|
|
| Line 1: |
Line 1: |
| {{Adams
| | #REDIRECT [[IR 05000293/1986017]] |
| | number = ML20205P914
| |
| | issue date = 05/16/1986
| |
| | title = Insp Rept 50-293/86-17 on 860412-0425.No Violation Noted. Major Areas Inspected:Spurious Group 1 Primary Containment Isolation on 860404 & 12 & Failure of MSIV to Reopen After Isolations
| |
| | author name = Kister H, Strosnider J
| |
| | author affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
| |
| | addressee name =
| |
| | addressee affiliation =
| |
| | docket = 05000293
| |
| | license number =
| |
| | contact person =
| |
| | document report number = 50-293-86-17, CAL-86-10, NUDOCS 8605280047
| |
| | package number = ML20205P911
| |
| | document type = INSPECTION REPORT, NRC-GENERATED, INSPECTION REPORT, UTILITY, TEXT-INSPECTION & AUDIT & I&E CIRCULARS
| |
| | page count = 44
| |
| }}
| |
| See also: [[see also::IR 05000293/1986017]]
| |
| | |
| =Text=
| |
| {{#Wiki_filter:- -- _ _
| |
| : .
| |
| .
| |
| .i .
| |
| U. S. NUCLEAR REGULATORY COMMISSION
| |
| AL'GMENTED INCIDENT RESPONSE TEAM
| |
| Report No. 50-293/86-17
| |
| Docket No'. 50-293
| |
| Licensee: Boston Edison Company M/C Nuclear
| |
| ATTN: Mr. William D. Harrington
| |
| Senior Vice President, Nuclear
| |
| 800 Boylston Street ,
| |
| Boston, Massachusetts 02199
| |
| 4
| |
| Facility Name: Pilgrim Nuclear Power Station
| |
| Inspection At: Plymouth, MA
| |
| Inspection Conducted: April 12, 1986 through April 25, 1986
| |
| Team Leader: J. Strosnider, Chief,
| |
| Section 18, DRP, RI
| |
| Team Members: L. Doerflein, Martin McBride, Senior
| |
| Project Engineer,RI Resident Inspector, Pilgrim
| |
| K. Murphy, R. Fuhrmeister
| |
| Technical Assistant, DRS,RI Reactor Engineer, RI
| |
| M. Chiramal, Section Chief, AE00 S. Pullani
| |
| Fire Protection Engineer,
| |
| ; DRS, RI
| |
| Reviewed By .
| |
| /J/ Strosnider, Chief
| |
| LProj cts Section 1B, DRP
| |
| '
| |
| Approved By:
| |
| !
| |
| fi. Kistdfl Chief
| |
| Protects Branch No. 1, DRP
| |
| l
| |
| r ,
| |
| 8605280047 860516
| |
| PDR ADOCK 05000293
| |
| G PDR _
| |
| (J
| |
| | |
| -
| |
| .
| |
| .-
| |
| 2
| |
| TABLE OF CONTENTS
| |
| Page
| |
| 1.0 Introduction ............................................. 4
| |
| 2.0 Summary of Events ........................................
| |
| -
| |
| 5
| |
| 2.1 Ap ri l 4, 1986 Reactor Scram. . . . . . . . . . . . . . . . . . . . . . . . . . 5
| |
| 2.2 April 12, 1986 Reactor Scram......................... 6
| |
| 3.0 Evaluation of Inadvertent Closure of the MSIVs ........... 7
| |
| 3.1 Background .......................................... 7
| |
| 3.2 PCIS Trip Logic Circuit and MSIV Control Circuit
| |
| Designs ............................................. 8
| |
| 3.3 Investigation ....................................... 9
| |
| 3.4 Root Cause and Safety Significance ................. 12
| |
| 3.5 Conclusions and Recommendations .................... 12
| |
| 4.0 Evaluation of MSIV Problems ............................. 14
| |
| 4.1 Chronology of Events ............................... 14
| |
| 4.2 Valve Design and Operation ......................... 15
| |
| 4.3 : Investigation ...................................... 16
| |
| 4.4 Root Cause and Safety Significance ................. 18
| |
| 4.5 Conclusions and Recommendations..................... 19
| |
| 5.0 Evaluation of LPCI Injection Valve Leakage .............. 20
| |
| 5.1 Chronology of Events .............. ................ 20
| |
| 5 . 2. RHR Isolation Valve Descriptions ................... 21
| |
| 5.3 Past System Leakage Experience ...... .............. 22
| |
| i.
| |
| !
| |
| r
| |
| | |
| O
| |
| ~
| |
| 3
| |
| 5.4 History of RHR Valve Refurbishment and Leak Testing... 22
| |
| 5.5 As-Found RHR Walkdown and Valve Leakage Measurements.. 24
| |
| 5.6 Root Cause and Safety Significance ................... 25
| |
| 5.7 Conclusions and Recommendations ...................... 25
| |
| 6.0 Overall Summary and Conclusions ....................... ... 27
| |
| Figures / Pictures
| |
| Attachments
| |
| _-
| |
| | |
| .
| |
| ~
| |
| 4
| |
| 1. INTRODUCTION
| |
| On April 4 and 12, 1986, the Pilgrim reactor scrammed from low power
| |
| during routine reactor shutdowns. Both scrams were caused by unexpected
| |
| group I primary containment isolations. In both cases, the isolation
| |
| signal was promptly reset, but the four outboard main steam line
| |
| isolation valves (MSIVs) could not be promptly reopened. As a result,
| |
| the main condenser was not available as a heat sink during a portion of
| |
| - ~~
| |
| the reactor cooldown. The shutdown.on April lith was initiated because
| |
| the residual heat removal (RHR) system had been pressurized by leakage of
| |
| reactor coolant past a check valve and two closed injection valves in
| |
| the "B" RHR loop. An Unusual Event was declared because of the RHR
| |
| valve leakage.
| |
| NRC management discussed concerns about the recurring isolation and RHR
| |
| valve leakage problems with senior licensee management and issued Con-
| |
| firmatory Action Letter (CAL) No. 86-10 on April 12, 1986. This letter
| |
| required that all affected equipment be maintained in its as-found condi-
| |
| tion (except as necessary to maintain the plant in a safe shutdown con-
| |
| dition) until an NRC Augmented Inspection Team (AIT) was onsite to inspect
| |
| and reconstruct the events. The letter also required that the licensee
| |
| provide a written evaluation to the NRC cf 1) intersystem leakage through
| |
| RHR injection valves in the RHR system, 2) the spurious primary containment
| |
| isolation that occurred on April 12, and 3) the failure of the outboard
| |
| MSIVs to reopen after the isolation. The licensee agreed to seek authori-
| |
| zation for restart of the reactor from the Regional Administrator of NRC
| |
| Region I. The CAL is included in this report as Attachment 1. An AIT
| |
| was dispatched to the site on. April 12, 1986.
| |
| | |
| .
| |
| ~
| |
| 5 I
| |
| 2.0 SUMMARY OF EVENTS
| |
| 2.1 April 4, 1986 Reactor Scram
| |
| -At 1:00 p.m. on April 4, 1986, a reactor shutdown was initiated after
| |
| oil leakage was detected in the main turbine control oil system. The
| |
| low pressure coolant injection (LPCI) system was considered inoper-
| |
| able at that time due to an unrelated problem, water leakage past a
| |
| block valve, MD-1001-36A, in the residual heat removal system torus
| |
| cooling line.
| |
| At 8:15 p.m. on April 4, a group I primary containment isolation
| |
| (resulting in a reactor scram) occurred as reactor pressure decreased
| |
| to 898 psig in the shutdown sequence. The two low main steam line
| |
| pressure alarms (set to approximately 880 psig) were received at the
| |
| time of the isolation. The reactor mode switch had been moved from
| |
| the "run" to the "startup" position 45 minutes prior to the isolation.
| |
| The low steam line pressure containment isolation function is active
| |
| in the run mode but is bypassed when the mode switch is placed in the
| |
| startup mode.
| |
| The containment isolation signal was promptly reset following the
| |
| scram, however, the outboard MSIV's could not be reopened for
| |
| approximately one and a half hours. The inboard MSIV's were opened
| |
| during that time period. As a result of the closed MSIVs, most of
| |
| .the subsequent reactor cooldown was controlled by directing reactor
| |
| steam to the high pressure coolant injection (HPCI) turbine. The
| |
| HPCI system was operated in the test mode and dio not inject water
| |
| into the reactor.
| |
| During the review of this event the licensee concluded that all the
| |
| contacts in the reactor mode switch did not close properly when the
| |
| switch was transferred from the run to the startup mode during the
| |
| shutdown. As a result, the low pressure containment isolation func-
| |
| tion was still active when steam line pressure dropped below the
| |
| trip setpoint (about 880 psig). The licensee determined that proper
| |
| positioning of the mode switch required removing the Key from the
| |
| switch each time it was moved to a different mode. Training for all
| |
| control room operators on proper mode switch operation was conducted
| |
| prior to the subsequent reactor startup. Additional details of the
| |
| licensee's evaluation of the inadvertent closure of the MSIVS are
| |
| discussed in Section 3.0 of this report.
| |
| The licensee also concluded that an air leak in the "A" outboard
| |
| MSIV, A0-203-2A, (coupled with repeated attempts to open the valves)
| |
| probably lowered air pressure to the four outboard valves, preventing
| |
| them from fully opening. The air leak was attributed to foreign
| |
| materials in the MSIV pneumatic control valve. Additional details
| |
| of the licensee's evaluation of the problem with the MSIVs failing to
| |
| open upon demand and corrective actions are discussed in Section 4.0
| |
| of this report.
| |
| | |
| .
| |
| ~
| |
| 6
| |
| The evaluations of the Mode Switch and MSIV problems were reviewed
| |
| by the Operational Review Committee (0RC) on April 8, 1986. The
| |
| reactor was restarted at 2:46 a.m. on April 10, 1986.
| |
| 2.2 April 12,1986 Reactor Scram
| |
| Periodic RHR system high pressure alarms (400 psig) were received on
| |
| April 10 and 11, indicating that the RHR system was being pressurized
| |
| by reactor coolant leakage. The RHR piping in the "B" loop was warm,
| |
| indicating the leakage was coming through the normally closed injec-
| |
| tion valve, M0-1001-298, and an inline check valve, 1001-688. At
| |
| 2:16 p.m. on April 11, a second "B" loop injection valve, MO-1001-28B,
| |
| was closed in the RHR system in an attempt to stop the leakage. The
| |
| low pressure coolant injection (LPCI) subsystem of the RHR system
| |
| was declared inoperable at that time. However, leakage continued
| |
| into the RHR system causing a high pressure alarm two and a half
| |
| hours later. At 4:53 p.m. on April 11, 1986, a reactor shutdown
| |
| was initiated from about 92*s power and an unusual event was
| |
| declared due to the leaking valves.
| |
| At 1:56 a.m. on April 12, a group-one primary containment isolation
| |
| (with an associated reactor scram) occurred during the shutdown se-
| |
| quence. Reactor pressure was 908 psig at the time of the isolation.
| |
| The mode switch had been mcved from the "run" to the "startup" posi-
| |
| tion and the key removed from the mode switch twenty minutes earlier,
| |
| at 1:36 a.m. The isolation and scram occurred about 30 seconds after
| |
| the two main steam line low pressure alarms annunciated in the con-
| |
| trol room.
