IR 05000344/1987034
| ML20236X814 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 11/25/1987 |
| From: | Ramsey C, Richards S NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML20236X774 | List: |
| References | |
| 50-344-87-34, GL-86-10, IEIN-84-16, IEIN-86-017, IEIN-86-17, NUDOCS 8712100290 | |
| Download: ML20236X814 (19) | |
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U. S. NUCLEAR REGULATORY COMMISSION
REGION V
Report No. 50-344/87-34 Docket No. 50-344 License No. NPF-1 Licensee:
Portland General Electric 121 S. W. Salmon Street Portland, Oregon 97204 Facility Name:
Trojan Nuclear Plant Inspection at:
Rainier, Oregon
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Inspection Conducted:
September 21-25, 1987 Inspected by:
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///s s//g,y C. Ramsey, Reactor Irfspector Date Signed
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jfff[g7 Approved by:
e S. Richards,' Chief, Engineering Section Da'te 6igned Summary:
Inspection on September 21-25, 1987 (Report No. 50-344/87-34)
Areas Inspected:
Routine unannounced inspection of the licensee's fire protection program implementation and action on previous NRC inspection I
findings.
Inspection procedures 64704 and 92701 were used.
l Results:
Of the areas inspected, three violations were identified, geum 6%
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DETAILS 1.
Persons Contacted.
Portland General Electric
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t D. W. Cockfield, Vice President Nuclear
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- D.'R. Swanson, Manager, Nuclear Safety Branch
- B. Schmitt, Manager, Operations and Maintenance
- J. Reid, Manager, Plant Services
- C. Brown, QA Operations Branch Manager
- P. Morten, Manager, Plant Engineering Branch
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- D. Swan, Manager, Technical Services
- D. Nordstrom, Compliance Engineer
- D. Modeen, Nuclear Engineer
- H. Caballero, Safety Coordinator
- A. Sanchez, Fire Protection Engineer
- R. Russell, Operations Supervisor
- R. Reinhardt, Safety Coordinator
- E. Davis, Electrical Branch
- D. Farrell, NQAD Operations Branch
- G. Bair, Plant Engineering NRC
- G. Suh, Resident Inspector
- Denotes personnel present at the exit interview.
2.
Licensee Action on Previous Inspection Findings A.
(Closed) Information Notice No. 86-17, " Potential Failure of Automatic Sprinkler Corporation Model C Deluge Valves."
This Information Notice provided licensees with notification of a problem involving the operational failure of deluge and pre action fire protection water control valves, identified as "Model C,"
manufactured by Automatic Sprinkler Corporation.
This notice contained supplemental information to information provided in Information Notice No. 84-16 issued on March 2, 1984.
In response to Information Notice No. 84-16 regarding the subject
"Model C" fire protection water control valves, the licensee determined that successful test results for 14 - 2 inch and 2 - 6 inch Model C type deluge valves at Trojan demonstrated that no further action to this notice was required.
In response to Information Notice No. 86-17, during the inspection, the licensee provided the inspector with a revised testing procedure for these valves that incorporated the valve manufacturers enhanced i
instructions for periodic testing and maintenance.
According to the valve manufacturer, these enhanced instructions contain the solution l
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for the problems identified in the Information Notices that were associated with these valves.
Therefore, this item is closed based on the licensee's corrective actions.
B.
(0 pen) Items 1 through 18 of Inspection Report No. 50-344/83-18.
The items identified in this inspection report relate to the licensee's failure to meet the technical requirements of Appendix R to 10 CFR 50.
For these conditions, the licensee implemented interim compensatory measures and has been involved in modifying the plant to meet these requirements on a schedule that has been agreed upon by NRR.
This is further discussed in Paragraph 48 of this report.
3.
Routine Fire Protection Program Implementation General Design Criterion 3 of Appendix A to 10 CCR 50 requires that each nuclear power plant be provided with fire detection and fire fighting systems of appropriate capacity and capability to minimize the adverse effects.of fires on structures, systems and components important to safety.
The NRC guidance on the implementation of General Design Criterion 3 of Appendix A to 10 CFR 50 for the Trojan Nuclear Plant is contained in Appendix A to NRC Branch Technical Position 9.5-1 as stated in the NRC's Safety Evaluation Report (SER) for the facility dated March 9,1978.
Based on the licensee's response to Appendix A to Branch Technical Position 9.5-1 and alternatives proposed by the licensee, the NRC concluded in Section 2.1 of the SER that the overall objectives of General Design Criterion 3 were met except as described in Amendment P!o.
22 to Facility Operating License No. NPF-1.
The licensee document entitled " Trojan Nuclear Plant Fire Protection Review" and Section 9.5-1 of the FSAR contains the licensee's response to Appendix A to Branch Technical Position 9.5-1 and alternatives proposed by the licensee.
