gpm AP.103/3500 gpm
- Excessive flow rates on DHS pump while lined up to take suction from the emergency sump increases the potential for loss of NPSH g and Vortex Formation
- 2. Maximum successive starts \
for DH Pump Motor @ Ambient Temp: A.8/2 AP.23.10/3
- 3. Maximum successive starts for DH Pump motor @ Hot: A.8/ AP.23.10/2 Several successive starts of a large motor can cause stator damage from heat buildup caused by high starting current without sufficient 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 DH Pump is greater than the 300 psig' design; pressure of the suction piping. '
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~ , REQUEST FOR INFORMATION (RI)
RI NO: 016 SYSTEM CODE: RCS ISSUE DATE: 07-22-87
SUBJECT:
REACTOR COOLANT PUMP COMBINATIONS DURING COOLDOWN DEPARTHENT: NUCLEAR OPERATIONS COORDINATOR: RICH MACIAS : TEAM HEMBER: D. SATPATHY TEAM LEADER: KEITH PRINCE i S. CARMICHAEl~ POTENTIAL CONCERN /0VESTION: Plant Cooldown Procedure B.4 calls for operation of one Reactor Coolant Pump per loop. 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. j Horeover, there are added limits and precautions to minimize running a ! single RC pump per loop, i.e.,1/1 or 1/0 pump combination (to less than i 5 minutes).
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r REQUEST FOR INFORMATION (RI) RI NO: _ __017' SYSTEM CODE: DHS ISSUE DATE: 07-30-87 SttBJECT: DECAY HEAT REMOVAL COOLER CAPABILITIES DEPARTHENT: SYSTEM ENGINEERING COORDINATOR: JOHN ITTNER TEAM MEMBER: D. SATPATHY S. CARMICHAEL TEAM LEADER: KEITH PRINCE POTENTIAL CONCERN /00ESTION: Design bases document, Section 4.;l (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 (0.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. k
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'. REQUEST FOR'INFORMATION (RI)
~RI NO: 018 SYSTEM CODE: DHS ISSUE DATE: 07-22-87
SUBJECT:
JACKETING AROUND REACTOR BUILDING SUHP ISOLATION VALVES DEPARTHE!!T: NUCLEAR ENGINEERING COORDINATOR: RON LAWRENCE TEAM HEMBER: S. CARMICHAEL TEAM LEADER: KEITH PRINCE POTENTIAL CONCERN /0UESTION: Decay heat system design bases document Section 4.1(8) page 13 of 49; under Licensing Design Bases' states, "A jacket s .(11 be installed at the , exterior of the Reactor Building. Enclosed piping up to and including l 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: . 1 "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&W in the 18K1 manual " Duke type PHR Nuclear Steam System" issued 05/15/68. The jacket is provided to prevent drainage of the reactor building ' emergency sump in the event of failure of the stop valve body or breakage of the suction line between the Reactor building penetration and the valves. During walkdown of the LH suction line, it was noted that the jacket around the valve body (HV-26105) is not leak tight. There is no jacket l around the piping from valve HV-26105 to the penetration. Without 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 ESH to the redundant decay heat pump.
3 REQUEST FOR INFORMATION (RI) RI NO: 023 SYSTEM CODE: DHS ISSUE DATE: 07-22-87
SUBJECT:
DECAL HEAT SYSTEM LICENSING DESIGN BASES DEPARTMDIT: NUCLEAR ENGINEERING C0ORDINATOR: R. LAWRENCE TEAM MEMBER: D. SATPATHY TEAM LEADER: KEITH PRINCE DENNIS MARTIN POTENTIAL CONCERN /00ESTION: The Design Bases document, Section 4.1 under Licensing Design Bases. (Paragraph 13) requires the DHS to be completely filled at initiation so that venting is not required. This criteria is from " Duke type PHR Nuclear Steam System" 18K1 manual issue 05-15-68. However, no means has been provided to keep the DH system filled with water. DH pump "A" discharge piping does not have high point vents and adequate drain locations to vent the system properly. .