| |
| As before, the outboard MSIVs could not be opened for approximately
| |
| one and a half hours after the isolation signal was reset and the
| |
| HPCI system (in the test mode) was used to cool the reactor. The
| |
| reactor was placed in cold shutdown and the unusual event terminated <
| |
| at 9:08 a.m. on April 12, 1986.
| |
| | |
| - . .- -. . - - -- - . .- -
| |
| ,
| |
| -
| |
| i
| |
| ~
| |
| ; 7
| |
| i
| |
| i
| |
| *
| |
| 3.0 Evaluation of Inadvertent Closure of the Main Steam Isolation Valves
| |
| :
| |
| .
| |
| '
| |
| Following the scram on April 12, 1986, the licensee promptly organized a
| |
| team consisting of approximately 14 technical and support personnel to
| |
| investigate potential failures of the reactor mode switch (RMS), other
| |
| <
| |
| potential problems in the PCIS circuitry, and operator errors which could
| |
| have contributed to this event. The scope of the investigation included
| |
| a thorough analysis of previous events and included trouble shooting
| |
| '
| |
| plans, procedures and special tests. Members of the NRC Augmented Inspec-
| |
| tion Team (AIT) monitored the activities of the licensee team and assessed
| |
| the operational anomalies that occurred in the PCIS circuitry.
| |
| ~~~ --~~
| |
| 3.1 Background ,
| |
| <
| |
| On April 4 and April 12, 1986, while shutting down the reactor, the ,
| |
| ; Pilgrim unit experienced a reactor trip due to inadvertent closure
| |
| of all eight main steam isolation valves (MSIVs). On both occasions -
| |
| the reactor mode switch was in the "Startup/ Hot Standby" position and
| |
| the inadvertent closure of the MSIVs occurred after the operators
| |
| '
| |
| received alarms indicating main steam line pressure was less than 880
| |
| psig. During the April 4th event, the reactor scram due to MSIV
| |
| closure occurred almost immediately following the alarm; while on
| |
| April 12, the scram apparently occurred 30 to 40 seconds after the
| |
| alarms came in.
| |
| '
| |
| 1
| |
| j Following investigation and analysis of the April 4th event, the
| |
| ; licensee had concluded that the cause of inadvertent closure of the
| |
| !
| |
| MSIVs and subsequent reactor scram was due to failure of some
| |
| l contacts of the reactor mode switch. The contacts in question are
| |
| , in the primary containment isolation system (PCIS) logic channel
| |
| circuits and are designed to inhibit the actuation of the trip
| |
| i circuits on a low steam line pressure condition. That is,
| |
| i the mode switch contacts, when the mode switch is in any position
| |
| other than "Run", bypass the low steam line pressure trip of the
| |
| PCIS.
| |
| 4
| |
| l Based on testing of a spare mode switch the licensee also determined
| |
| ; that, as a means of assuring that the mode switch contacts function
| |
| properly, the operators should remove the key from the mode switch
| |
| '
| |
| ; handle after the switch is operated. The mode switch key can be re-
| |
| moved from the handle only if the switch is aligned fully in one of
| |
| the four required positions, i.e., the key cannot be removed if the
| |
| switch is in an intermediate position. All operators were trained on
| |
| proper mode switch operation, using the spare mode switch, prior to
| |
| the reactor startup on April 10, 1986.
| |
| '
| |
| On April 12, 1986, while shutting down, the mode switch was moved
| |
| j from the "Run" position to the "Startup/ Hot Standby" position and
| |
| the key was removed from the handle. However, 30 to 40 seconds
| |
| ; following the expected alarms indicating steam line low pressure,
| |
| i
| |
| _ _ _ . __ _______ . _ - - -
| |
| _ _ _ _ _ _ _ _ _ _- . . _ - . _ .
| |
| | |
| .
| |
| -
| |
| 3
| |
| the reactor scrammed due to the unexpected closure of the MSIVs.
| |
| Once again the reactor mode switch contacts in the PCIS trip logic
| |
| channel circuits associated with the MSIVs were suspected to have
| |
| caused the inadvertent closure of the MSIVs.
| |
| 3.2 PCIS Trip Logic Circuit and MSIV Control Circuit Designs
| |
| 3.2.1 PCIS Trip Logic Circuit
| |
| _
| |
| -~~~
| |
| The PCIS trip logic scheme consists of four trip logic
| |
| channels (designated A1, A2, B1, and B2) arranged in a
| |
| one-out-of-two taken twice logic (i.e., Al or A2 and 81 or
| |
| B2) to cause a trip. Figure 3.1 is an elementary diagram
| |
| showing the trip logic channel Al of the PCIS for the MSIVs,
| |
| main steam line drain valves and reactor water sample
| |
| valves. When the reactor mode switch is in the "Run" mode,
| |
| the following conditions will cause the actuation of the
| |
| PCIS trip logic channels (i.e., deenergization of relay
| |
| 16A-K7A, B, C, and D):
| |
| (1) Main steam line low pressure (<880 psig)
| |
| (2) Low low reactor water level
| |
| (3) Main steam line high radiation
| |
| (4) Main steam line high flow
| |
| (5) Main steam tunnel high temperature
| |
| These are referred to as isolation conditions 1, 2, 3, 4,
| |
| or 5 in the discussion that follows.
| |
| .
| |
| When the mode switch is in other than the "Run" mode (i.e.,
| |
| shutdown, refuel or Startup/ Hot Standby), a main steam
| |
| line low pressure condition will not cause the actuation of
| |
| the PCIS trip logic channels. This feature enables the
| |
| MSIVs to remain open while the reactor pressure is less
| |
| than 880 psig during a normal reactor startup. However, a
| |
| high reactor water level condition during these modes
| |
| (i.e., other than "Run") will cause the actuation of-the
| |
| PCIS trip logic channels.
| |
| As stated before, the actuation of the PCIS trip logic
| |
| means deenergization of relays 16A-K7A, B, C and D.
| |
| Contacts of these relays, arranged in a one-out-of-two
| |
| taken twice logic, actuate the MSIV control circuits
| |
| discussed below and close the MSIVs.
| |
| _
| |
| | |
| . .
| |
| ~
| |
| 9
| |
| 3.2.2 MSIV Control Circuit
| |
| Each MSIV is controlled by two, three-way, direct acting,
| |
| solenoid valves; one powered by 120 VAC and the other'120
| |
| VDC. The MSIV pilot system is arranged so that when one or
| |
| both solenoid valves are energized, normal air supply pro-
| |
| vides pneumatic pressure to an air operated pilot valve
| |
| which in turn directs air pressure to the MSIV valve opera-
| |
| tor so that the MSIV can be opened against the action of
| |
| the spring. When both the solenoids are deenergized by a 3
| |
| PCIS trip logic actuation or a manual closure signal, the
| |
| air pressure is directed to the opposite side of the valve
| |
| operator piston which along with action of the spring
| |
| closes the MSIV.
| |
| 3.3 Investigation
| |
| Investigation revealed that the reactor scrams which occurred on
| |
| April 4 and 12, 1986 were initiated by the actuation of the Reactor
| |
| Protection System (RPS) due to closure of the Main Steam Isolation
| |
| Valves (MSIVs). The closure of the MSIVs was initiated by the PCIS
| |
| trip logic circuitry discussed in Section 3.2.1.
| |
| On both the April 4 and 12, 1986 events, it was initially determined
| |
| that the only PCIS signal present at the time of the isolation was
| |
| main steam line low pressure. On both occasions, the reactor mode
| |
| switch was in the "Startup/ Hot Standby" position and the reactor
| |
| pressure was being reduced below 880 psig during the controlled cool
| |
| down of the reactor. The PCIS trip signal from the four main steam
| |
| line low pressure switches (261-30A through D) should have been
| |
| inhibited by the previously performed operator action of transferring
| |
| the mode switch from the "Run" position to the "Startup/ Hot Standby"
| |
| position.
| |
| The reactor mode switch is a pistol grip, key locked, four position
| |
| control switch. The four positions are: " Shutdown", " Refuel",
| |
| "Startup/ Hot Standby", and "Run" (see Figure 3.2). The switch is
| |
| made up of four banks of General Electric Model SB-1 rotary control
| |
| switches (see Figure 3.3), having 8 stages per bank (i.e.,16 sets
| |
| of cam operated contacts per bank). The banks are coupled together
| |
| by gears. The pistol grip handle is attached to the second bank
| |
| from the left hand side of the switch.
| |
| Reactor mode switch malfunctions causing problems of a similar nature
| |
| have occurred at several other nuclear plants and were the subject of
| |
| IE Information Notice 83-43. Pilgrim had experienced a problem with
| |
| the mode switch in 1983 (Reference ORC Meeting Minutes 84-104 and
| |
| Failure & Malfunction Report 83-133). General Electric Information
| |
| Letters (SIL) Number 155 and its supplements 1 & 2; and SIL 397 dis-
| |
| cuss instances of failure of "SB" model switches and recommend
| |
| actions to be taken.
| |
| i
| |
| 4
| |
| | |
| .
| |
| ~
| |
| 10
| |
| Previously, in accordance with SIL 397, during Refuel Outage VI, an
| |
| SB-9 model mode switch was bought and tested. The SB-9 mode switch
| |
| was a used unit rebuilt by GE. Testing was also performed on a S.8-1
| |
| model switch. Both switches functioned properly and each had a
| |
| definite ' feel' during a transfer operation. The SB-1 required a
| |
| specific technique be used to ensure proper align. ment of contacts
| |
| while the SB-9 operated in a stiff and hard manner.
| |
| Following this testing, Operations personnel visited the test site
| |
| and familiarized themselves with the feel and technique used to
| |
| properly transfer the existing SB-1 switch. This familiarization
| |
| ~~~--
| |
| reduced the concern for the proper operation of the SB-1 Mode Switch.
| |
| This experience coupled with the knowledge that the new SB-9 switch
| |
| operated in a stiff and hard manner and the extensive time required
| |
| to change out and post-work test the replacement switch contributed
| |
| to a subsequent licensee decision to continue operation with the
| |
| existing SB-1 Mode Switch.
| |
| 3.3.1 Analysis and Evaluation of the April 4, 1986 Event
| |
| During the shutdown on April 4th, the mcde switch was trans-
| |
| ferred by an operator-in-training under direct supervision
| |
| of the Nuclear Watch Engineer. The watch engineer " wiggled" -
| |
| the mode switch to " feel" that it was in the right position.
| |
| The mode switch key was not removed from the switen handle
| |
| following the transfer.
| |
| The operator who had transferred the mode switch in the
| |
| control room prior to tne April 4, 1986 scram, had not :
| |
| been trained on the SB-1 Mcdel Switch at Pilgrim and had
| |
| no previous experience with it. Even though tne watch '
| |
| engineer checked the position of the mode switch, it is
| |
| possible that the switch was not actually in the correct .'
| |
| position because the key was not removed (as a positive
| |
| verification of proper positioning) after this trantfer. -
| |
| In retrospect, inadequate training of the operatcr could :
| |
| have contributed to the event. -
| |
| Following this event and in accordance with the recommenda-
| |
| tions in SIL 155, an inspe: tion of the Reactor Mode Switch
| |
| was conducted at Pilgrim on April 5, 1986. No ir.dication
| |
| of cracking or broken contacts or of any other adverse con- ;
| |
| dition was observed. Examination did indicate that proper '
| |
| preloading of the switch contacts existed.