Plant Technical Specifications set forth limiting conditions for operation and surveillance requirements for various aspects of the program.
The results of the inspectors review of the licensee's fire protection program implementation, and the licensee's compliance with the above requirements, are as follows:
A.
Qualified Staff Involvement Technical Specification No. 6.8.2 requires that written procedures
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be established, implemented and maintained covering the activities of the fire protection program.
Technical Specification No. 6.5.3 requires that each procedure (or changes thereto) implementing the j
fire protection program and proposed changes or modifications to plant nuclear safety-related structures, systems and components be reviewed by a qualified individual of the appropriate discipline as designated by approved written procedures.
Individuals responsible for such reviews are required to be members of the technical review groups designated by approved procedures and each review is required to be documented, whether or not an unreviewed safety question is
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involved.
Additionally, the licensee's procedures and the NRC i
Branch Technical Position 9.5-1 guidelines require that personnel f
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charged with implementing the fire protection program be qualified and designated by procedure.
The licensee's response to NRC position No. 1.0 of Appendix A.to
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Branch Technical Position 9.5-1, as stated in the " Trojan Nuclear Plant Fire Protection Review," states in part, " existing procedures are considered to be in conformance with the guidelines."
The inspector reviewed the licensee's. staffing in the fire protection area and determined that, in addition to a corporate loss prevention engineer, the licensee employs an onsite qualified fire protection engineer who is involved in implementation of limited aspects of the program.
Periodically, the licensee employs outside
consultants to assist in implementing the program.
Also, there are two safety coordinators assigned responsibility for the program's implementation at both the site and corporate levels.
Despite this compliment of qualified fire protection personnel in the licensee's organization, overall responsibility for implementation of the program requirements is fragmented and dispersed throughout the organization to individuals of varying expertise.
None of these
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individuals, including the fire protection engineers and safety
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coordinators, appeared to be in a position within the organization to provide appropriate levels of management with the necessary technical, operational and administrative input required to formulate, implement and assess the effectiveness of the program.
Additionally, the program description in Topical Report PGE-1012 does not accurately describe the role of the fire protection engineer's involvement in implementing the program.
There appeared to be no one in a management position overseeing implementation of the program and providing a single point of control and contact for all contingencies.
This lack of program direction and coordination
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is viewed by the inspector as a significant programmatic weakness which may have contributed to the following examples of the licensee's failure to adhere to Technical Specification requirements:
(1) Failure of Procedure No. NDP 200-1 to Require Fire Protection Review by Qualified Individual of the Appropriate Discipline The licensee's procedure No. NDP 200-1 requires a fire protection review of Design Change Packages (DCPs) to assure
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that fire protection requirements are adhered to.
However, the procedure does not require that a qualified fire protection engineer (the appropriate discipline) perform this review prior to or subsequent to the Plant Review Board's (PRB) approval of procedures, changes or modifications to plant nuclear j
safety-related structures, systems or components.
According to the licensee, a qualified fire protection engineer has reviewed DCPs prior to PR8 approvals in some cases.
However, the licensee acknowledged that there was no consistent policy in place to require this review of DCPs by the appropriate discipline engineer as required by the plant technical specification.
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Failure to require that a qualified fire protection engineer review des'gn changes is considered a violation of Technical Specification 6.5.3.1 (344/87-34-01).
(2) Failure to Follow Procedures Requiring Qualified Personnel to Issue Welding and Cutting Permits
Technical Specification No. 6.5.3.1.d requires that individuals responsible for reviews of fire protection program implementing procedures be qualified in the appropriate discipline.
In response to NRC Guideline Position No. 1.2.1(c)(1) of Appendix A to Branch Technical Position 9.5-1 as stated in the " Trojan Fire Protection Review," the licensee's statement indicates that welding and cutting permits must be completed and approved by an appropriate cognizant supervisor.
The NRC guideline statement reads in part, " Procedures governing such work should be reviewed and approved by persons trained and experienced in fire protection."
i Item No. 5 of the licensee's Triennial Audit No. JLD-7-87 of the Fire Protection Program dated January 12, 1987, identified the lack of training and experience in fire protection for supervisors issuing welcing anc. cutting permits.
The licensee's Nuclear Quality As urance Department (NQAD) issued Nonconforming Activity Report (NCAR) No. P86-123 and recommended that designated personnel initiating welding and cutting permits be specifically trained in fire protection and items to be considered in completing the permits.
The NCAR recommendation pointed out that this item was previously identified in the Triennial Audit Report of December 1983.
During the inspection, the licensee was unable to provide
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documentation or other evidence of training and experience in
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fire protection for all supervisors issuing welding and cutting permits.
Furthermore, the licensee provided no documentation to demonstrate that a root cause evaluation was performed or that comprehensive corrective action was taken in response to NCAR No. P86-123.