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l REQUEST FOR INFORMATION (RI) RI NO: 038 SYSTEM CODE: GENERIC ISSUE DATE: 07-27-87
SUBJECT:
CONTROL OF VENDOR TECHNICAL MANUALS DEPARTMENT: ADMINISTRATIVE COORDINATOR: HARRY ELLER TEAM MEMBER: D. A. LOGAN TEAM LEADER: KEITH PRINCE D. SATPATHY EDTENTIAL CONCERN /0UESTION: - Revision 4 of AP.46 dated 3-30-87 " Control of Vendor Technical Manuals" is not being fully implemented; training to AP.46 requirements has not 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.
. Enclosure 8.1 of AP.46 " Technical Manual Review Sheet" is not being utilized as required by Section 6.2. Although, it should be noted that the review sheet currently used, requires essentially the same '
information as Enclosure 8.1. j
- Distribution of technical manuals for review and approval (Ref. 6.3) is not by NEDC, but by the Technical Library.
. NEDC personnel, when queried of the requirements of AP.46 indicated that they have no knowledge of its contents. Similarly, Nuclear Procurement was not aware of the requirements imposed 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.)
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. RI-038- (Continue'd)
. Why doesn't AP.46 make reference to AP.42 " Maintenance Information !
Management System" (MIMS) procedure? Contained in the MIMs program ) is the Master Equipment. List (MEL). The MEL, in part, provides 1 reference to Manufacturer's Instruction Book No.'s (Ref. AP.42, I Rev. 5, pg. 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.)- l
. 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.
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. REQUEST FOR INFORMATION (RI)
RI NO: 044' SYSTEM CODE: DHS ISSUE DATE: 07-27-87 1
SUBJECT:
FLOODING IN DECAY HEAT PUMP ROOMS DEPARTMENT: NUCLEAR ENGINEERING COORDINATOR: R. LAHRENCE TEAM MEMBER: D. MAD. TIN. S. CARMICHAEL TEAM LEADER: KEITH PRINCE DILLAP-SATPATHY POTENTIAL CONCERN /00ESTION: USAR' Table 14.l'-17. Item 22, indicates that a barrier is provided between the'DH pump rooms toistop room-to-room floodir.g. Walkdown.of.the DH pump rooms determined that the DH pump room "A" has a-dimension of'approximately 16 ft. x 60 ft. There 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 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.
I FloodingLis 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. l [ t j
REQUEST FOR INFORMATION (RI) RI NO: 052 SYSTEM CODE: DHS ISSUE DATE: 07-27-87
SUBJECT:
DECAY HEAT PUMP TRIP / RESTART LOGIC
. DEPARTMENT: TRAINING COORDINATOR: _. PAUL TURNER TEAM MEMBER: TOM FAUBLE TEAM LEADER: KEITH PRINCE POTENTIAL CONCERN /00ESTION:
An Operator could violate successive start limits on the Decay Heat Pumps. by attempting to start the pump with 62/TD0 closed, not realizing the breaker closed when the start button was pushed. Training materials for DHS state that the decay heat pumps cannot be l restarted for 3 minutes if either of the suction valves (HV-20001 or HV-20002) leaves its open seat. Per drawing E-203, Sheet 3, it appears i 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 errergizes and before 62A/TD0 opens, the pump breaker will retrip when the st rt button is released.
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i . REQUEST FOR INFORMATION (RI) RI NO: 054 SYSTEM CODE: DHS ISSUE DATE: 07-31-87
SUBJECT:
DHS SUCTION VALVE CIRCUITRY AFFECTS SYSTEM RELIABILITY NUCLEAR ENGINEERING COORDINATOR: RON LAWRENCE ! DEPARTHENT: TEAM HEMBER: THOMAS FAUBLE TEAM LEADER: KEITH PRINCE POTENTIAL CONCERN /00ESTION: 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 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-20001; and a lifted lead on the core flood tank valve HV-26514 interlock during maintenance on the CFT valve. Refer to elementary diagrams E203 sheets 600, 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: g "A review will be performed concerning the necessity for auto-close interlocks (re-energ.izing 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 (reference: AE00/C503
" Decay Heat Removal Problems at U.S. Pressurized Water Reactor's dated December 1985). Any appropriate design changes deemed necessary by this review will be 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 required.)
. The amber light on the breaker cu',1cle is wired such that it will be lit if the pressure interlocks are met, but is not affected by the status of the CFT valve 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.
. The 62A/TDC contact that was installed by ECN A-2487 to override the pressure interlocks after a loss of SFAS A or B power also overrides the CFT valve interlock.