| |
| In summary, the licensee concluded that the most probable !
| |
| root cause of this event was that at laast two of the tode ,
| |
| switch contacts (10, 26, 42 & 58) did not close or remain +
| |
| closed after the mode switch was transferred from "Run" to
| |
| "Startup/ Hot Standby". Corrective actions taken as a result
| |
| .
| |
| l
| |
| I
| |
| i
| |
| >
| |
| ,e aw-~.wewe,e . e w e -
| |
| | |
| -.
| |
| . .
| |
| 11
| |
| 4
| |
| .
| |
| of this event included development of a prescribed technique i
| |
| for transferring the modo switch and training of Operations i
| |
| personnel in its application. ;
| |
| 3.3.2 Analysis _andEvaluationoftheApril 1_2, 1986 Event
| |
| As discussed earlier in this report, the containment isola- ,
| |
| tion and reactor scram on April 12, 1986 were similar to
| |
| the April 4th event. However, on April 12, the tcram oc-
| |
| curred 30 to 40 seconds after the main steam line low
| |
| ~~ -"~~
| |
| pressure alarms cane in. Also, on April 12, the transfer
| |
| was performed by an experienced operator and the mede '
| |
| switch key was removed.
| |
| The licensee investigation team's cyaluation of possible
| |
| means by which the MSIVs could close was comprehensive. It
| |
| considered loss of instrument air, failure of the MSIV's AC !
| |
| and DC solenoid valves, loss of AC and DC control power, ,
| |
| simultaneous actuation of MSIV test switches or associated +
| |
| circuits, operation of MSIV hand switches, failure of
| |
| relays associated with the MSIV close circuit, and the PCIS
| |
| logic circuits. The team analyzed available event data,
| |
| interviewed plant operators, reviewed past history for
| |
| similar events, performed adaitional functional tests and i
| |
| calibration tests, conducted special tests and conducted
| |
| walk-downs of the associated systems.
| |
| The NRC inspectors reviewed test documents to assess their i
| |
| technical adequacy, evaluated the safety consequences of l
| |
| these actlyities, and analyzed the test results to ascer- i
| |
| tain that the components functioned as intended. No signi- -
| |
| ficant problems were identified. Attachment 3 lists the ;
| |
| ;. tests reviewed and performed as of April 26, 1986.
| |
| a .
| |
| 4' ; Ouring the performance of one test, surveillance test
| |
| ; 8.M.1-19, an unanticipated closure of the MSIVs occurred. !
| |
| After the initial round of tests and analyses the licensee
| |
| decided that the inadvertent closure of the MSIVs was due '
| |
| *
| |
| ;
| |
| '
| |
| to actuation of the PCIS trip logic circuits. Based on the
| |
| results of the tests conducted, it was further concluded '
| |
| p that testing of the reactor mode switch was necessary. On
| |
| i April 19, 1986, a special test of the switch was conducted.
| |
| l The test involved monitoring of the mode switch contacts in
| |
| the suspect PCIS trip logic circuits and multiple operations ,
| |
| of the mode switch in its various positions. To consider
| |
| t the human factors aspect of the mode switch operation,
| |
| [ several operators were used in the manipulation of the
| |
| f switch. The switch was moved from the "Run" to "Startup" -
| |
| position approximately thirty tinies during this test.
| |
| , During this testing the contacts in the node switch were
| |
| ,
| |
| instrumented in order to determine if they were opening
| |
| j. and closing properly. ,
| |
| 4-
| |
| e
| |
| +nw.we n
| |
| | |
| . - __
| |
| .
| |
| .
| |
| *
| |
| 12
| |
| Review of the mode switch special test data showed that the -
| |
| moc:e switch contacts ia the PCIS trip circuits functioned
| |
| consistently as designed. It could also be concluded that,
| |
| discounting random failures, the mede switch was not the
| |
| root cause of the events of April 4 and April 12, 1986.
| |
| Following the mcJe switch test, the 11censee's team con-
| |
| centrated in identifying and testing fcr other potential
| |
| failu*es affectir.g at least two channels of the PCIS trip .
| |
| logic circuits. Possible causes such as icose wires and
| |
| tarminations, voltage surges on c7rcuit neutrals, ground
| |
| circuit anomalies, and wiring errors during the recent
| |
| replacement cf RPS and PCIS relays vere assessed through
| |
| te stir.g and ir.spection. Inis testing and inspectior did
| |
| not confl.vm the cause of the unanticipated containment
| |
| isolations.
| |
| 3.4 hot Cause and Saf 1 tyjhnjfjcance
| |
| The licersee and its special teens are continuing their investigation
| |
| into the root cause of the inadverter.t closures of the MSIVs that
| |
| occurred on Aprfl 4 and April 17, 1985. No root cause for the un-
| |
| expected containment isolations riad been identified at the conclusion
| |
| of this inspection, although a . mode switch failure was suspected,
| |
| Until a root cause is established, the possibility that these inad-
| |
| vertent closures c0uld cccur 1.n any mode cf reactor operation cannot
| |
| be ruled out. The safety functicn of the main steam isolation vaives
| |
| is to close when needed to isolate the reactor primary systeA.
| |
| Although inadvertent cicsure of the MSIVs aligns the valves in their
| |
| safe configuration, such closures are of concern for the following
| |
| reasons:
| |
| 1. Inadvertent closure can lead to a reactor trip, a turbine trip, *
| |
| and a loss of the normal heat sink and normal pressure control
| |
| of the reactor.
| |
| ,
| |
| .
| |
| 2. Closure could cause challenges to safety related systems such
| |
| as the main steam line safety and relief valves, the RPS, HPCI,
| |
| and RCIC. ,
| |
| 3. Closure could result in increasing the stress level of the opera-
| |
| tors, as a result of the potential transients identified in
| |
| items 1 and 2 above.
| |
| 3.5 Conclusions and Recommendations
| |
| The licensee has worked hard to determine the root cause of inadver-
| |
| tent closures of the MSIVs. However, the root cause or causes of
| |
| the problem have not been established as yet. Due to the concerns
| |
| raised by the inadvertent closures of MSIVs, the root cause should
| |
| I
| |
| - - -. .
| |
| | |
| .
| |
| '
| |
| 13
| |
| be determined prior to restart of the unit or prior to operation
| |
| under conditions where an unanticipated containment isolation could
| |
| significantly challenge reactor safety systems or operators.
| |
| In addition, considering the important safety functions the mode
| |
| switch performt, its operation should not be subject to an operator's
| |
| " feel", or a prescribed technique for its transfer operation. The
| |
| licensee should continue to work on resolving these noted problems.
| |
| Licensee activities in this area will be evaluated in future
| |
| ,_ . _ _ _ . -
| |
| inspections (86-17-01).
| |
| o
| |
| i
| |
| | |
| .
| |
| *
| |
| 14
| |
| 4.0 EVALUATION OF MAIN STEAM ISOLATION VALVE (MSIV) PROBLEMS
| |
| This section discusses the failure of the outboard Primary Containment
| |
| Main Steam Isolation Valves (MSIVs) to open upon demand following the
| |
| reactor trips on April 4 and 12, 1986. For reference, a simplified
| |
| drawing of an MSIV and an enlarged drawing of the valve pilot poppet
| |
| assembly are included as figures 4.1 and 4.2 respectively. A list of
| |
| procedures and other documents reviewed is included in attachment 4.
| |
| 4.1 Chronology of Events
| |
| On April 4, 1986 at 8:15 p.m., during a planned reactor shutdown,
| |
| the reactor tripped due to all eight MSIVs closing (Group I
| |
| Isolation). Following the trip, the operators reset the Grcup I
| |
| Isolation signal and attempted to open the outboard MSIVs. The MSIV
| |
| control switches were left in the open position for approximately one
| |
| minute. During this time, the operators observed both red (open) and
| |
| green (closed) position indication on the outboard MSIVs, however,
| |
| the valves did not go full open. When the operators placed the con-
| |
| trol switches to the closed position, they observed the valve indica-
| |
| tion went green (full closed) in less than one second. The inboard
| |
| MSIVs were then successfully cycled open and closed The High Pres-
| |
| sure Coolant Injection (HPCI) system was used in the full flow test
| |
| lineup to control reactor pressure which is the normal method of
| |
| pressure control if the MSIVs can't be opened. Approximately one and
| |
| a half hours after the MSIV isolation was received the outboard MSIVs
| |
| opened upon demand.
| |
| The licensee considered four possible causes for the failure of the
| |
| outboard MSIV's to reopen: 1) simultaneous mechanical binding of the
| |
| four outboard MSIV's, 2) excessive differential pressure across the
| |
| valves, 3) low instrument air pressure, and 4) loss of electrical
| |
| control power. During the followup investigation, the licensee dis-
| |
| covered a large air leak on the control system for the "A" outboard
| |
| MSIV which continuously ported the under piston area of the MSIV air
| |
| cylinder. During the repair of the air leak, debris (paper and
| |
| yellow plastic) was found lodged in the pneumatic four way valve.
| |
| Some of the pieces of paper were folded, indicating that they were
| |
| manually placed in the controller rather than blown in from the in-
| |
| strument air system. The entire air distribution manifold on the
| |
| "A" outboard MSIV (last disassembled during the 1984 outage) was
| |
| removed for cleaning. Inspections for debris were also performed on
| |
| the air distribution manifolds of the "A", "B" and "C" outboard MSIVs
| |
| as well as the "A", "C" and "D" inboard MSIVs with negative results.
| |
| The "0" outboard and "B" inboard MSIVs were not inspected as they had
| |
| recently been worked on. The licensee's evaluation of the source of
| |
| the debris was ongoing during the AIT and will be examined during a
| |
| future it'spection (86-17-02).
| |
| - - .. .. -. _ - - - --. .
| |
| | |
| r
| |
| .
| |
| *
| |
| 15
| |
| Following the inspections of the MSIt air system, testing was perform-
| |
| ed to determine if reduced air pressure would preclude the MSIVs
| |
| from achieving full stroke. The test results indicated that approxi-
| |
| mately 40 psig supply pressure would open the MSIV one half inch,
| |
| resulting in both red and green valve position indication and that
| |
| full valve stroke could not be achieved when normal supply pressure
| |
| was introduced slowly to the air cylinder. As no other problems were
| |
| identified duri :g the followup investigation, the licensee concluded
| |
| that the failure of the outboard MSIVs to open upon demand was most
| |
| likely caused by a lowered cylinder air supply pressure due to the
| |
| leak on the "A" outboard MSIV. The reactor was restarted on April 10,
| |
| 1986.
| |
| On April 12, 1986 at 1:56 am, during another planned shutdown, the
| |
| reactor tripped due to all MSIVs closing. Approximately four and a
| |
| half minutes after the MSIV closure, the operators reset the Group I
| |
| Isolation signal and attempted to open the outboard MSIVs. As during
| |
| the previous event, the MSIV control switches were left in the open
| |
| position for approximately one minute, operators observed both red
| |
| and green valve position indication, and the MSIVs failed to open.
| |
| The control switches for the outboard MSIVs were placed in the closed
| |
| position. Then with personnel stationed in the steam tunnel to ob-
| |
| serve MSIV stem movement, operators made several attempts to open
| |
| only the "A" and "C" outboard MSIVs. In one case the MSIV control
| |
| switch was left in the open position for approximately five minutes.