According to the NCAR, the licensee's corrective action was limited to counseling the supervisor and worker involved in the. incident discussed in the NCAR.
Failure to take corrective action to preclude repetition is considered a violation of 10 CFR 50, Appendix B, Criterion XVI (344/87-34-02).
8.
Technical Specification Surveillance Testing of Fire Protection Systems Various aspects of the Trojan Nuclear Plant fire protection system are governed by the limiting conditions for operation (LCO) and surveillance requirements of plant technical specifications.
Procedures for tests which affect plant nuclear safety are required
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to be prepared, reviewed and approved prior to implementation by qualified personnel of the appropriate discipline.
The results of
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the inspectors review of the licensee's fire protection technical specification surveillance testing program are as follows:
(1) Fire Suppression Water Supply System Operability Flow Test Every 3 Years Technical Specification No. 4.7.8.1.1 requires that the fire suppression water supply system be demonstrated operable at least once every 3 years by performing a flow test of the system in accordance with Section 11, Chapter 5 of the NFPA handbook, 14th edition.
Regarding the fire suppression water supply system installation for nuclear power plants, the licensee's response to NRC guideline position statement No.
5.2.1.1 of Appendix A to BTP 9.5-1 (as stated in the " Trojan Nuclear Plant Fire' Protection Review") reads in part, "A 10 inch underground yard fire m.in loop is installed in accordance with NFPA 24."
Section 11, Chapter 5, of the NFPA hendbook specifies that a flow test of fire protection water supply systems be made to determine the available water flow and to evaluate the performance of the system by comparison with recognized standards.
The hydraulic principles underlying the water flow and pump testing procedures are specified and are to be used to measure appropriate test parameters to ensure proper operation.
The provisions of the NFPA handbook that were not appropriately addressed in the licensee's test procedure No. PET 11-1 for the years 1981, 1983 and 1986 were as follows:
Limitations are placed on the range of test gauges to obtain the best measurement of water flow.
The range of test gauges used in the licensee's test to take velocity i
pressure readings (water flow) exceeded the maximum range specified in Section 11, Chapter 5 of the handbook.
Pressure gauge ranges were allowed up to 300 psig vice 200 psig as specified in the handbook.
Utility connections (i.e., sprinkler and standpipe systems) are required to be isolated for the test.
The licensee's procedure did not require these connections to be isolated.
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The method of measuring velocity pressure readings is required to be controlled such that readings are to be kept within a specified pressure limit (between 10 and 30 PSI for best accuracy).
Additionally, the handbook places limitations on the insertion of test devices into the flow stream.
The licensee's test procedure did not specify limitations on the method used to take these readings.
Motor or engine controllers are required to be equipped
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The licensee operated both pumps in parallel and not sequentially for the tests.
l Section 11, Chapter 5 of the handbook, states that the hydraulic principles underlying water. flow and pump testing procedures should be in accordance with Section 11, Chapters 1, 2, 4 and 6 of the handbook.
The licensee's test procedure did not consider the appropriate
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l hydraulic principles referenced in Section 11, Chapter 5 of the handbook.
One result of the test is an
" extrapolated flow at 20 psig" value.
The acceptance criteria stated in the licensee's procedure was 5800 -
6200 gpm.
The test results for 1981, 1983, and 1986 were 6390, 6695, and 7200 gpm, respectively.
The licensee accepted the latest test result, concluding that the higher than expected value was the result of the fire pumps performing better than required.
Apparently no effort was made to determine whether testing inaccuracies could have caused the wide variance in test results for the-last three tests.
This is considered an unresolved item (344/87-34-03).
(2) Fire Pump Operability Test Every 18 Months The licensee's response to NRC guideline position statement No.
1.3.1 of Appendix A to BTP 9.5-1 reads in part, "a test program will be established and implemented to assure that testing is performed to demonstrate conformance with design and system readiness requirements." The licensee's procedure No. PT-10 does not contain instructions for testing the station fire pumps in accordance with the design criteria of NFPA 20.
Regarding the fire pump installation, the licensee's response to NRC guideline position statement No. 5.2.3.2 of Appendix A-to BTP 9.5-1 reads in part, "the fire pump installation was designed in accordance with NFPA 20."
NFPA 20 parameters of fire pump assemblies required to be tested annually include the following:
Measurement of 3 points on the pump performance curve (shutoff churn test with no flow; approximate rated capacity and 150 percent of rated capacity or peak load).
Measurement of suction and net discharge pressures at minimum, rated and peak' loads.
Measurercent of electric current for electric motor driven pumps at minimum, rated and peak loads.
Measurement of pump speed at minimum, rated and peak load.
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Measurement of pump vibration at minimum, rated and peak loads.
