IR 05000354/1986041
| ML20215F619 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 09/24/1986 |
| From: | Norrholm L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20215F617 | List: |
| References | |
| 50-354-86-41, NUDOCS 8610160291 | |
| Download: ML20215F619 (24) | |
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U. S. NUCLEAR REGULATORY COMMISSION
REGION I
50354-860808 Report No.
50-354/86-41 Docket 50-354 License NPF-57 Licensee:
Public Service Electric and Gas Company Facility:
Hope Creek Generating Station Conducted:
August 13, 1986 - September 2, 1986 Inspectors:
P.. W. Borchardt, Senior Resident Inspector D. JA F1 rek, Lead Reactor Engineer J.p.}Pe1, a tor Engineer Approved:
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L. $oJ % olm, Chief Reactor, Projects
/ datd SeYtion 2B Inspection Summary:
Inspection on August 13, 1986 - September 2, 1986 (Inspection Report Number 50-354/86-41)
Areas Inspected: Special onsite inspection by the NRC resident and region based inspectors of the causes for the inoperability of the reactor building to suppression chamber pressure relief system.
This inspection involved 93 hours0.00108 days <br />0.0258 hours <br />1.537698e-4 weeks <br />3.53865e-5 months <br /> by the inspectors.
Results: The inoperability of the reactor building to suppression chamber
' pressure relief system was identified as an apparent violation of Technical Specifications. Although this condition was licensee identified, it is viewed as significant because of the duration that this discrepancy existed and because a system which ensures primary containment integrity after certain postulated accident conditions was inoperable.
8610160291 861006 DR ADOCK 05000354 PDR
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DETAILS 1.
Overview At 11:45 a.m. on August 8, 1986, a reactor shutdown was commenced and an unusual event declared after it was determined by the licensee's staff that'the reactor building to suppression chamber pressure relief system was inoperable and the plant was operating in violation of Technical Specifications. Technical Specification (T.S.) 3.6.4.2 does not allow plant operations in operational conditions 1, 2 or 3 with both Reactor Building to Suppression Chamber vacuum breaker assemblies inoperable, and therefore the plant was shut down as required by T.S. 3.0.3 which states that when a Limiting Condition for Operation is not met, except as provided in the associated action requirements, within one hour action shall be initiated to place the unit in an Operational Condition in which the specification does not apply. The Unusual Event was terminated at 3:35 p.m. when the unit entered operational condition 3 (Hot Shutdown).
The unit entered operational condition 4 (Cold Shutdown) at 7:54 p.m..
The subsequent investigation into this event determined that a design drawing error made in 1983 during plant construction caused this system to be inoperable.
In the event that a vacuum was created in the suppression chamber, the butterfly isolation valves in series with each -
vacuum breaker would have remained shut, which would have prevented the vacuum breakers frem fulfilling their design safety function. This condition had existed since 1983, and remained undetected until August 8, 1986.
A design engineer's apparent misunderstanding of the reactor building to suppression chamber pressure relief system resulted in the reversed installation of the pressure differential transmitters intended to sense a vacuum in the suppression chamber. The as-installed configuration would have resulted in the 24 inch butterfly valve opening on high torus pressure rather than on high torus vacuum. A review of the construction program design controls identified no other similar discrepancies. A review of post construction activities indicates that once the original design error was made, no subsequent test would have identified the design error because both preoperational testing as well as surveillance testing isolated the sensing lines while a controlled pressure was applied directly to the pressure differential transmitter.
Despite the conditions described above, the ability of the 24 inch butterfly isolation valves to be manually operated from the control room was not affected.
This report provides a description of the reactor building to suppression chamber pressure relief system, discusses how the problem was identified, examines how the design error was made, and evaluates if other programs should have identified the deficiency earlier.
The licensee's corrective actions are also discussed.
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System Description The reactor building to suppression chamber pressure relief (or vacuum breaker) system is designed to limit the presssure differential between the reactor building and the suppression chamber to less than 3.0 psid.
This safety feature is intended to protect the suppression chamber from negative pressure loading in the event containment spray is inadvertently initiated after a loss of coolant accident. The reactor building to suppression chamber pressure relief (RBSCPR) system consists of two 24-inch independent vacuum relief assemblies, each providing a relief path from the reactor building air space to the suppression chamber. Each vacuum relief assembly consists of an outboard mechanical check valve and an inboard, normally closed, air operated butterfly valve. The two 24-inch mechanical check valves are designed to be fully open at 0.25 psid.