| |
| personnel in the steam tunnel reported that, during the attempts to
| |
| open the "A" and "C" outboard MSIVs, the valve stem would travel ap-
| |
| proximately one half inch and then stop. There was no sound of steam
| |
| flow when the MSIVs stroked the one half inch. It was also observed
| |
| that MSIV air cylinder supply pressure was normal.
| |
| Again, as during the April 4, 1986 event, the operators were able to
| |
| open the inboard MSIVs (which were left open) and HPCI was used to
| |
| control reactor pressure. Approximately one and a half hours after
| |
| the Group I Isolation, the outboard MSIVs opened upon demand. The
| |
| operators noted that the differential pressure across the outboard
| |
| MSIVs was 30 psi when the valves were opened. Reactor pressure at
| |
| that time was approximately 310 psig.
| |
| 4.2 Valve Design and Operation
| |
| Valve Design
| |
| The Main Steam Isolation valves, manufactured by Atwood and
| |
| Morrill Company Inc., are 20 inch globe valves having a "Y"
| |
| pattern body. The valves have a cylindrical main disc (poppet)
| |
| moving in a centerline 45 degrees upward from the axis of the
| |
| horizontal main steam inlet line. An air cylinder is utilized
| |
| to operate the isolation valve. Air for the outboard valves
| |
| and air or nitrogen for the inboard valves is used to open the
| |
| | |
| .
| |
| -
| |
| 16
| |
| valve while springs and/or air (nitrogen) close the valve. The
| |
| air cylinder is capable of opening the MSIV with the design
| |
| differential pressure of 200 psi across the main poppet. The
| |
| MSIV also contains an internal pilot valve whose seat is in the
| |
| middle of the main poppet. The pilot valve provides a means of
| |
| balancing the pressure across the main poppet, just before the
| |
| main poppet is lifted and while it is off its seat. The first
| |
| three quarters of an inch stem travel only opens the pilot
| |
| poppet after which the main poppet is lifted of." its seat. The
| |
| total MSIV stem travel from full close to full open is nine and
| |
| one half inches.
| |
| Due to a history of problems with leak tightness and two valve
| |
| stem failures in 1978 and 1982, the licensee modified all eight
| |
| MSIVs during the sixth refueling outage, which ended in December
| |
| 1983. These modifications included: new main poppets with an
| |
| elongated poppet nose to position the poppet in a proper seating
| |
| position; increasing stem diameter and fillet radius on the
| |
| backseat surface; addition of main poppet anti-rotation devices;
| |
| and addition of self-aligning pilot poppets.
| |
| Valve Operation
| |
| Opening MSIVs with the reactor pressurized, such as following a
| |
| Group I Isolation, is described by procedure. Basically the
| |
| sequence requires that all the , outboards MSIVs be opened first
| |
| to allow trapped condensation to drain. The outboard valves
| |
| should open after the pilot poppet reduces the differential
| |
| pressure across the main poppet to within 200 psi. The steam
| |
| line drain valves (numbers MOV 220-1, MOV 220-2 and MOV 220-3)
| |
| are then opened to equalize pressure across the inboard MSIVs.
| |
| When the differential pressure across the inboard MSIVs is
| |
| within 50 psi (administrative limit), as measured between
| |
| reactor pressure and main steam pressure upstream of the turbine
| |
| stop valves, the inboard MSIVs are opened and the drain valves
| |
| are shut.
| |
| 4.3 Investigation
| |
| Following the MSIV isolation and reactor trip on April 12, 1986, the
| |
| licensee formed a multi-disciplined team to investigate and determine
| |
| the cause of the outboard MSIV failure to open upon demand. Activi-
| |
| ties of the team were observed by the NRC inspectors who found that
| |
| the evaluation team performed a detailed review and analysis of the
| |
| MSIV problem. Actions taken by the team included: bringing a valve
| |
| vendor representative onsite to review valve characteristics; operator
| |
| interviews; review of surveillance test data; review of all previous
| |
| trip reports for similar events; system walkdowns; identification and
| |
| discussions of potential causes; and contact with the Institute of
| |
| Nuclear Power Operations and other utilities to identify similar
| |
| occurrences at other facilities.
| |
| | |
| . _ .
| |
| .
| |
| *
| |
| 17
| |
| ,
| |
| The evaluation team identified the following seven possible causes
| |
| for the failure of the MSIVs to reopen: electrical failure; air
| |
| supply problems; all pilot poppets broken off; insufficient time
| |
| allowed by operator for area above main poppet to bleed off; inboard
| |
| MSIVs leaking so that the differential pressure across outboard
| |
| '
| |
| .
| |
| M3IVs could not be reduced to less than 200 psi; mechanical binding
| |
| of main poppet; and mechanical binding of pilot poppet. Based on
| |
| system walkdowns, functional tests, etc., the team concluded that ,
| |
| the most probable cause of the outboard MSIVs failure to open upon
| |
| '
| |
| - _
| |
| demand was the pilot poppet becoming detached from the valve stem.
| |
| Prior to the sixth refueling outage, during which the MSIVs were
| |
| modified, the pilot valve was an integral part of the MSIV stem. No
| |
| cases were found, prior to this outage, where the MSIVs could not be
| |
| opened following an isolation with the reactor pressurized. Follow-
| |
| ing the modifications and plant;startup in December 1983, only three
| |
| MSIV isolations occurred with the reactor pressurized. Two of the
| |
| three were the events of April 4 and 12, 1986 during which the out-
| |
| board MSIVs would not open upon demand. .The third event occurred
| |
| during a planned shutdown on June 15, 1985. However, in this case
| |
| ; no attempt was made to reopen the MSIVs.
| |
| The modification to the MSIV pilot valve involved installation of a
| |
| " floating" pilot poppet. The design was intended to provide a
| |
| laterally floating pilot poppet to improve leakage characteristics
| |
| and reduce MSIV stem bending stresses. As seen in Figure 4.2, the
| |
| pilot poppet is attached to the stem by tnreading tae poppet onto the
| |
| pilot poppet nut which is held on the stem by the split retaining
| |
| ring installed in the stem groove. A set screw is installed and
| |
| >
| |
| staked into the pilot poppet to prevent the poppet from unscrewing
| |
| itself from the pilot poppet nut.
| |
| The evaluation team developed a test to verify their conclusion that
| |
| the pilot poppet had become disconnected from the stem. The test
| |
| '
| |
| consisted of pressurizing the volume between a pair of MSIVs to 23
| |
| psig and then slowly increasing the air supply pressure to the out-
| |
| board MSIV air cylinder to slowly open the valve. Expected results
| |
| would be that within the first three quarters of an inch stem travel
| |
| the pilot poppet should lift and depressurize the volume between the
| |
| MSIVs. After three quarters of an inch stem travel (the limit of
| |
| pilot poppet travel) the main poppet would open to depressurize the
| |
| volume between the MSIVs. The inspector reviewed the test procedure
| |
| '
| |
| to verify it was technically adequate and approved by the Operations
| |
| Review Committee. In addition, the inspector observed the test per-
| |
| formed on the "A" outboard MSIV. The results of this test clearly
| |
| indicated that the~ pilot poppet was not attached to the valve stem.
| |
| ;
| |
| Similar tests were run on the remaining outboard MSIVs and, although
| |
| the results were not as definitive, there were indications the pilot
| |
| poppets were not opening as soon as expected.
| |
| I
| |
| . . . . _ _ . --.. , -_ _ __ . _ _ - . _ _ _ . _ _ . _ ,
| |
| | |
| _ .- .-.- ..
| |
| '
| |
| .
| |
| *
| |
| 18
| |
| Based on the test results, the licensee disassembled all eight MSIVs
| |
| for inspection. The results of these inspections were: on two MSIVs
| |
| ("A" outboard and "C" inboard) the pilot poppets were detached from
| |
| the valve stem; on the "D" outboard MSIV the pilot poppet became de-
| |
| tached during MSIV disassembly; three other pilot poppets ("0"
| |
| inboard, "B" and "C" outboard) had started to unscrew themselves from
| |
| the pilot poppet nut and exhibited 3/8 to 5/16 of an inch axial play;
| |
| -and the remaining two MSIVs ("A" and "B" inboards) had the pilot
| |
| poppet fully engaged to the pilot poppet nut. In those cases where
| |
| the pilot poppet had started to unscrew itself, the threads on the
| |
| -
| |
| poppet and nut were damaged.
| |
| Prior to disassembly the licensee also performed Local Leak Rate
| |
| Testing (LLRT) of all MSIVs. The results of the LLRT are included
| |
| in the following table. Leakage rates are in standard liters per
| |
| minute (sim).
| |
| MSIV Leakage
| |
| "A" inboard (IA) 44.5 slm
| |
| "A" outboard (2A) 5.5 slm
| |
| "B' inboard (IB) 23.2 slm
| |
| "B" outboard (28) 2.8 slm
| |
| "C" inboard (1C) 4.03 slm
| |
| "C" outboard (2C) 0.47 slm -
| |
| "D" inboard (1D) 33.5 slm
| |
| "D" outboard (20) 8.5 slm
| |
| The inspector noted that the Technical Specification limit for valve
| |
| leakage is 5.43 sim. However, the inspector also noted that the
| |
| measured leak rates were significantly lower than those from the two
| |
| previous LLRTs.
| |
| 4.4 Root Cause and Safety Significance
| |
| The cause of the outboard MSIV failure to open upon demand was the
| |
| pilot poppets b u ming detached from the valve stem or inhibited
| |
| from fully opening so that the differential pressure across the main
| |
| '
| |
| poppet would prevent the MSIV air cylinder from opening the valve.
| |
| ~
| |
| At the end of the AIT inspection the cause for the pilot poppets
| |
| becoming unscrewed and/or detached from the pilot poppet nut was
| |
| still under analysis by the licensee to determine if it was due to
| |
| > an installation error or design error. However, it was clear that ,
| |
| the set screw did not prevent tne pilot poppets from unscrewing from
| |
| the pilot poppet nut.
| |
| Subsequent to the AIT inspection the licensee concluded that the
| |
| lack of positive set screw engagement was due to an inadequate
| |
| ,
| |
| installation procedure coupled with the absence of a torque
| |
| ^
| |
| requirement between the pilot poppet and poppet nut allowing imposed
| |
| rotational / vibrational forces to unscrew these assemblies.
| |
| ;
| |
| - - - .
| |
| _ - - - , - _ -- .__ _ _. _ _ _ _ _ _
| |
| | |
| _
| |
| I%
| |
| .
| |
| I
| |
| v '
| |
| '
| |
| '
| |
| .
| |
| 1_9
| |
| .
| |
| Analysis by the lidensee is on going' to ensure that, with the problems
| |
| - identified, the MSIVs met the safety design basis as stated in the
| |
| Final Safety Analysis Report. However, the safety objectives of the
| |
| MSIVs are to close to limit the loss of reactor coolant and limit
| |
| the release of radioactive materials. The design of the valve is
| |
| such thatJapparently even a detached pilot poppet cannot become dis-
| |
| lodged and prevent the MSIV from fulfilling the safety cbjective.
| |
| This was reinforced by the LLRT results. Nonetheless, failure of
| |
| the valves to reopen did result in using a safety system to control
| |
| reactor pressure and temperature and presented a.dditional challenges
| |
| to the reactor operators. I'n, addition, based on this event and on
| |
| ,
| |
| reports from other facilities, there may be generic safety implica-
| |
| .