Comparison of test results to pump characteristic curves to determine performance is maintained between 5 and 10 percent of original design.
Technical Specification No. 4.7.8.1.1 only requires that the station fire pumps be demonstrated operable every 18 months by demonstrating their ability to achieve 2,000 GPM at 125 PSI i
(rated capacity).
An example of why the licensee would be
prudent to perform more extensive testing is that for the fire I
pumps to provide a backup water supply to the Service Water System, in accordance with the licensee's emergency procedure for a loss of the Service Water System, the licensee's calculation No. DJP-83-59, which supports the adequacy of the emergency procedure, is based on each pump operating above rated capacity and delivering 2500 GPM at approximately 95 PSI.
By not testing the pumps in accordance with NFPA 20 test criteria, the licensee 'is not periodically demonstrating the pump's ability to achieve this flow.
Licensee's Position During the inspection, the licensee's representative stated that the purpose c,f performing test procedure No. PT-10 was to demonstrate that Technical Specification requirements (2000 GPM at 125 PSI or above) were met and that no other testing was required.
Although the inspector concluded that the licensee was in compliance with the minimum surveillance requirements,
the inspector observed that this philosophy in system testing i
is contrary to the licensee's stated intention of performing I
surveillance at more than the minimum level.
This is considered an open item (344/87-34-04).
(3) Fire Detection System Operability Test Every Six Months Technical Specification No. 4.3.3.7.1 requires that fire
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detection instruments protecting safe shutdown areas of the plant be demonstrated operable by performance of a channel i
functional test.
Regarding the installation of fire detection systems for nuclear power plants, the licensee's response to NRC guideline position statement No. 4.7 of Appendix A to BTP 9.5-1 (as stated in the Trojan Nuclear Plant Fire Protection Review) reads in part, "The Trojan plant fire detection system complies with NFPA 72D."
Further, in response to NRC guideline position statement No. 2.0, the licensee's statement reads in part, " Fifty ionization-type detectors with adjustable
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sensitivity are provided for monitoring."
Article 350, Section 3532, of NFPA 720, requires that suitable and practical facilities be provided to permit periodic testing for sensitivity of fire detection instruments.
Section 3520 of I
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this article refers to NFPA 72E regarding these installations.
l Chapter 7, Section 7-4.1, of NFPA 72E requires that
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manufacturer's instructions be consulted prior to the cleaning, checking of operation and sensitivity adjustment for each fire detection instrument.
This section further states that
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" Ionization and photoelectric smoke detectors may require l
periodic cleaning to remove dust or dirt which has accumulated.
The frequency of cleaning will depend on the local ambient conditions." The licensee procedures have no provisions for l
cleaning of detectors.
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i The licensee's JJne 30, 1987, submittal of Amendment No. 3 to the Trojan Nuclear Plant FSAR identified the licensee's failure to perform sensitivity tests for all fire detectors based on an interim justification.
During the inspection, the inspector reviewed the licensee's interim justification and questioned the appropriateness as follows:
a.
Sensitivity Testing at Initial Installation Was Not Performed Initial sensitivity testing of the original fire detection syr, tem was not performed, and there is no indication that testing has been performed since initial installation.
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Sampling Method
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The licensee estimated an 11 percentpf'ailure rate based on 33 detectors actually tested witt no failurer of a total
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population of 210 detectors.
The sampiing method did not consider the different environments that detectors are installed in and the need for cleaning and sensitivity adjustment of individual detectors based on environmental i
conditions.
c.
Detector Sensitivity Drift Outside of Design Limit The l'icensee's fire detection system includes several types of fire detection devices but primarily ionization type fire detectors are used.
The referenced NFPA standards recognize that the sensitivity setting of fire detectors is subject to drift.
In this regard, the licensee's justification did not consider sensitivity drifts that would cause a detector to fail to alarm outside of its design limit.
Instead the licensee concluded that it is recognized that ionization detectors l
become more sensitive with time due to the principle of I
their design.
According to the licensee, this would only result in increased nuisance or false alarms which could i
be tolerated by use of a fire watch until the
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malfunctioning detector could be repaired or replaced by maintenance personnel.
The licensee apparently did not consider the potential for non-conservative setpoint drift
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to occur such that any increase in sensitivity, due to w _
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particle accumulation in the detector, could be more than offset, d.
Licensee Proposed Sensitivity Tests Although no sensitivity testing was performed following the original detector installations and, the licensee could not verify that the sensitivity of detectors have or have not been field-adjusted since original installation, the licensee concluded that due to the cost associated with performing sensitivity testing, only new or replacement detectors will now receive sensitivity tecting in order to establish that their sensitivity is within design limits upon installation.
Also, following maintenance or troubleshooting, the licensee plans to perform sensitivity testing of individual detectors on a case by case basis.
Thus, eventually, all detectors can be tested for sensitivity during the life of the plant.