A pressure differential transmitter senses the pressure difference between the reactor building and the suppression chamber. When this
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differential reaches a specified limit, a pressure differential switch provides an input signal to a solenoid valve, associated with the butterfly valve, which energizes and directs gas from the primary contain-ment instrument gas system (PCIGS) to open the butterfly valve. The path through the check and butterfly valves allows the reactor building
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atmosphere to enter the suppression chamber and equalize the pressure di f ferent.e. When the pressure differential is reduced to a specified level, the solenoid valve de-energizes and the butterfly valve returns to the closed position. The check valve prevents the suppression chamber atmosphere from venting into the reactor building should the butterfly valve fail to'close. Valve position switch contacts are monitored on each butterfly valve and each check valve, providing fully open and fully closed valve position indication in the main control room.
i 2.1 RBSCPR System Testability The RBSCPR system controls and instrumentative are capable of being tested from sensors through actuating devices during.ormal power operation.
Each sensor (pressure differential transmitter) can be valved out of service and functionally tested using an appropriate test source. This test verifies proper circuit operation from the transmitter input through the actuating device but does not test sensing line operability.
The calibration of the alarm units fpressure differential switches) can be checked from the appropriate cabiaet in the control equipment room without initiating operation of the actuating device. When an alarm unit is placed "in test", an output is provided to illuminate an indicating light in the main control room to advise the control room operator of the
"in test" status. This indicating light is automatically extinguished when the alarm unit is placed back in operatio _
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Each check valve and butterfly valve in the RBSCPR system can be individually tested from the main control room by the operation of an associated pushbutton test switch. When the pushbutton test switch is depressed, the solenoid valve associated with the valve being tested is energized and directs gas from the PCIGS to open the butterfly or check valve. When the pushbutton test switch is released, the solenoid valve de-energizes and the valve under test returns to its normal operating status. Satisfactory operation is determined by observation of the expected valve position indicating light patterns during the test.
2.2 Design Bases As stated earlier, the RBSCPR system is designed to limit the presssure differential between the reactor building and the suppression chamber to less than 3.0 psid.
The negative pressure differentials (negative corres-ponding to an inward loading) across the drywell and suppression chamber walls, as analyzed in the FSAR, are caused by the following events:
cooling cycles;
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inadvertent containment spray actuation during normal operation; and,
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steam condensation following reactor coolant system (RCS) pipe ruptures with inadvertent containment spray actuation.
Cooling cycles result in minor pressure transients in the drywell, which occur slowly and are normally controlled by heating and ventilating equip-ment.
Inadvertent spray actuation during normal operation results in a more significant pressure transient and becomes important in sizing the suppression chamber-to-reactor building vacuum breaker assemblies.
Steam condensation following RCS pipe ruptures with inadvertent containment spray actuation within the drywell results in the most severe pressure transients.
Following an RCS rupture, the drywell atmosphere is purged to the suppression chamber free space, leaving the drywell full of steam.
Subsequent condensation of the steam in the drywell can be caused either by ECCS spillage from the rupture or by inadvertent containment spray actuation following a LOCA.
Pressure transients within the drywell and the suppression chamber free space, due to inadvertent containment spray actuation for post-LOCA steam condensation, are evaluated in the FSAR. The results of containment
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depressurization transients are provided in FSAR Table 6.2-29.
Details of l
the limiting transient, including the analytical models, assumptions, and methods used, are provided in FSAR Appendix 6A.
In response to this event the licensee has performed an additional analysis which shows the maximum allowable external pressure on the drywell to be in excess of 4 psi. With both vacuum breaker assemblies inoperable and assuming the worst case scenario, the negative pressure
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f loading on the. primary containment was calculated to be approximately 3.4 psid. The licensee therefore concluded that primary containment-integrity would not have been compromised.
3.
Problem Identification
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i On July 4,L1986, the control room operators observed that the torus
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pressure instrument indicated a pressure of -0.6 psig.