| |
| tions with regard to the use of set screws.
| |
| 4.5 Conclusions and Recommendations
| |
| -The MSIV evaluation team did a thorough job in identifying the cause
| |
| . sor the MSIVs failing to open on demand. Based on the observations
| |
| r
| |
| and testing performed during the,first event of April 4, 2986, the
| |
| inspector could not fault the licensee for not identifying the prob-
| |
| lem then. ,Also, based on the LLRT results, it appears that the MSIV
| |
| modifications have significantly reduced the valve leakage problems
| |
| roted/prvviously.
| |
| ,
| |
| The licensee, sbould continue the root cause analysis, to identify why
| |
| the set screws did.not prevent the MSIV pilot poppets from unscrewing
| |
| off the-poppet nut.in order that a perma.nent fix can be implemented.
| |
| The corrective actions including proposed design changes will be
| |
| evaluated when they'are available (86-17-03).
| |
| :
| |
| ,
| |
| I
| |
| e
| |
| 1
| |
| a s , , , - , - .:,, . ~ - - -
| |
| - . - - - . .
| |
| | |
| .
| |
| *
| |
| 20
| |
| 5.0 EVALUATION OF LPCI INJECTION VALVE LEAKAGE
| |
| -5.1 Chronology of Events
| |
| Tables 5.1, 5.2, and 5.3 summarize the chronology of events signifi-
| |
| cant to the RHR valve leakage question. The events begin with the
| |
| pulling of control rods on April 10, 1986 and end with the securing
| |
| from the Unusual Event (declared as a result of RHR valve leakage) on
| |
| April 12, 1986.
| |
| ~
| |
| At about 10:00 a.m. on April 10 after reactor pressure was increased
| |
| '-
| |
| to about 500 psig, the "B" RHR Rosemount flow transmitter indicated
| |
| an increase in pressure was occurring in the RHR system. Normal
| |
| system pressure is 105 psig controlled by the keepfill system via a
| |
| connection from the condensate transfer system. Though no pressure
| |
| indicator exists for the segment of piping in question, the flow
| |
| transmitter was determined to be pressure sensitive; as pressure in-
| |
| creases the "B" RHR flow indicator is driven negative. This anomaly
| |
| was substantiated by the inspector by observation, discussion with
| |
| control room operators, and by a contact made with the meter manufac-
| |
| turer, Rosemount, Inc. The flow transmitter reading is charted in
| |
| the control room and thus a permanent record exists that records all
| |
| of the pressurization events that have occurred. At approximately
| |
| 11:00 a.m. the "RHR Discharge or Shutdown Cooling Suction High
| |
| Pressure" alarm (hereafter referred to as RHR Hi alarms) sounded,
| |
| indicating pressure of about 400 psig. This alarm had been frequent-
| |
| ly received in the past. The alarm response procedure requires the
| |
| operator to diagnose the source of the leakage and to depressurize
| |
| the system by opening valves that lead to the tcrus. A number of
| |
| closely spaced RHR Hi alarms were subsequently received, approximately
| |
| once every fifteen minutes. In the afternoon the alarms continued
| |
| to be received, approximately once every half hour. During this time
| |
| the unit was placed on the line as the normal startup continued.
| |
| During the afternoon and into the night maintenance personnel at-
| |
| tempted to control the RHR leakage by increasing the closing torque
| |
| on Valve 298, this valve was diagnosed as the leaky valve because the
| |
| outboard piping was hot to the touch. The valve torque was increased
| |
| three or four times, each time the second MOV (288) was closed and
| |
| -29B opened then torqued closed, after which 28B was reopened. Ad-
| |
| justing-the closure torque to its highest allowable design value had
| |
| no affect on valve leakage.
| |
| At 2:15 p.m. on Friday, April 11, a decision was made by the operating
| |
| staff to close valve 28B in an attempt to stop the leakage. This
| |
| required that the plant enter a seven day LCO. Several hours after
| |
| the closure of 28B the RHR Hi alarm again sounded indicating that the
| |
| leakage continued. As both the MOV's appeared to leak the possibility
| |
| of violating containment integrity forced the operating staff to de-
| |
| clare an Unusual Event and start a slow, controlled shutdown of the
| |
| plant. The shutdown was subsequently speeded up after discussions
| |
| with NRC and by the next morning the reactor was in cold shutdown.
| |
| . __ ., _ __
| |
| | |
| .
| |
| '
| |
| 21
| |
| 5.2 RHR Isolation Valve Descriptions
| |
| Three valves form the isolation barrier between the high pressure
| |
| reactor coolant and the low pressure piping of each of the two RHR
| |
| loops. Figure 5.1 shows a schematic of the RHR loop B isolation
| |
| valves. The three isolation valves are:
| |
| -
| |
| Valve 688
| |
| A Rockwell, 900 lb., 18" testable, tilting disc, 316ss, check
| |
| valve. The valve has been modified by the removal of the post-
| |
| tion indication and air cylinder that provided the testability
| |
| feature. Thus, the valve is now a simple check valve. The
| |
| valve was once required to undergo containment leak testing but
| |
| was taken off the Appendix J list via an exemption granted by
| |
| NRC in 1977.
| |
| -
| |
| Valve 29B
| |
| A 600 lb.,18" x 14", 316ss, gate valve rated for 1250 psig at
| |
| 586 degree F. The valve is operated by a limitorque motor opera-
| |
| tor controlled from the control room. The valve will automati-
| |
| cally open in the event of a LOCA in combination with a reactor
| |
| pressure less than 400 psig. The valve is interlocked with
| |
| valve 288 preventing the inadvertent opening of both valves when
| |
| reactor pressure is greater than 400 psig. The valve is required
| |
| to undergo local leak rate testing as part of the containment
| |
| integrity program.
| |
| -
| |
| Valve 28B
| |
| A 600 lb., 18", 316ss, globe valve with a plug type disc,
| |
| rated for 1250 psig at 586 degrees F. This valve is operated
| |
| by a limitorque motor operator controlled from the control
| |
| room. As with 29B,_it is sent an automatic open signal in the
| |
| event of a LOCA in combination with a reactor pressure of less
| |
| than 400 psig. The valve can be throttled by manual control
| |
| room operation for flow control purposes. The valve is required
| |
| to undergo local leak rate testing as part of the containment
| |
| integrity program (it has replaced valve 68B as the second iso-
| |
| lation valve for the purpose of containment integrity).
| |
| Design requires one of the two motor operated valves be closed
| |
| during normal standby. Valves 28A and 28B of the two loops were the
| |
| original valves to be kept closed. In early 1986, leakage past 28B
| |
| began causing RHR Hi alarms. At that time, the valve was judged to
| |
| ,
| |
| have remained within LLRT leakage limits by trending of past leak
| |
| tests, but to eliminate the RHR Hi alarms it was decided to operate
| |
| with 29B as the closed valve. This change took place on February
| |
| 26, 1986. The A loop valves were left as is with 28A being the
| |
| closed MOV in that loop.
| |
| ;
| |
| . - - .-
| |
| | |
| ..
| |
| *
| |
| 22
| |
| 5.3 Past System Leakage Experience
| |
| The RHR Hi' alarm indicative of RHR isolation valve leakage has been
| |
| received in the past. The most recent period in which RHR Hi alarms
| |
| were received prior to April 10 began on January 10, 1986 and con-
| |
| tinued through February 26, 1986 (the day that Valve 29B replaced
| |
| valve 288 as the normally closed valve). During this period the
| |
| average time between RHR Hi alarms was about 10 hours. From
| |
| February 27, 1986 through March 8, 1986 no alarms were received.
| |
| The reactor was then shutdown until April 10 as a result of leaks
| |
| found in the Head Spray and Reactor Level Instrumentation systems.
| |
| The inspector made a bounding calculation in attempt to estimate a
| |
| conservative value of the amount of leakage from the valves.by ob-
| |
| taining the time and quantity of torus water that had been trans-
| |
| ferred to the rad waste tanks for processing. Between January 10
| |
| and February 27 a total of 81,500 gallons of torus water had been
| |
| pumped. Since the reactor was at pressure for 47 days during this
| |
| period the estimated leak rate of between one and two gpm is calcu-
| |
| lated. As there are other sources of water draining to the torus,
| |
| e.g., HPCI turbine pump exhaust, this estimate should be considered
| |
| as a high bound for the leak rate. Also this calculation assumes
| |
| that torus level at the beginning and end of the period was the
| |
| same. Though there is a chance of error in this calculaticn, it
| |
| providet some evidence that the isolation valve leak rate was not
| |
| substantial during the January 10 - February 27 time period.
| |
| 5.4 History of RHR Vaive Refurbishment and Leak Testing
| |
| 5.4.1 Check Valve 68B
| |
| Valve 68B was disassembled and rebuilt in May-June 1984.
| |
| The inspector reviewed the rebuilding documentation and
| |
| interviewed the engineer and technicians that worked on
| |
| the valve. In addition, the valve design was reviewed to
| |
| judge if the internal components provided reliable support
| |
| of the disc. The visual inspections upon disassembly in-
| |
| dicated that the valve internals were acceptable, other
| |
| than a light lapping of the valve seating surfaces no
| |
| component degradation was noted. The cover nuts and bolts
| |
| were found not to be acceptable and were replaced. As the
| |
| valve is no longer on the Appendix J valve list no leak
| |
| test was performed. Only a visual bluing technique was
| |
| ;
| |
| used to assure that the seating surfaces were in full
| |
| contact. The fact that the valves as-found condition was-
| |
| generally acceptable and that the design of the valve disc
| |
| and hinge is substantial, with little chance of misposition-
| |
| ing of the disc, provided evidence to the inspector the
| |
| : Valve 68B would reliably perform its closure function. As
| |
| will be discussed in Section 5.5, a significant pressure
| |
| .
| |
| differential, which forces the seating surfaces together,
| |
| | |
| . ..
| |
| .
| |
| 23
| |
| is required to tightly seat the valve. However, without a
| |
| pressure differential, as is the normal case with one of
| |
| the MOVs closed, the check valve is likely to provide no
| |
| additional resistance to leakage flow.
| |
| 5.4.2 Globe' Valve 288
| |
| In early 1986, this valve was diagnosed as leaking. At
| |
| that time it was predicted that the valve leak rate was
| |
| ~-~ - ~ -
| |
| still within the 7.89 standard liters per minute (sla)
| |
| Appendix J limit for a single penetration. The local leak
| |
| rate testing of Valve 28B is as follows:
| |
| 1980 - As found 0.1 sim, as left 0.2 slm
| |
| 1982 - As found 0.5 sim, as left 0.5 slm
| |
| 1984 - As found 1.9 sim, as left 1.9 slm
| |
| Utility staff calculated that (assuming an exponential
| |
| trend) a valve leak of 4.8 slm would be predicted for early
| |
| 1986. It was then concluded that the valve, though leaking,
| |
| was still acceptable. No other mechanical problem with the
| |
| valve was identified with one exception; electrical main-
| |
| tenance personnel had found that the closing amperages of
| |
| the valve were not initially repeatable. This led to the
| |
| disassembly and inspection of the motor operator, no
| |
| abnormalities were found.
| |
| ,
| |
| 5.4.3 Gate Valve 298
| |
| This valve has a history of failing the local leak rate
| |
| testing (LLRT). The valve failed its LLRT on January 7,
| |
| 1984 and again on October 11, 1984. Based on an interview
| |
| with the valve maintenance contractor representative, the
| |
| valve has a design deficiency that makes it difficult to
| |
| .