Using this rationale, the licensee concluded that this will remove the possibility that any detectors could have been insensitive upon initial installation, e.
Purpose of Sensitivity Testing The licensee's semi-annual functional testing of fire detectors in accordance with procedure No. POT 10-5 and the Technical Specifications, only verifies the operability of the electrical circuit remote from the detector instruments (i.e., broken leads or faileo circuit components that should be recognized by the supervision capability of the circuit and cause a trouble alarm to annunciate in the control room.)
Fire detector sensitivity testing is a method of electrically \\erifying that the alarm setpoints of detector instruments remain within design limitations.
One way that this can be accomplished is by applying a voltage signal to the detector to simulate a rising concentration of smoke causing the detector to alarm.
The original detector alarm setpoints and the detector response time establish the acceptance criteria for operability of the detector.
By not performing sensitivity testing, the licensee is not ensuring that a detector will alarm within its design limit.
This is considered an open item (344/87-34-05).
C.
Seismic Qualification of Interior Standpipe Hose Stations j
In response to NRC guideline position statement No. 5.3.5.1 of j
Appendix A to BTP 9.5-1 (as stated in the " Trojan Nuclear Plant Fire Protection Review"), the licensee takes exception to the position that " provisions should be made to supply water at least to standpipes and hose connections for manual fire fighting in areas i
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within hose reach of equipment required for safe shutdown in the evert of a safe shutdown earthquake (SSE).
The standpipe system serving such hose statior.s should be analyzed for SSE loading and be provided with supports to assure system pressure integrity." The licensee's response to this position reads in part, "This capability is not required at Trojan in accordance with the new guidelines."
It is not clear which "new guidelines" the licensee's statement makes reference to.
However, the latest NRC fire protection guidelines are contained in Generic Letter No. 86-10.
The NRC's response to industry question No. 7.2 of Generic Letter 86-10 restates NRC guideline position statement No. 5.3.5.1 of Appendix A to BTP 9.5-1 (as stated in the " Trojan Nuclear Plant Fire Protection Review") and further states that post-seismic procedures should include damage surveys and a determination of whether any fires were initiated as a result of the seismic event.
In addition, the generic letter notes that these guidelines are not applicable to plants reviewed and approved under BTP 9.5-1.
The Trojan Nuclear Plant was reviewed and approved by the NRC under Appendix A to BTP 9.5-1.
Therefore, it appears that the guidelines for seismically qualified interior standpipe hose stations are applicable to the facility design.
The licensee took exception to this position, however, written documentation from the NRC in the form of an approved deviation or exemption request had not been received.
Furthermore, the licensee's post-seismic event procedure No. ONI-52 did not include instructions to operators regarding damage surveys and a determination of whether any fires were initiated as a result of the seismic event.
To resolve these concerns, the licensee indicated that efforts would be made by appropriate levels of licensee management to obtain
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written documentation from the NRC documenting the facility relief from this NRC guideline position.
In addition, the licensee indicated that post-seismic event procedure No. ONI-52 would be revised to include instructions to operators for performing fire damage surveys following any seismic event.
This is considered an open item (344/87-34-06) pending further
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licensee and NRC action.
D.
Manual Fire Fighting Capability In resporise to NRC guideline position statement No. 1.2.1 of Appendix A to BTP 9.5-1 (as stated in the " Trojan Nuclear Plant Fire Protection Review"), the licensee describes manual fire fighting and fire prevention training as being in conformance with the intent of NFPA 27, 194, 197, 601 and other NFPA standards on these subjects that are generally recognized as industry accepted practice.
Speci fication Nos. 6.2.2(F), 6.4.2 and 6.8.2 of Technical Specification No. 6.0 require that a 5-man fire brigade be i
maintained onsite at all times; the training program for the fire brigade to meet or exceed the requirements of NFPA 27-1975 and
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procedures be established to implement manual fire fighting and fire prevention capability.
The results of the inspectors review of these aspects of the l
licensee's fire protection program implementation are as follows:
(1) Fire Brigade Training NFPA standard Nos. 27 and 197 require that fire brigade
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training programs be adapted to the purpose of the particular brigade.
The training program is required to be upgraded in order to keep pace with new hazards presented to them so that their ability to perform the operations they are expected to
carry out is maintained.
The licensee's discussion of fire l
brigade training in the " Trojan Nuclear Plant Fire Protection
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Review" states that the licensee is in full compliance with the
NRC position that each member of the brigade must have knowledge and experience in fighting the type of fires expected to occur in nuclear plants (internal structural fires).
Paragraph 44 of NFPA 27 provides guidance on obtaining l
assistance on setting up training programs for fire brigades l
utilizing outside agencies such as municipal fire departments or state fire schools.
l The inspector's interviews with licensee fire brigade members disclosed the following:
The site fire brigade receives no training in internal structural fire fighting.