The operators then took action to restore.the torus pressure to within the Technical Specification limits of -0.5 psig to 1.5 psig and incident report 86-119 i
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was. generated to document <this occurrence. As part of the incident report followup, the_ licensee investigated why the reactor building to suppression chamber vacuum breakers had not operated to relieve this presssure differential as designed. The presssure differential trans-mitters which control the inboard butterfly valves were verified to be within calibration.and the butterfly valves were shown to be free to
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' operate. Recent testing had also shown that the vacuum breaker check valves were operable.
It was not untti August 8, while performing a
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system review,- that the system engineer identified a problem with the butterfly valves. His review indicated that a problem existed in the valve control logic. Once it was established that the RBSCPR system was inoperable, the unit was. shut down to operational condition 3.
The in -
operability of the RBSCPR system during operational conditions 1, 2 and 3 is an apparent violation of Technical Specification Limiting Condition
.for Operation 3.6.4.2 (86-41-01).
On August 8, 1986, the Nuclear Safety Review Group was tasked to perform an independent ~ assessment of the causes for the RBSCPR system's inopera-t bility and to recommend corrective actions. The Safety Review Group i
directed that a complete system walkdown be performed, and during this evolution, the improperly installed pressure differential transmitter (PDT) tubing was identified. The high side of the PDT's were found to be connected to the torus and the low side to the reactor building, which is opposite of the arrangement required for proper system operation. The result of this error is that if a vacuum were drawn in the torus, the
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butterfly isolation valves would remain shut thereby preventing the vacuum j
breakers from relieving the pressure differential. Additionally, if a
_ positive pressure was generated in the torus, the butterfly valves would
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open causing the vacuum breakers to'be subjected to primary containment
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pressure.
The inspector has subsequently reviewed the local leak rate i-test data for the vacuum breakers and determined that the results were within specifications and that prin.ary containment leakage rates would have been acceptable even if the vacuum breakers were subjected to primary
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The system walkdown identified two other discrepancies that impacted system operability.
The PDT for one butterfly valve was found to be
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valved out of service. This would have prevented the valve from operating even if the PDT were installed correctly. Also, a sensing line for the
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same PDT had tape over its end, which may have impacted the PDTs ability
to accurately sense reactor building pressure.
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Although the system engineer's conclusion that a logic error was the cause for the system inoperability was not correct, it did serve to focus attention on the lack of control placed on the staple jumper configuration for Bailey 745 Alarm modules.
This is a separate issue from the subject of this report and will therefore be discussed in inspection report 50-354/86-40.
4.
Design and Construction Process The pressure differential transmitters (PDT 5029 and 5031), used to sense suppression chamber vacuum, were designed and installed in accordance with the architect-engineer's field engineering program. Under this program, the conceptual system design comes from Bechtel's area office in San Francisco (SFAO).
SFAO also approves and provides the instrument vendor prints and a " Design, Installation and Test Specification" (DITS) package for each major system. The DITS package provides a description of how each sub-system is to function, as well as design, installation and testing criteria.
After walking down the proposed routing of the system, the design engineer, using the above criteria, develops a diagrammatic field instal-lation drawing (FSK). The FSK is independently reviewed by another design engineer and approved by a lead design engineer. The drawing is then turned over to the Field Engineering group who is responsible for working with craft personnel to install the system according to the drawing.
Field Engineering uses a " red line" system for marking and identifying any necessary changes to the routing. Once installed, the " red lined" drawings are sent back to the design group, where any changes are reviewed by the design engineer and incorporated into an appropriate. revision of the draw-ing. This revised drawing is subsequently reviewed by an independent engineer, the lead design engineer, and the lead field engineer.
For subsequent revisions to the drawing, the design engineers review only the changes made to the drawing and not the whole drawing itself.
Therefore, an error made in the original design drawing will not necessarily be identified during subsequent drawing revision reviews. Also, the Bechtel QA/QC department reviewed the drawings to verify that all administrative criteria had been met and not, technical accuracy of the design.
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Revision 0 of the FSK for PDT 5029 and 5031 were both completed on April 11, 1983. However, SFAO did not approve and issue the vendor prints for these instruments until July 1983. The vendor prints clearly show that the high pressure port of the PDT is on the left side and the low pressure port on the right side of the installed sensor. Revision 0 of the FSK for PDT 5029 correctly showed the routing of the tubing to the PDT.
However, Revision 0 of the FSK for PDT 5031, incorrectly routed the tubing for the high and low pressure ports. The same designer and reviewer were responsible for both FSKs.