| |
| '
| |
| maintain low leakage over a long period of use. The
| |
| distance between the bottom of the valve wedge and the
| |
| bottom of the valve housing is only 3/16 inch. As the
| |
| seating surfaces wear the' wedge must drop lower and with
| |
| enough wear will bottom out on the housing. During last
| |
| November the wedge was removed and the seating surfaces
| |
| built-up and ground smooth. The post maintenance LLRT
| |
| done on November 29, 1984 resulted in zero leakage, so at
| |
| that time the valve was leaktight. As indicated
| |
| previously this valve replaced valve 28B as the normally
| |
| closed valve on February 26, 1986. The valve has been
| |
| considered for replacement, but no hard schedule exists.
| |
| :
| |
| i
| |
| [
| |
| - , --.
| |
| | |
| .-
| |
| .
| |
| 24
| |
| 5.5 As-Found RHR Walkdown and Valve Leak Measurements
| |
| During the period between April 13 and April 19 extensive RHR system
| |
| walkdowns and isolation valve leak measurements were conducted. NRC
| |
| inspectors observed these activities. The following summarize the
| |
| findings of these efforts.
| |
| 5.5.1 System Walkdowns
| |
| An as-found visual inspection of the RHR "B" loop system
| |
| piping, components, and structural supports was planned.
| |
| The inspection was to determine if any adverse effects had
| |
| resulted from the isolation valve leakage or possible
| |
| water hammer events associated with depressurizing the
| |
| piping after the RHR Hi alarm annunciated or as a result
| |
| of recent events involving water hammer events of the head
| |
| spray piping. Evidence of overheating, overpressurization,
| |
| or piping / component movement was of most interest. The
| |
| utility staff's planning and conduct of the walkdown was
| |
| careful and detailed. Drywell, "B" RHR quadrant, and torus
| |
| room entries were made and potential defects for each pipe
| |
| segment, component, and support was documented. No defects
| |
| were identified which could be associated with thermal,
| |
| overpressurization, or component movement, thus it was
| |
| determined that no visual evidence existed suggesting any
| |
| adverse conditions as a result of the isolation valve
| |
| leakage.
| |
| 5.5.2 Water Leak Measurements
| |
| A special water leak rate test was designed that would
| |
| simulate reactor water pressure on the reactor side of
| |
| valve 688. The test was conducted to determine the amount
| |
| of water leakage associated with the as-found Valve condi-
| |
| tion. Thus the leak rate of the check valve 688, the
| |
| closed gate valve 29B, and the closed globe valve 28B, in
| |
| series with one another, was to be determined. In addi-
| |
| tion, the test continued until pressurization of the RHR
| |
| piping was achieved and the RHR Hi alarm sounded in the
| |
| control room. Each segment of piping between the isolation
| |
| valves and between valve 28B and the RHR pump check valve
| |
| was monitored for pressure.
| |
| The test was conducted on April 17, 1986. The normally
| |
| locked open manual valve, 338, near the reactor was closed
| |
| and water pumped betweei it and valve 688. Table 5.4 sum-
| |
| marizes the test results as recorded by the NRC inspector.
| |
| The pressure between valves 33B and 68B was increased to
| |
| 300 psig. The technician found it difficult to hold pres-
| |
| sure constant. At one point the pumping was stopped com-
| |
| pletely for several minutes and then varied between 10 and
| |
| | |
| _
| |
| . :-
| |
| -.-
| |
| 25
| |
| 20 strokes / minute. An air operated positive displacement
| |
| water pump was used. It became apparent to the inspector
| |
| that as the technician increased pressure, valve 688 became
| |
| an effective barrier until such time as the pressure dif-
| |
| ference across the valve equalized, after which the valve
| |
| had no affect. The pressure was increased to 600 psig and
| |
| then to 950 psig. It was held at 950 psig for 95 minutes
| |
| at which time the RHR Hi alarm sounded in the control room.
| |
| During_this period the pump flow rate, on average, was
| |
| _
| |
| about 1/2 gpm.
| |
| 5.5.3 Appendix J Measurements
| |
| After the water test was concluded, the RHR piping was
| |
| drained and the air testing was conducted on April 18, 1986
| |
| for each of the three valves. Even though an LLRT is not
| |
| required for 68B an informational test was conducted. The
| |
| following are the LLRT results of the as-found valves:
| |
| 688 - 76 standard liters per minute
| |
| 298 - 1.0 standard liters per minute
| |
| 288 - 1.5 standard liters per minute
| |
| 5.6 Root Cause and Safety Significan e
| |
| The root cause of the RHR Hi pressure alarms was an approximate 1/2
| |
| gpm water leak past the loop "B" RHR isolation valves in conjunction
| |
| with apparently relatively leak tight RHR pump discharge check
| |
| valves. This condition caused a build up in pressure in the inter-
| |
| vening pipe segments to about 390 psig resulting in the alarm. There
| |
| was no indication that the potential for sudden failure of all three
| |
| isolation valves and resultant sudden overpressurization of the RHR
| |
| piping existed.
| |
| Prior to the decision to-shutdown, the operating staff could not know
| |
| the extent of isolation valve leakage and their decision to shutdown
| |
| was a good one. Though the low leak rate of the isolation valves
| |
| poses no safety problem, the inability of the operating staff to
| |
| determine significance due to instrument and procedural inadequacies
| |
| should be addressed. In addition greater utility attention should
| |
| be focused on the isolation function of valves that protect low
| |
| pressure ECCS systems from the high pressure reactor coolant.
| |
| 5.7 Conclusions and Recommendations
| |
| The licensee did a thorough job in evaluating the LPCI injection valve
| |
| leakage and recurring RHR pressurization events. The low leak rates
| |
| which were measured do not pose a safety problem. However, continued
| |
| power operation with the recurring pressurization of the RHR piping
| |
| and the resultant RHR High Pressure Alarm is unsatisfactory because
| |
| 1) the operator's attention is frequently drawn to an alarm that has
| |
| | |
| --
| |
| .:
| |
| '
| |
| 26
| |
| uncertain and undefined operational / safety significance, and 2) the
| |
| excessive cycling of the two safety related isolation MOVs (valves
| |
| 348 and 368) used to vent the pressure to the suppression pool contri-
| |
| butes to premature wearout of these valves. The licensee should
| |
| eliminate the cause of the recurring pressurization of the RHR piping.
| |
| In addition, several areas were identified where improvements are
| |
| needed to. ensure the significance of similar events in the future can
| |
| be determined and/or minimized. These include: periodically verify-
| |
| ing that the LPCI injection check valve properly seats with a dif-
| |
| ferential pressure across the valve; installation of pressure monitor-
| |
| ing equipment on the RHR piping; and development of a method to quan-
| |
| titatively measure the LPCI injection valve _ leakage during reactor
| |
| operation. Licensee activities in this area will be reviewed in
| |
| future inspections (86-17-04).
| |
| <
| |
| | |
| .
| |
| .
| |
| 27
| |
| 6.0 OVERALL SUMMARY AND CONCLUSIONS
| |
| The AIT reviewed three recent operational problems at Pilgrim: 1) the
| |
| spurious group-one primary containment isolation on April 4 and 12,1986,
| |
| 2) the failure of the main steam line isolation valves to open after the.
| |
| 1solations, and 3) recurring pressurization events in the residual heat
| |
| removal (RHR) system.
| |
| The team noted that the licensee's problem solving approaches were
| |
| carefully structured and appeared thorough. In addition, the team drew
| |
| -~~
| |
| the following conclusions for the three areas of concern:
| |
| --
| |
| No root causes for the spurious primary containment isolations on
| |
| April 4 and 12, 1986 were identified during the. inspection period,
| |
| despite considerable licensee effort. The team did not identify any
| |
| weaknesses in the licensee's problem solving approach.
| |
| --
| |
| The failure of the outboard main steam line isolation valves (MSIV)
| |
| to re-open following the containment isolations on April 4 and 12
| |
| was caused by partial or complete mechanical separation of the valve
| |
| pilot poppets from the MSIV valve stem assemblies. Pilot poppet set
| |
| screws did not prevent the poppets from unscrewing from the stem
| |
| assemblies.
| |
| --
| |
| The RHR pressurization events reflect slow leakage (about 0.5 gpm)
| |
| past a check valve and two motor operated injection valves in the
| |
| "B" RHR loop. Lack of RHR pressure instrumentation and the lack of
| |
| periodic tests of the RHR injection check valves inhibit a more
| |
| thorough diagnosis. No apparent RHR valve failure mechanism has
| |
| been identified as the reason for this leakage.
| |
| --
| |
| The licensee's conduct of the reactor shutdown on April 11 and 12,
| |
| 1986, was prudent in light of the recurring RHR pressurization
| |
| events.
| |
| The licensee's root cause evaluations were not completed and corrective
| |
| actions were not finalized during the AIT inspection. NRC review of
| |
| these actions should be conducted prior to startup from this outage.
| |
| Based on the AIT review, the first four items in CAL No. 86-10 have been
| |
| completed. The fifth and final item will be closed when the licensee
| |
| submits a written report on the three areas of concern to the Regional
| |
| Administrator and the Administrator authorizes reactor restart.
| |
| | |
| . . -
| |
| ''
| |
| 23
| |
| TABLE 5.1 - EVENTS OF THURSDAY APRIL 10, 1986
| |
| Plant RHR System
| |
| Time Conditions Conditions Comments
| |
| 0246 Started pulling RHR in standby with Reactor
| |
| rods cross connect open startup
| |
| between A & B loops. begins
| |
| Pressure at-105 psig
| |
| provided by keepfull
| |
| system.
| |
| 0345 Critical
| |
| 0700 300 psig
| |
| 11% steam flow
| |
| 1000 500 psig RHR flow chart in-
| |
| 12% steam flow dication showing
| |
| pressure rise in RHR
| |
| piping.
| |
| 1100 660 psig RHR Hi alarm; 6 alarms First alarm
| |
| 12% steam flow once every 15 mins. indicated on RHR
| |
| flow chart, no log
| |
| entry.
| |
| 1200 >900 psig Reactor at
| |
| 12% steam flow pressure.
| |
| 1300 Turbine Rolling RHR Hi alarm; 4 alarms,
| |
| once-every 30 mins.
| |
| 1336 Unit on Line
| |
| 1500 STA log -
| |
| indicates look-
| |
| ing into valve
| |
| '29B leakage
| |
| 1600 NWE Log (1600 to
| |
| 2400) -
| |
| maintenance is
| |
| torquing up
| |
| valve 29B
| |
| 1800 RHR Hi alarm; 4 alarms,
| |
| once every 30 mins.
| |
| | |
| _. _
| |
| .
| |
| *
| |
| 29
| |
| Plant RHR System
| |
| Time Conditions Conditions Comments
| |
| 2200 No RHR Hi alarms
| |
| between 2200 and
| |
| 0200
| |
| 2400 Reactor near
| |
| 100's steam flow
| |
| _
| |
| . -_- .- ._ - _.
| |
| . . _ - . . - - . . - - .