According to the licensee, such training was provided for the fire brigade at one time earlier during plant operation but was discontinued.
One fire brigade leader interviewed by the inspector indicated that he had gained actual internal structural fire fighting experience while in the U. S. Navy some years ago, but had not actually fought this type of fire since that experience.
Another fire brigade leader interviewed by the inspector indicated that neither he,
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nor any member of his brigade, had any experience or training in fire extinguishment inside of structures.
All of the brigade members interviewed by the inspector expressed appreciation for the classroom instruction and training provided during drills by the Safety Coordinator.
Hnwever, the brigade members indicated to the inspector that they recognized their need for additional and i
reinforced training in the type of internal structural
fire fighting strategies and techniques they would have to
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rely on when fighting an actual internal structural fire in the plant.
As part of quarterly drills / practice sessions, the licensee's past practice included the fire brigade's extinguishment of actual flammable liquid or ordinary
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g combustible fires in outside open areas.
This practice was discontinued by the licensee due to environmental l
concerns.
Therefore, the brigade no longer receives
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training in external structural fire f.ighting strategies I;
and techniques.
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The licensee's internal audit No. JLD 7-87 (finding No. 8)
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identified the lack of fire brigade internal structural
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firefighting training as a deficiency in the fire brigade i
training program.
Regarding this nonconforming activity,
corrective actions had not been taken on this matter.
(2) Offsite Fire Department Assistance The licensee's response to NRC guideline position statement No.
1.2.1 of Appendix A to BTP 9.5-1 (as stated in the " Trojan Nuclear Plant Fire Protection Review") and NFPA 27-1975, require that training of the plant fire brigade be coordinated with the local offsite fire department so that responsibilities and duties are delineated in' advance.
The local fire
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department is required to be educated in the operational l
precautions when fighting fires on nuclear plant sites, and are to be made aware of the need for radiological protection of personnel and of the special hazards associated with nuclear
power plant sites.
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The results of the inspectos assessment of the licensee's offsite fire department assistance training and readiness
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capability are based on the inspector's tour of'the offsite fire department facility,. observation of fire fighting equipment, review of pertinent written material and interviews with personnel.
Based on these activities, the inspector obwerved the following:
l The offsite fire department had received no tours of plant vital areas in order to familiarize themselves with the l<
plant layout and simulate the use of fire fighting
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equipment in preparation for an actual fire event in which
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a particular area is filled with heat, smoke, hot gases and contains insufficient lighting.
Although the offsite fire department has participated in annual fire fighting drills with the site fire brigade,
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the scope and death of these exterior structural fire
fighting evolutions have not adequately preparta the
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offsite fire department to support the site fire brigade fire fighting activities inside or outside of vital areas 01 the plant.
The offsite fire department did not have a I
high level of confidence in their role is backun to the j
site fire brigade.
In addition to limited practice j
e 'olutions during annual drills, the offsite fire
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dt.?artment personnel were uncertain about specific responsibilities that might be delegated to them during an actual fire at the site.
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The offsite fire department had not been educated in the operational precautions and specific radiological hazards associated with various areas of the plant.
Offsite fire department personnel indicated that they had received the radiological training that the licensee provides general employees through the General Employee Training (GET)
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program.
However, this training was not specific to the
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abnormal conditions that could be present during fire
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fighting activities in the various areas of the facility,
Despite an offsite fire department agreement that became effective on February 6, 1974, which specified that all offsite fire department personnel assisting in firefighting activities be provided with security clearances in order to gain emergency access to the site.
offsite fire department personnel have received no security clearances or site access badges.
Therefore, the offsite fire department has consistently been denied or delayed access when responding to the site for fire i
emergencies, fire alarms or followup on fire occurrences which were reported to have been extinguished prior to the offsite fire department's arrival.
According to the offsite fire department personnel interviewed, the continuing experience of denied access has had some negative impact on personnel motivations about responding to the site since the average number of responses to the site has been approximately 15 times per year over'the past four years.
Offsite fire department access to the site apparently has not been permitted on any of these occasions.
Scheduled fire drills with the fire brigede appears to be the exception to the overall offsite fire department access problem.
With some delay, the offsite fire department has been permitted onsite to participate in these drills.
The offsite fire department had one copy of outdated pre-fire plans for the site which was maintained by the fire chief in an office.
There were no copies of the plans on fire apparatus that wou d be accessible to fire companies responding to the plant site from locations remote from the main fire station.
In aduition, there appeared to be no reliable system in place for the
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licensee's timely updating of the plans or assurances that the offsite fire depertment will be kept abreast of
changes to the plans as they occur.