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The tubing installation and rework for PDT 5029 was completed in October 1984. At this time, the tubing lines were reversed per the field engineer's instructions. This reversed installation was then approved and i
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incorporated into Revision 1 of the FSK. A third design engineer reviewed and approved Revision 1, but the same engineer who initially checked the earlier revision also checked this revision.
.The tubing installation for PDT 5031 was not completed until April 1985.
This installation was made per the drawing, which was subsequently determined to be incorrect.
Revision 1 and all subsequent revisions of this drawing show the tubing being routed to the wrong ports of the PDT.
Each of the design engineers involved with this event had significant experience in design work at nuclear power plants; and, it appears that, after reviewing the Bechtel design engineering and installation process, that sufficient engineering design controls were established.
In addition, no other similar disempancies were identified.
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Review of Preoperational Test Results Data The inspector reviewed portions of preoperational tests PSSUG-PTP-GS-1, Containment Atmosphere Control System, and PSSUG-PTP-GP-1, Primary
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Containment Integrated Leak Rate Test, to determine if preoperational testing contained precursors of the vacuum breaker problem or was inappro-priately performed. Specifically, the inspector reviewed the methodology of performing the preoperational testing on 1-GS-PDT-5029 and 1-GS-PDT-5031 and reviewed the containment integrated leak rate test (CILRT) pre-operational test to determine if the 24 inch butterfly valve opened during the test. The inspector concluded that the methodology utilized to per-form the functional testing of the pressure differential transmitter through operation and cycling of the 24 inch butterfly valve was not capable of identifying that the sensing lines to the PDTs were installed in the reverse manner.
In addition, during the CILRT, the PDTs were isolated per procedure so that no data or conclusion could be drawn from this test relative to the sensing line configuration.
The methodology used to test the PDTs in the preoperational test program
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required connecting a variable pressure test source to the PDT and increasing the pressure until the valve opened.
Interviews with various plant personnel indicated that test personnel are instructed to always utilizc the high side of the PDT to impose pressure for testing. This would require isolating the reactor building and torus sensing lines to the PDT, and installing a pressure source to the high side of the PDT.
The required pressure differential is then applied via a high side test connection.
Since the butterfly valve control logic assumes the high side
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of the transmitter is connected to the reactor building air space, as the test pressure is increased the butterfly valves should open.
(NOTE: a positive pressure to the high side (reactor building) is equivalent to a vacuum on the low side (torus)). As installed, the tubing on the high side of the PDT was incorrectly connected to the torus.
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The preoperational test (PSSUG-PTP-GS-1) was performed several times.
Several test exceptions were identified due to recalibration of the pressure differential transmitter.
No concerns were identified regarding the configuration of the sensing lines.
The inspector also reviewed several Test Pack Release Documents for PDT-5029 (TPR-GSC - 9, 12, 22, 156, 186, 323, 441 and 444) and for PDT-5031 (TPR-GSC - 30,72,104,187,190,308,414,420). The method used to perform these calibrations was similar to the actual preopera-tional test. The sensing lines were isolated and a regulated pressure supply connected to the pressure differential transmitter. The inspector determined that none of the testing identified the problem.
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Review of PDT Setpoint Calculations The inspector reviewed the following calculations to determine if the calculations used the "as-built" configuration of the sensing lines:
Calculation 113 Reactor Building to Suppression Chamber Differential Pressure High, dated January 10, 1985 Calculaticn GS-18 Process Setpoints for Suppression Pool to Reactor Building Breakers, dated August 5, 1986 Calculation SC-GS-Setpoint Calculation for Reactor Building Atmosphere 0101 Control, dated June 7, 1986 Setpoint register J040Z, Revision 1 dated January 15, 1986, was also reviewed. No indication existed that the calculations of setpoints assumed anything other than calculating pressure differential with high pressure on the reactor building side relative to the torus.