| |
| | |
| . _. . _ . -_ ._.__. _ .. . . _ _ __.. .._ _ .__ . __
| |
| l
| |
| l
| |
| *
| |
| 30
| |
| TABLE 5.2 - EVENTS OF FRIDAY APRIL 11, 1986
| |
| Plant RHR System
| |
| Time Conditions Conditions Comments
| |
| 0200 RHR flow test Checks opera-
| |
| bility of all
| |
| four RHR pumps
| |
| 0219' RHR Hi alarm First notation
| |
| of RHR Hi alarm
| |
| found in control
| |
| room log
| |
| 0315 RHR Hi alarm
| |
| 0336 "B" RHR loop in torus Pressurization
| |
| cooling mode of RHR pre-
| |
| vented when loop-
| |
| open to torus
| |
| 1115' RHR secured from
| |
| torus cooling
| |
| 1158 RHR Hi alarm
| |
| 1415 RHR Hi alarm; valve Declared LPCI
| |
| 288 closed,'both loop "B" inoperative
| |
| MOVs (28B & 24B)
| |
| no closed
| |
| 1653 RHR Hi alarm
| |
| 1710 Initiated a con- Declared an
| |
| trolled shutdown Unusual Event,
| |
| steam flow decrease notified NRC
| |
| rate of 5% per hour
| |
| 2000 960 psig, steam
| |
| flow decrease rate
| |
| increased to 30%
| |
| per hour
| |
| 2200. 930 psig, 33%
| |
| steam flow
| |
| 2215 RHR Hi alarm Notified NRC
| |
| | |
| .
| |
| *
| |
| 31
| |
| TABLE 5.3 - EVENTS OF SATURDAY, April 12, 1986
| |
| Plant RHR System
| |
| Time Conditions Conditions Comments
| |
| 0030 Turbine off line
| |
| _ _ _ __.
| |
| 0136 Out of run mode
| |
| 0200 HPCI in recir- Initiate torus cooling
| |
| culation mode mode of RHR
| |
| for reactor
| |
| pressure control
| |
| 0215 Significant
| |
| pressure
| |
| reduction begins
| |
| 0400 <100 psig
| |
| 0645 Out of torus cooling
| |
| RHR loop A placed in
| |
| shutdown cooling mode
| |
| 0908 Reactor Secured from
| |
| <212 degrees F Unusual Event
| |
| | |
| O
| |
| ~
| |
| 32
| |
| Table 5.4
| |
| Summary of Water Leak Test Data Recorded By Inspector
| |
| April 17, 1986
| |
| Approximate
| |
| Pump Strokes Pressure Between Valves, PSIG
| |
| Per Minute
| |
| ~--
| |
| - Time 338/68B 688/298 298/28B 28B/ Pumps Comments
| |
| ~ 1500 0 22 25 65 104
| |
| ~ 1510 0-20 300-500 290 100 104 5 min after
| |
| reaching 300
| |
| ~ 1520 0-20 600-700 575 330 145 10 min after
| |
| reaching 600
| |
| 1540 0-20 950 - - -
| |
| 1606 4-8 950 950 700 185
| |
| 1715 4.75 975 975 725 375 to 380 RHR Hi Alarm
| |
| received
| |
| Note: 4.75 pump strokes per minute is equivalent to ~ gpm.
| |
| | |
| -
| |
| 33
| |
| FIGURE 3.1
| |
| -
| |
| PCIS INITIATION LOGIC FOR CHANNEL A-1
| |
| (Typical of Channels A-2, B-1 and B-2)
| |
| ~
| |
| h
| |
| l -
| |
| i
| |
| <
| |
| d . 16 A - K4 A (.OPEN ON .-
| |
| 5 A-S1 ( REACTOF. MODE
| |
| V
| |
| M SL Lo Priss,4980 psi swi7cp : ray eAss
| |
| $ STM, LINT LO. PR TRIP;
| |
| * O PE W IN"RUN"
| |
| $ MODE ONLY)
| |
| '
| |
| W
| |
| , I 6 A - K I A ( OPErd ON \ 16 A-k !9 A ( O PFN ON M'
| |
| $ LO LO RX. WTR. LEVEL) Rx. W ATER
| |
| LE\E L #\
| |
| I ,
| |
| IGA- R44 A ( OPEN ON } ZZ
| |
| [
| |
| t
| |
| MSL Nl RAD.) J
| |
| ,
| |
| 16 A - k A COPEN Orl )
| |
| t'
| |
| ; HSL Hi T E M P.) b '
| |
| >
| |
| o . 16 A - K 3 A (OPEN Or]
| |
| 9 MSL H! FLOW)
| |
| 16 A - x '7 A ( 6 ROUP 1 PCI.5
| |
| INITI ATioN CHANNEL A.i RELAY
| |
| -
| |
| ,
| |
| C -
| |
| :
| |
| .
| |
| blO TE : RELAYS ARE NOR>1 ALLY C!)ERGl~Ely i BUT SHCw A) It)
| |
| 'DE E!, ER G l2 ED (SIDEL F) Co t3 DI T ION "
| |
| | |
| . _ _ _
| |
| -
| |
| 34
| |
| ~
| |
| FIGURE 3.2
| |
| REACTOR MODE SWITCH
| |
| !
| |
| . -
| |
| -
| |
| REACTOR MODE I
| |
| --
| |
| -
| |
| .-
| |
| '
| |
| .
| |
| X REFUEL
| |
| STA R T* {
| |
| ?
| |
| $44
| |
| 0
| |
| g
| |
| l HO T S TBY
| |
| .
| |
| m
| |
| 9
| |
| \ /
| |
| l
| |
| '
| |
| .
| |
| *
| |
| ca -RUN
| |
| l 'L , ,
| |
| :
| |
| .-
| |
| .
| |
| i
| |
| .
| |
| I
| |
| i .
| |
| o
| |
| | |
| ._._.-._. _ -- . _ _ _ .-. . . _ . _ . _ - _ - _ - _ _ _ _ _ _ _ _ _ _ _ _
| |
| . 35
| |
| i
| |
| FIGURE 3.3 !
| |
| l .
| |
| GENERAL ELECTRIC SB-1 MODEL CONTROL SWITCH l
| |
| l
| |
| l
| |
| l
| |
| l l
| |
| '
| |
| l
| |
| I I
| |
| l
| |
| l
| |
| i
| |
| .
| |
| vuiitacts numcerec for easy
| |
| *
| |
| Silver-to-silver contacts.
| |
| Circuit designation plate is
| |
| marked for easy identification -
| |
| of switch functions.
| |
| 4' i
| |
| !
| |
| .
| |
| l
| |
| .
| |
| ,
| |
| I Good selection of functional f l
| |
| handles. ,
| |
| I
| |
| f
| |
| l
| |
| l
| |
| l
| |
| Front support spaces body of
| |
| f switch % inch from rear of the
| |
| panel, allowing ample rocm for
| |
| inserting leads into the switch.
| |
| Escutcheon plates of per.
| |
| manent-finish molded material Protective cover (not shown) f
| |
| '
| |
| are neat in appearance and completely covers all live
| |
| uniform in size. parts, meets NEMA 1 require-
| |
| ments for panel mounted
| |
| 80407s3 switches. i
| |
| Type SD-1 Switch With Cover Removed
| |
| !
| |
| L
| |
| | |
| . 36
| |
| FIGURE 4.1
| |
| ,
| |
| MAIN STEAM ISOLATION VALVE
| |
| .
| |
| I 1
| |
| I T
| |
| l *
| |
| (?
| |
| o
| |
| d"
| |
| ,E
| |
| -
| |
| -
| |
| M[ a;
| |
| AIR CYLINDER
| |
| N .. s
| |
| hp}? V HYORAULIC
| |
| N
| |
| '
| |
| , DASH POT
| |
| ~
| |
| HEllCAL $PRINC1
| |
| 5PmhG Gu1DE
| |
| '
| |
| ! SPEED COPITROL VALVF
| |
| ACTUATOR SUPPORT AND
| |
| h[ k;, ; $PRING CUl0E $ HAFT
| |
| SPRING SEAT WEW9ER
| |
| -
| |
| -
| |
| _
| |
| / ST EW PACKING
| |
| GL
| |
| l l
| |
| 'Jh LEAK OFF CONNECTION
| |
| BONNET BOLT 5
| |
| CLE ARA NCE BONNET
| |
| b ,
| |
| PILOT SPRING ~
| |
| j~Q R p(gg
| |
| _ "_
| |
| x t=)- /
| |
| Ni ; 1. -
| |
| \ : *
| |
| )
| |
| %
| |
| t ? /I? / POPPI t(PLUG WAIN 0isk) .
| |
| / ~
| |
| >
| |
| uAiN vatvt 5f at
| |
| '
| |
| ,
| |
| /
| |
| \ Pit 0i 1( At
| |
| Pot ot
| |
| '
| |
| /
| |
| ,
| |
| L
| |
| | |
| -
| |
| .
| |
| 37
| |
| .
| |
| FIGURE 4.2
| |
| . MSIV pit.0T POPPET ASSEMBLY ,
| |
| l
| |
| .
| |
| m- ]
| |
| .
| |
| *
| |
| i
| |
| I
| |
| . ) {,
| |
| .
| |
| - ,
| |
| .s
| |
| F7 >
| |
| $
| |
| h jl }-l .
| |
| -
| |
| }
| |
| -
| |
| l 'o :s-
| |
| g% k x. '
| |
| l
| |
| TI
| |
| ,
| |
| A
| |
| : N
| |
| :
| |
| ~
| |
| n h . \'N . '
| |
| !@
| |
| ~ l
| |
| l
| |
| !AMl,}yn[,T
| |
| NW
| |
| '; k I
| |
| w-
| |
| i
| |
| "-
| |
| n:^4
| |
| t)k.g -
| |
| ..,
| |
| _ ,,/,
| |
| g' g - - . . . . . .
| |
| '
| |
| ,
| |
| ,
| |
| , . ;,;,
| |
| ,.
| |
| ,
| |
| .
| |
| < ,
| |
| j
| |
| ,. 3' n,.s s
| |
| ,.
| |
| .. 9l _.. lg'/d ''
| |
| j
| |
| .< i ,
| |
| i
| |
| i : t
| |
| ,,
| |
| ''
| |
| .e .. w o- ,
| |
| ~,
| |
| k
| |
| 3': .
| |
| l'1
| |
| : Gr l -{1' ! (.. ,
| |
| ,
| |
| , :
| |
| ! e-~ .
| |
| 4
| |
| &
| |
| ! %
| |
| ty
| |
| f'f5
| |
| :
| |
| 'l
| |
| :
| |
| er I
| |
| J9
| |
| L
| |
| *
| |
| i-
| |
| 1
| |
| g
| |
| ! [1
| |
| G I
| |
| .O
| |
| I
| |
| s''
| |
| g= - STEM
| |
| -
| |
| M .
| |
| l
| |
| .
| |
| efy JB <
| |
| k's'
| |
| "
| |
| - '
| |
| a. i [Q %:
| |
| '' #
| |
| sE+s
| |
| '
| |
| . .K.( g' ~3
| |
| I , N, _
| |
| ,__.
| |
| j.',s PILOT POPPET
| |
| s
| |
| gsgj L/ 'g'W # .
| |
| ,
| |
| ' NUT
| |
| SCREW ; gw ._ p%(% .
| |
| ~
| |
| '' ' s' ,
| |
| =
| |
| '
| |
| I 4' '' '
| |
| t ._-..
| |
| n
| |
| ig'2sa$f9....tti4 '4 '1 8$$$.$
| |
| i
| |
| .