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Although the offsito fire department chief interviewed by
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offsite fire department and the site could work with improvements, the inspector determined that the offsits fire department assist & ace agreement with the site was l
inadequate in that it did not clearly specify the
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responsibilities and expectations of both parties.
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This (Items 3D(i) and (2)) is considered a violation of
Technical Specifi ution No. 6.4.2 (344/87-34-07).
(3) General Employee Trai ing Program (GET)
l As a result of the inspector's review of the licensee's GET program on fire protection at the site, tiie inspector made the following observations about the program content:
The program instructs general employees to make technical assessments about the type and size of fire occurrence before attempting to extinguish fires but the program does not provide the degree of training and experience to
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i general employees that would enable.them to accurately make such decisions consistently.
By directing general emplop as to attempt to extinguish fires first and notify the control room / fire brigade af ter extinguishment of the fire, the program places first-line responsibility for fire extinguishment on general employees.
By fellowing these directions, a fire brigade response can be delayed.
The program contains terminolo0y that may not be easily comprehended by the novice to fire protection.
The instructions given to general employees appear in some instances to be in conflict with the requirements of other regulatory agencies such as the Occupational Safety and Health Administration (OSHA).
In response to the n inspector observations, the licensee indicated that the fire protection portion of the GET program would be review and revised where necessary; particularly in the area where the potential exists for general employees to delay fire brigade responses to actual plant fires.
This is considered an open item (344/87-34-08) pending Region V j
followup on the licensee's corrective actions.
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E.
Schedular Fire Protection Modifications Required by Amendment No. '22
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to Facility Operating License NPF-1.
By letter dated March 9, 1978, the NRC issued Amendment No. 22 to i
Facility Operating License No. NPF-1, which required the licensee to complete certain fire barrier modifications as identified in Sections 3.1 of the Trojan Nuclear Plant Safety Evaluation Report (SER) dated October 1, 1978, prior to the licensee's return to operation for cycle 3.
According to the licensee, these modifications were completed in i
accordance with the Amendment 22 schedular requirements.
However, as a result of reassessing the plant configuration for Appendix R
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compliance, the licensee indicated that some of the fire barriers
required by Amendment No. 22 may no lo: ger be. required by techn cal i
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i specification due to the predesignation of fire area boundaries l:
permitted under Appendix R and NRC Generic Letter No. 86-10.
In each case where.a fire' barrier was required by Amendment No. 22, the licensee was uncertain whether a 50.59 review was appropriate for the deletion of the fire barrier, and the licensee was uncertain whether each deleted barrier was identified to the NRC in Appendix R or updated FSAR submittals.
In response to the inspector's concern, the licensee indicated that a review of the fire area boundary redesignations would be performed in order to assure that fire barriers required by Amendment No. 22-that are no longer being maintained have been properly i
dispositioned.
This is considered an open item (344/87-34-09) pending further licensee action and Region V followup.
4.
Safe Shutdown Capability In June 1983, the NRC conducted an inspection of the Trojan Nuclear Facility in order to determine the extent of the licensee's compliance with Appendix R to 10 CFR 50.
Eighteen significant inspection findings were identified and documented in Region V Inspection Report No.
50-344/83-18.
All of these in';pection findings remain open pending the licensee's final implementation of plant modifications and procedures to achieve Appendix R compliance.
According to the licensee, the final implementation schedule of June 1988, as stated in the licensce's May 20, 1987 submittal to the NRC, will be met.
l Following the June 1983 Appendix R inspection by the NRC, the licensee implemented a number of compensatory measures pending final implementation of the Appendix R modifications.
An important element of.
I these interim measures was plant operating procedures developed to l
achieve post-fire safe shutdown.
The results of the inspector's review of the licensee's interim post-fire safe shutdown procedures are as follows:
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A.
Operator Training in Interim Post-Fire Safe Shutdown Methodology Licensee procedure Nos. EFP-0, EFP-1, EFP-1.1 and EFP-2 implement tne interim post-fire safe shutdown methodology.
Procedure No.
EFP-0 establishes operator actions for a fire occurrence in most locations at the plant.
Procedure No. EFP-1 establishes operator actions for achieving control room alternative post-fire safe i
shutdown.
Procedure No. EFP-1.1 establishes operator actions for fire damage assessments upon a control room evacuation and procedure 10. EFP-2 establishes operator actions for alternative post-fire safe shutdown upon a complete loss of service water.
According to the inspector's review of available records and interviews with operations personnel, operators receive Reactor Operator licensing initial tuining, interim training, initial watch station training, annual req wlification training and daily shift
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crew training'on the procedures.
Operators were knowledgeable of various aspects of the procedures and displayed confidence in their ability to implement them.
No violations or deviations were identified.
B.
Use of Offsite Fire Department to Provide Gackup Water Supply to the Service Water System Potential loss of the service water system is described in Section 9.2.1 of the FSAR by the licensee.