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Surveillance Procedure Review The following surveillance procedures were reviewed to determine if they identified the sensing line installation error:
IC-SC.GS-008 Containment Atmosphere Control - Channel A
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PDT-5029 (Revision 0 dated September 4, 1986, and Revision I dated May 27,1986)
IC-CC.GS-006 Containment Atmosphere Control Division 1 Channel
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PD-5029 (Revision 0 dated October 4, 1985, Revision 1 dated June 2, 1986 and Revision 2 dated June 11, 1986)
IC-CC.GS-005 Containment Atmosphere Control Channel B P-5031
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(Revision 0 dated October 4, 1985, Revision 1 dated June 2, 1986 and Revision 2 dated June 2, 1986)
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The methodology to perform the surveillances is_similar to that of the preoperational test, in that the sensing lines are isolated for the PDT calibration and a controlled pressure device is used to calibrate the PDT. Similarly for the logic portion of.the circuit, the PDT is electrically isolated and a digital _ calibrator is utilized. The inspector concluded.that the. surveillance procedures would not determine that the sensing lines to the PDT were reversed.
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Licensee's Corrective Actions As a result of the discrepancies identified during the review of this event, the licensee initiated a number of corrective actions. These actions can be grouped into three categories, listed by the time frame of their completion.
1.
lImmediate (Prior to Unit Startup)
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A Design Change Package was completed, correcting the PDT tubing
_ error.for PDT 5029 and 5031.
A review of Q-Listed vacuum. applications for differential
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pressure transmitters was conducted to search for other tubing errors or misapplications. bk) problems were identified.
Instrument and Control personnel conducted a complete
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instrument valve lineup verification for all instruments in the reactor building. One discrepancy was identified, in that a MSIV sealing system pressure transmitter used for an alarm-annunciator was found isolated. This was corrected.
All Q, F and R Temporary Modifications were reviewed to verify
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that safety evaluations and unreviewed safety question determinations had been made.
This action was taken to provide added assurance that no changes had been made to the plant design that could possibly impact system operability. The licensee identified no additional problems.
SORC Meeting 86-196, held on August 11, 1986, approved a
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program for identification and position verification of
Q-Listed Instrument Valves.
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Short Term Recommendations i
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Verify that all non Q-Listed vacuum pressure differential pressure applications have correct instrument tubing routing
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installation to perform their intended design function.
l Develop a schedule for the expeditious conversion of existing
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temporary modifications to Design Change Packages.
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Revise Administrative Procedure AP-13 to require performance of
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a safety evaluation for all temporary modifications.
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Long Term Recommendations
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Terminate the practice of using temporary modifications in lieu
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of DCPs for permanent plant modifications.
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Establish a policy of completing the incident report, root cause assessment / draft LER within 20 days of the incident.
Develop, implement, and maintain a formal program to identify
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and document the design basis for Hope Creek.
On August 22, 1986, the Region I staff met with the licensee at the NRC Region I office in King of Prussia, Pennsylvania to discuss this event and the licensee's corrective actions.
Enclosure 1 to this report
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identifies the attendees and the information provided by PSE&G.
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Exit Interview The inspectors met with licensee and contractor personnel periodically and at the end of the inspection report to summarize the scope and findings of their inspection activities. Written material was not provided to the applicant during the exit.
Based on Region I review and discussions with the licensee, it was determined that this report does not contain information subject to 10 CFR 2 restrictions.
The inspector summarized that the as-built configuration of the reactor building to suppression chamber pressure relief system resulted in the system not being operable and was an apparent violation of Technical
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Specification 3.6.4.2.
Enforcement pertaining to this issue would be addressed separately from the inspection report.
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Enclosure 1 August 22, 1986 Meeting Between PSE&G and NRC Region I List of Attendees Name Title Organization T. Murley Regional Administrator NRC Region I
W. Kane Director, Division of Reactor Projects NRC Region I R. Summers Project Engineer NRC Region I S. Collins Deputy Director, Division of Reactor NRC Region I Projects D. Allsopp Resident Inspector, Hope Creek NRC Region I R. Gallo Chief, Reactor Projects Branch 2 NRC Region I T. Kenny Senior Resident Inspector, Salem NRC Region I S. Ebneter Director, Division of Reactor Safety NRC Region I J. Durr Chief, Engineering Branch NRC Region I D. Florek Lead Reactor Engineer NRC Region I P. Eapen Chief, Quality Assurance Section NRC Region I
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C. A. McNeill Vice President - Nuclear PSE&G R. Salvensen General Manager - Hope Creek Operations PSE&G B. A. Preston Manager - Licensing & Regulation PSE&G G. Peet Lead I&C System Engineer PSE&G/ System W. Braver Principal Safety Review Engineer PSE&G R. Burricelli General Manager Engineering and PSE&G Plant Betterment J. Mackinnon General Manager Nuclear Safety Review PSE&G D. Sullivan Resident Project Engineer Bechtel
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TECHNICAL SPECIFICATION 3.