| |
| gMMkW
| |
| SPLIT RING MAIN POPPET
| |
| PILOT P0PPET
| |
| < 1
| |
| | |
| 38
| |
| .
| |
| FIGURE 5.1
| |
| . SIMPLIFIED DIAGRAM 0F RHR LOOP B
| |
| .
| |
| .,
| |
| '
| |
| To
| |
| t*Yuna_s ^
| |
| 2EE(To,? SfRAT
| |
| 9tEnde ro oreNS ' ~N0
| |
| #
| |
| sg u el 5'Y# 788
| |
| $ oNc6
| |
| -nb
| |
| l
| |
| .ns #8
| |
| 'A N
| |
| */013
| |
| RECstC
| |
| vr
| |
| Cross ~ Tor ,
| |
| To
| |
| '# #
| |
| N Loo ? 'h'
| |
| $(-nB
| |
| Mots: M L L. J4Lur NuessR S
| |
| ART P RFCE 2, cc thy " loof ** [#M MNfatR
| |
| A$'
| |
| *tS O '191 &
| |
| .. kua
| |
| 7e R w ', -(,n E- /:9
| |
| 292 2%R
| |
| fumP i Pump
| |
| y .o- ) 'B
| |
| Stok
| |
| sum,
| |
| gg MS /G C001143%
| |
| SutTrov
| |
| -
| |
| - -
| |
| - 7B
| |
| L
| |
| | |
| ,. ' **..,' UNITCl3 S T AT cs
| |
| ** ',<
| |
| ''
| |
| [ '* ,
| |
| ( NUCt.l AH REGUI.ATORY COMMISSION ATTACHMENT 1
| |
| -
| |
| , j' ncGION I
| |
| hJ1 F ARK AVLNUI
| |
| _, 5; p
| |
| *s f KING 08 Puusst A. PENNSYL VANIA 194(4
| |
| ....+
| |
| April 12,1986
| |
| CAL No.: 86-10
| |
| Docket Nu@ce: 50-293
| |
| Bosten Edison Company M/C Nuclear
| |
| ATTN: Mr. William D. Harrington
| |
| Senior Vice President, Nuclear
| |
| 800 Boylston Street
| |
| Boston, Massachusetts 02199
| |
| Gent.lcmen:
| |
| Subject: Confirmation of Actions to be Taken with Regard to the Pilgrim
| |
| Plant Events Which Occurred on April 11-12, 1986
| |
| Pursuant to our telephone conversation on April 12, 1986 with Mr. Oxsen it is
| |
| our understanding that you have taken or will take the following actions:
| |
| 1. Maintain all af fected equipment related to the events which occurred
| |
| on April 11-12, 1926 in its as-found condition-(except as
| |
| nu essary to maintain the plant in a r,afe >liutdown t.undition) In
| |
| order to preserve any evidence which would be needed to inspect
| |
| or reconstruct the events.
| |
| 2. Deveinp troubleshooting plans and procedures and provide those to
| |
| the NRC Augmented Inspection Team (Ali) for their review and
| |
| comment prior to initiating any troubleshooting of the affected
| |
| equipment.
| |
| 3. Advise the AIT leader prior to the conduct of any troubleshooting
| |
| ar,tiv ities. *
| |
| 4. Make available to the NRC AII relevant written material related to
| |
| previous problems with the affected equipment.
| |
| 5. Provide a written report to the ftegional Administrator prior to
| |
| restart that contains your evaluatlon of the following:
| |
| a. Intersystem leakage through the motor-operated injection
| |
| valves (including the check valve) of the residual heat
| |
| removal system;
| |
| '
| |
| b. The primary containment isolation which occurred daring
| |
| shutdown af ter the reactor mode switch was repositioned
| |
| from the run mode to the startup mode;
| |
| f _
| |
| r I r1 ~d Il
| |
| h b y ~/ ' I
| |
| | |
| i
| |
| *
| |
| 2
| |
| .
| |
| ..
| |
| l
| |
| C. The failure of the outboard main steam isolation valves' to
| |
| reopen after resetting the primary containment isolation
| |
| signal.
| |
| This report should include the underlying causes for the above
| |
| noted events, an assessment of their relationship to previous events
| |
| including the events of April 4, 1985, corrective actions taken and
| |
| your basis for restart, including the criteria used and your analyses
| |
| associated with these criteria.
| |
| Further we understand that restart will not occur until you receive authoriza-
| |
| tion from the Regional Administrator.
| |
| If your understanding of the actions to be taken are different than those
| |
| described above, please contact this of fice within 24 hours of the receipt of
| |
| this letter.
| |
| Thank you for your cooperation.
| |
| Sincerely,
| |
| .
| |
| Thomas E. Hurley
| |
| Regional Administrator
| |
| cc: L. Oxsen, Vice President, Nuclear Operations
| |
| C. J. Mathis, Station Manager
| |
| Joanne Shotwell, Assistant Attorney General
| |
| Paul Levy, Chairman, Department of Public Utilities
| |
| Plymouth Board of Selectmen
| |
| Plymouth Civil Defense Director
| |
| Senator Edward P. Kirby
| |
| Public Document Room (PDR)
| |
| local Public Document Room (LPDR)
| |
| Nuclear Safety Information Center (MSIC)
| |
| NRC Resident Inspector
| |
| Commonwealth of Massachusetts (2)
| |
| '
| |
| .
| |
| | |
| A
| |
| ,
| |
| o
| |
| ATTACHMENT 2
| |
| PERSONS CONTACTED
| |
| The following is a partial listing of the licensee personnel that were
| |
| contacted during the inspection.
| |
| W. Harrington, Senior Vice President, Nuclear
| |
| L. Oxsen, Vice President, Nuclear Operations (Senior Licensee Manager Present
| |
| at the Exit Interview)
| |
| C. Mathis, Nuclear Operations Manager
| |
| P. Mastrangelo, Chief Operating Engineer
| |
| K. Roberts, Director Outage Management
| |
| N. Brosee, Maintenance Section Head
| |
| T. Sowdon, Radiological Section Head
| |
| J. Seery, Technical Section Head
| |
| E. Ziemianski Management Services Section Head
| |
| S. Wollman, On-Site Safety and Performance Group Leader
| |
| R. Sherry, Chief Maintenance Engineer
| |
| E. Graham, Compliance and Administrative Group Leader
| |
| P. Smith, Chief Technical Engineer
| |
| W. Clancy, Nuclear Engineer, FS and MC Group Leader
| |
| T. McLoughlin, Nuclear Operations Sr. Electrical Engineer
| |
| A. Morisi, Operations Assistant to Director of Outage Management
| |
| | |
| .
| |
| ,
| |
| o
| |
| ATTACHMENT 3
| |
| Tests / Checks Performed During Mode Switch /PCIS Investigation
| |
| The licensee performed the following tests / checks of tne PCIS components,
| |
| including the reactor mode switch. The mode switch testing was performed
| |
| in all four mode positions under various human factor scenarios i.e.,
| |
| with and without key removed, pulling up or pushing down while turning
| |
| _ _
| |
| the mode switch, etc.
| |
| -
| |
| Surveillance Test Procedure 8.M.2-1.5.3.1, 2, 3, and 4 Primary Con-
| |
| tainment Isolation Logic Channel Test - Channels A-1, A-2, A-3, A-4,
| |
| respectively Revision 6; performed on April 14, 1986.
| |
| -
| |
| Inspection of contacts of the PCIS relays in Channels A-1, A-2, B-1
| |
| and B-2, in accordance with Procedure 3.M.3-8, Inspection / Trouble
| |
| Shooting - Electrical Circuits, Revision 6, performed on April 14,
| |
| 1986, along with the above 4 PCIS Logic Tests.
| |
| -
| |
| Surveillance Test Procedure 8.M.1-19, Reactor Water Level (RPS/PCIS),
| |
| Revision 13; performed on April 15, 1986. (While performing this
| |
| test, an inadvertent closure of the MSIVs and steam line drain valve
| |
| M0-220-2 occurred)
| |
| -
| |
| Trouble Shooting Procedure for.the investigation of inadvertent
| |
| closure of MSIVs and M0-220-2 during performance of the above Sur-
| |
| veillance Test Procedure (8.M.1-19) on April 15, 1986; performed in
| |
| accordance with procedure 3.M.3-8 on April 15, 1986.
| |
| -
| |
| Surveillance Test Procedure 8.M.2-1.4.4, Main Steam Line Low Pressure,
| |
| Revision 5, performed on April 16, 1986.
| |
| -
| |
| Trouble Shooting Procedure to check out the AC and DC solenoid circuits
| |
| of the MSIVs, performed on April 17, 1986.
| |
| -
| |
| Temporary Procedure TP86-59, Mode Switch Test for Steam Line Low
| |
| Pressure Bypass, Revision 0; performed on April 19, 1986.
| |
| -
| |
| Trouble shooting procedure 3.M.3-8 to check out the effect of vibra-
| |
| tion on reactor vessel level Yarway level indicating switches;
| |
| performed on April 21, 1986.
| |
| -
| |
| Trouble shooting procedure 3.M.3-8 to confirm the vibration effect
| |
| observed during the above test; performed on April 21, 1986.
| |
| -
| |
| Trouble shooting procedure 8.M.1-19 to investigate the cross charnel
| |
| interaction of relays suspected during the performance of the above
| |
| two' tests; performed on April 21, 1986.
| |
| | |
| O
| |
| e
| |
| 0
| |
| -
| |
| Trouble shooting procedure 3.M.3-8 to investigate the vibration / cross
| |
| channel interaction observed as the April 21, 1986 testing; performed
| |
| on April 23, 1986.
| |
| -
| |
| Trouble shooting procedure 3.M.3-8 to check out the contact resis-
| |
| tances of the relays in the PCIS trip circuitry, performed on
| |
| April 23, 1986.
| |
| -
| |
| Surveillance test procedure 8.M.2-1.4.3, Main Steam Line High Flow,
| |
| Revision 1; performed on April 24, 1986.
| |
| -
| |
| Surveillance Test Procedure 8.M.1-12, Main Steam Line High Radiation,
| |
| Revision 11; performed on April 24, 1986.
| |
| -
| |
| Temporary Procedure TP 86-68, Mode Switch Resistance, Revision 0;
| |
| performed on April 24, 1986.
| |
| -
| |
| Trouble shooting procedure 3.M.3-8 to check out loose wire in the
| |
| PCIS circuitry and the RPS grounding connection; performed on
| |
| April 24, 1986.
| |
| A
| |
| | |
| o-
| |
| t
| |
| O
| |
| ATTACHMENT 4
| |
| DOCUMENTS REVIEWED
| |
| Plant Design Change Request No. 83-48, "MSIV Refurbishment", dated
| |
| October 5, 1983
| |
| Atwood and Morrill Co. Inc., " Instruction Manual for 20" Main Steam
| |
| Isolation Valves".
| |
| - ~'
| |
| Procedure No. TP 86-61, "MSIV Plot disassociation Test", Revision 0,
| |
| dated April 17, 1986
| |
| Procedure No. 2.2.92, " Main Steam Line Isolation and Turbine
| |
| Bypass Valves", Revision 15, dated May 8, 1985
| |
| Procedure No. 8.7.4.4, "MSIV Trip", Revision 12, dated January 30, 1986
| |
| i
| |
| }}
| |