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. RI-054 Rev. 1 -(Continued)
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. Other B&H plants do not have the CFT valve interlock. Is this interlock required? (1)
. What'is the basis for having an open-enable pushbutton at the i breaker cubicle? (Plant has'not. responded.)
- Can the auto-close feature be' eliminated' entirely by providing an alternative means of overpressure protection? (2) 1 (Plant. responded that interlock is required to prevent overpressuritation of DHR system piping.)
2 (Plant responded that installation of other means of protection such-as larger relief valves allows other B&W plants not to have this interlock.) ! l
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- REQUEST FOR INFORMATION (RI)
RI NO: 067 SYSTEM CODE: DHS ISSUE DATE: 07-29-87
SUBJECT:
LOW RANGE AND HIDE RANGE PRESSURE INSTRUMEE.T LOOPS t NUCLEAR ENGINEERING COORDINATOR: R. LAHRENCE DEPARTHENT: ' and NUCLEAR OPERATIONS _R. MACIAS TEAM HEMBERS: S. CARMICHAEL. D. SATPATHY. J. AREVALO._D. MARTIN _ , , , _ , l TEAM LEADER: KEITH PRINCE POTENTIAL, CONCERN /00ESTION: 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 m 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 i transmitter loops have an accuracy of about z 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 operable up to about 215 psi (250-35) which is less than RC pump NPSH requirement for 1/0 RC pump operation, (Prodess Standard Curve AP.101-2B). DH System may be put in service when the RCS pressure is at about 285 psig (with potential 35 psig error) which violates the design i pressure criteria of DH piping and equipment. t.
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- REQUEST FOR INFORMATION (RI) .]
v RI N0: 068 SYSTEM CODE: DHS ISSUE DATE: 07-28-87 f1
SUBJECT:
CARBON STEEL BOLTING MATERIAL DEPARTMENT: NUCLEAR ENGINEERING COORDINATOR: R. LAWRENCE TEAM HEMBER: TOM FAUBLE TEAM LEADER: KEITH PRINCE 1 i
' POTENTIAL CONCERN /00ESTION: )
{ Carbon steel bolting materials are used in components that carry borated I water. An'IE Bulletin was recently issued concerning corrosion of carbon f 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 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. Howevec, not all affected valves received the replacement bolting. In the decay heat system, bolting was replaced on the Anchor valves in safety-related service, but not on Velan valves. This concern is applicable to RCS, SIM,-PLS, CBS, BHS, SFC, RCD and RHS, All of these systems have valves with A-193 Grade B carbon steel bolting.
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'l .. REQUEST FOR INFORMATION (RI)
RI NO: 072 SYSTEM CODE: DHS ISSUE DATE: 07-29-87
SUBJECT:
TESTING OF HPI PIGGYBACK HODE (o COORDINA(OR: R. LAWRENCE DEPARTMENT: SYSTEMS ENGINEERING TEAM MEMBER: TOM FAUBLE TEAM LEADER: KEITH PRINCE POTENTIAL CONCERN /00ESTION: Testing of the decaj heat cooler to HPI: suction line'(piggyback mode) did not cover the case of an HPI pump and makeup pump operating in parallel. The original test, TP 203-4, demonstrated the capability to deliver 528 gpm to the HPI pump suction with sufficient NPSH available to the pump. However, with both an HPI pump and makeup pump operating, it is 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 calc 91ation.
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. REQUEST'FOR INFORMATION (RI)
RI NO: 076 SYSTEM CODE: DHS ISSUE DATE: 07-29-87 l
SUBJECT:
VORTEX FORMING AND NPSH FOR DH PUMPS DEPARTMENT: NUCLEAR ENGINEERING COORDINATOR: R. LAHRENCE TEAM MEMBER: D. SATPATHY; TEAM LEADER: KEITH PRINCE S. CARMICHAEL ' EDTENTIAL CONCERN /00ESTION: 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 instrumentation accuracy and alarm indication used for operator action for switch over. - 4
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-. REQUEST FOR INFORMATION (RI)
RI NO: 077 SYSTEM CODE: DHS ISSUE DATE: 07-31-87
SUBJECT:
VIBRATION ON DECAY HEAT PIPING DEPARTMENT: NUCLEAR ENGINEERING COORDINATOR: R. LAHRENCE' TEAM HEMBER: -D. MARTIN TEAM LEADER: KEITH PRINCE
,,.S. CARMICHAEL D. SATPATHY
' POTENTIAL CONCERN /0UESTION:
'DH Cooler Bypass piping 26123-6"-GD is supported only by piping 26021-10"-GD and 26121-10"-GD. Plant operators have observed at as much as 3 inches radial movement of this piping.