6.
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"DRYWELL AND SUPPRESSION CHAMBER INTERAL PRESSURE SHALL BE MAINTAINED BETWEEN-0.
AND
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"THE LI MI T A TI O NS ON DRYWELL AND SUPPRESSION a
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CHAMBER INTERNAL PRESSURE ENSURE THAT THE
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EXTERNAL PRESSURE DIFFERENTIAL DOES NOT
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EXCEED THE DESIGN MAXIMUM EXTERNAL PRESSURE
DIFFERENTIAL OF
PSID.
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EVENTS o
7/4/86
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TORUS PRESSURE-0.
PSIG o
7/31 /86
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TORUS PRESSURE-0.
PSIG o
8/8/86
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FORCED SHUTDOWN
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EXCEEDED LCO 3.
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BOTH REACTOR BUILDING
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SUPPRESSION CHAMBER BUTTERFLY ISOLATION VALVES DECLARED INOPERABLE
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INVESTIGATION o
8/8/86
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TASK FORCE COMMISSIONED BY VICE
PRESIDENT TO INVESTIGATE FORCED SHUTDOWN
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TASK FORCE REPORT CONCLUDES ROOT CAUSE TO BE DESIGN DEFICIENCY j
C ISOMETRIC DRAWINGS SHOW TUBING
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TO PDT REVERSED)
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TORUS TO REACTOR BUILDING
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VACUUM BREAKER SYSTEMS / LINES ge, W l0F 8 ORYWELL ik TO TORUS VACUUM
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BREAKERS
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H 5031 5032 4-20 mADC 1-SV DC pg7 g
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TORUS COMPARTMENT
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X = TRIP = + 4.99'(+.18 PSID)
9 = RESET = + 4.71'(+.17 PSID)
RESPONSE CURVE FOR INSlHUMENT LOOP
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VACUUM BREAKER SYSTEMS / LINES y
10F 8 DRYWELL TO TORUS VACUUM
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ESB TORUS COMPARTMENT CORRECTIVE ACTION l
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4-55'
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X = TRIP = - 4.99'(- J8 PSID)
e = RESE1 = - 4.71'(.17 PS101 RESPONSE CURVE FOR INSTRUMENT LOOP WITH CORRECTIVE ACTION
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OBSERVATIONS MADE AS A
RESULT OF TASK FORCE INVESTIGATION o
SYSTEM TESTING PROGRAM WOULD NOT HAVE UNCOVERED REVERSAL OF THE SENSING LINES.
TESTING WAS r
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FULL INSTRUMENTATION CHANNEL TESTING VERSUS FULL SYSTEM OPERATIONAL TEST S,
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Setpoint Register J0402, Revision 1 IC-SC-GS-008, Revision 0 and Revision 1 Containment Atmosphere Control - PDT 5029 IC-CC-GS-006, Revision 0, Revision 1, Revision 2 Containment Atmosphere Control (C.A.C.) Division 1 PD 5029 IC-CC-GS-005, Revision 0, Revision 1, Revision 2 C.A.C. PDT-5031 10855-D3.40, Revision 4 - Design, Installation and Test Specification for Containment Atmosphere Control System for the Hope Creek Generating Station QC File No. 3G6-M57-1-1, QCIR No. - FSK-JD-1303-1-001-1-2-II.10 QC File No. 3G6-M57-1-1, QCIR No. - FSK-JD-1803-A-010-1-1-2 QC File No. 3G6-M57-1-1, QCIR No. - FSK-JD-1303-1-002-1-IC-Il QC Instruction No. 10855/I-1.10, Revision 3, Installation of Instruments
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Specification 10855-J-825(Q), Revision 8, Technical Specification for Instrument Installation SWP/P-057, Revision 1, Specific Work Plan / Procedure for Installation
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of Instrumentation SWP/P-010, Revision 19, Specific Work Plan / Procedure for Field Design Approval and Control
SWP/P-J-101, Revision 9, Specific Work Plan / Procedure for Instrument Field Design, Materials Installation, Surveillance / Inspection I
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