The results of the inspector's review of the licensee's mitigating actions for this condition are as follows:
(1) pomplete Loss of Service Water Pumps Coincident with a L'oss of Offsite Power For the case where the service water system pumps could be lost coincident with a loss of offsite power due to fire, instructions provided to operators in the licensee's emergency procedure No. ONI-15 direct operators to implement the operator actions for alternative post-fire safe shutdown required by procedure No. EFP-2.
To provide assurance that this capability is available, the licensee has provided fire barrier protection for electrical circuits to the electric fire pump located in the Intake Structure.
The electric fire pump ci_rcuitry has also been modified to enable the pump motor to be driven by onsite power supplied from the emergency diesal generators; thereby, ensuring that both the electric and diesel driven fire pumps are available to provide a backup water supply to the service water system in this event.
To support implementation of this capability, the licensee provided the inspector with engineering calculation No.
DJP-83-59, which indicates that with both fire pumps operating in parallel, a balanced flow of 4,500 GPM would be available from the pumps.
This method provides for supplying the service water system via 5 outside pump discharge headers and the number 1 fire hydrant using 6 lengths of 2 inch fire hoses (50 feet maximum) to feed 6 - 2 inch emergency siamese connections to the service water system train "A" supply piping.
With each fire pump operating in accordance with the original manufacturer's pump curves, the pumps would deliver 2,500 GPM at approximately 95 PSI.
Based on these values, it appears that operating in parallel, the pumps could deliver 4,500 GPM i
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to the branch tee at the train
"A" service water system 30 inch supply line.
The licensee indicated that there was no supporting test data available to demonstrate the adequacy of this backup capability.
The inspector concluded, that due to the unusual configuration of the system in this situation, a test of the capability of the system to be properly aligned and to provide the required flow would be a prudent action.
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(2) Use of Cooling Tower Basin as a Backup to the Service Water System In the event of a loss of all service water pumps or a loss of the intake structure, the licensee's emergency procedures direct the use of the cooling tower basin as an alternate cooling water supply for the service water system.
Depending on the circumstances, the cooling tower basin may require a make-up water supply.
Step No. 3.5.19 and 4.5.11.1 of the licensee's emergency operating procedure No. ONI-15 instructs operators to call the local fire department and initiate fire department pumper truck drafting from the river.
.Upon the inspector's request, the licensee was unable to provide a procedure for accomplishing this fire department evolution.
According to the offsite fire department personnel interviewed, offsite fire department personnel have informed the licensee that this evolution was not possible with existing fire department pumper capability.
Therefore, the licensee and the offsite fire department have been engaged in discussions about this capability for the past two years.
The offsite fire department indicated that using the licensee designated draft site, fire department pumpers would not be capable of supplying water from the Columbia River in this situation due to design limitations for drafting.
The fire department pumpers are Underwriters Laboratory (U.L.) certified
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for drafting water a maximum lift of 10 feet.
From the licensee's designated draft site, the required drafting lift appears to exceed 25 feet.
In response to the inspector's concern over this capability, the licensee indicated that supporting engineering calculations or actual test data demonstrating the adequacy of this backup capability were unavailable.
This (items B.(1) and B.(2)) is considered an unresolved item pending further NRC verification of this backup capability.
(344/87-34-10).
5.
Quality Assurance The inspector reviewed the licensee's Fire Protection Program Quality Assurance Audit Nos. JLD-7-87, CAO-27-86 and CAO-265-86.
Each audit contained a broad scope and in-depth review of the areas audited.
The audit findings were clearly documented with appropriate recommendations for program enhancement or associated Non-Conforming Activity Reports (NCARS) which identified areas of the licensee's failure to comply with the program requirements.
Audit Report No. CAO-265-86 identified additional Appendix R concerns in the area of fire protection features, safe shutdown systems, safe shutdown components, cable separation methodology, associated circuits and emergency lighting that will be further reviewed by the NRC as a
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result of followup on the NRC findings contained in Inspection Report No.
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No violations or deviations were identified.
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6.
Open Items Open items are matters which have been discussed with the licensee, which
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i will be reviewed further by the inspector, and which involve some action l
on the part of the NRC, the licensee, or both.
Open items disclosed j
during the inspection are discussed in Paragraphs 3.B, 3.C, 3.0, and 3.E.
1 7.
Unresolved Items
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Unresolved items are matters about which more information is required in order to ascertain whether they are acceptable items, items of noncompliance, or deviations.
Unresolved items disclosed during the inspection are discussed in Paragraph 3.B and 4.B.
i 8.
Exit Meeting l
An exit meeting was held with the licensee's staff on September 25. 1987.
The items of concern in this report were discussed at that time ard ir, previous meetings with the licensee.
The licensee acknowledged th'
content and scope of the inspection findings.
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