The procedure for decay heat initiation is being revised to start the pumps with 'B' Bypass valve HV-26038, located in the middle of line 26123-6", open. There is potential for loads on this piping during DH initiation that may cause excessive moveaent without proper supports. HV-26038 is a modulating valve controlled remotely from the control room without an operator assigned to the DH pump rooms. Problems that inight be caused by such movement would not be immediately known. Discussions 4 with the pipe support group indicated that proper supports'will be provided in the long term as priority 3. Failure'of this piping could affect ability to properly cooldown and maintain the plant in cold shutdown condition. i 1 1 1 1 I 4 j q l I l L 2
s. 6.0 ATTACHMENTS 6.1 List of RevioWed Documents 6.2 Status of RIs i
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LIST OF REVIEHED DOCUMENTS 1
- 1. . System Design Bases for DHS, NEP 5419, Rev. 1 (draft) !
- 2. System Status Report "DHS", Rev. 1
- 3. USAR Section 1.4, 6.2, 9.5.2, 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 1
- 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
- 13. Casualty Procedure C.12, Rev. 5
- 14. Casualty Procedure CP.101, Rev. 4
- 15. ' Casualty Procedure AP.103, Rev. 4
- 16. Ad"tinistrative Procedure AP.23
- 17. Administrative Procedure AP.42, Rev 5
- 18. Administrative Procedure AP.44, Rev. 11
- 19. Administrative Procedure AP.48, Original
- 20. Administrative Procedure AP.49, Original
'21. Emergency Opert, ting Procedures
- 22. Maintenance A kinistrative Procedure MAP.0006
- 23. Code of Federal Regulations 10CFR50 Appendix A,B,K,J 10CFR21 10CFR100 Appendix A
- 24. SYSTEMS Train'ing 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 M.522, Sh 1
- 29. P&ID M.521 Sh 1,2,3
- 30. P&ID H.544
- 31. P&ID M.545
- 32. Isometric Drawings: All for DHS
- 33. ECN A-2921 A-2931 R-0498
- 34. Temporary Changes to Operating Procedure A.8, from 01-01-85 to 06-01-87
- 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 1 l
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'a' .. LIST OF REVIEHED DOCUMENTS (Continued)
- 39. Surveillance Procedures: 18 19, Rev. 0 'li 29A 29B-
-203.05A, Rev. 19 j 203.05B, Rev. 19 203.06A, Rev. 8 203.06B, Rev. 10 203.06C/D, Rev. 11 203.09, Rev. 10 203.11, Rev. 7-204.03A, Rev. 20 200.02 204.038, 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' S-029
- 41. Work Request #134850 -
H/R 4776 H/R 67 H/R 12552 \ H/R 15609 , H/R 52036 H/R 80511 H/R 105763 H/R '98585 H/R 102132 W/R 102251 H/R 102676 N/R 100299 H/R 117875 H/R 129742 H/R 130460 H/R 133457 H/R 114607 ATTACHMENT 1 w .. , .. . .. .. . . .-. .. .. . . .. .. ..
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- o. STATUS OF~RIs.
i RI NUMBER SIATUS RSL NUMBER.
- 02: ACKN0HLEDGED RSL-RI-02
.03 OPEN 12 ACKNDHLEDGED RSL-RI-12 16 OPEN 17 OPEN.
18! ACKN0HLEDGED RSL-RI-18 23 ACKNOWLEDGED RSL-RI-23 29 1ACKN0HLEDGED RSL-RI-29 38 -OPEN 44 ACKNOWLEDGED RSL-RI-44 52 OPEN 54 ACKN0HLEDGED RSL-RI-54 167 ACKNOWLEDGED P' ,.-RI-67 68- ACKNOWLEDGED yu.-RI-68 . 72 OPEN 76 ACKNOWLEDGED RSL-RI-76 77 ACKN0HLEDGED RSL-RI-77
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