ML20116G136
| ML20116G136 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 04/18/1985 |
| From: | Fiedler P GENERAL PUBLIC UTILITIES CORP. |
| To: | Kister H NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| References | |
| NUDOCS 8505010384 | |
| Download: ML20116G136 (5) | |
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Forked River.New Jersey 08731-o388 609 971-4000 Writer's Direct Dial Number:
April 18, 1985 Mr. Harry B. Kister, Chief Projects Branch No. 1 Division of Reactor Projects U.S. Nuclear Regulatory Commission Region I 631 Park Avenue King of Prussia, PA 19406
Dear Mr. Kister:
Subject:
Oyster Creek Nuclear Generating Station Docket No. 50-219 Inspection 85-01 Notice of Violation Response i
Pursuant to 10CFR2.201, the attachment to this letter contains our response to the Notice of Violation in Appendix A of your letter dated March 14, 1985. Due to management reviews, several additional days beyond the due date were necessary to finalize our response. Messrs. E. L. Conner and H. Kister, Region I, were notified on April 16, 1985 that our response would be submitted on or before April 18, 1985.
Should you have any questions concerning the information therein, please contact Mr. Michael W. Laggart at (201)299-2341.
Very truly yours, g'
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J Vice President and Director Oyster Creek PBF/PFC/ dam Attachment cc: Dr. Thomas E. Murley, Administrator Region I U.S. Nuclear Regulatory Consnission 631 Park Avenue King of Prussia, PA 19406 NRC Resident Inspector Oyster Creek Nuclear Generating Station Forked River, NJ 08731 8505010384 850418 fI*6/
I PDR ADOCK 05000219 N'
GPU Nuclear theatiutiis a subsidiary of the General Public Utilities Corporation I \\
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ATTACHMENT I l
Violation A 10 CFR.50, Appendix B, Criterion XVI, Corrective Action, requires tnat measures shall be established to assure that conditions adverse to quality such as deficiencies are promptly identified and corrected.
Contrary to the above, measures were not established to promptly evaluate, take corrective actions, and report a condition in which the containment spray heat exchangers pressure was being maintained in a condition contrary to that described in the Facility Description and Safety Analysis Report (containment spray water pressure higher than the emergency service water pressure).
Response
We agree with the Notice of Violation (NOV) in that contrary to the description contained in the Facility Description and Safety Analysis Report, containment spray (CS) pressure was higher than the emergency service water (ESW) pressure, and that this condition was not promptly evaluated, corrected or reported.
The NOV requires us to submit, in accordance with 10CFR2.201 a written
. statement which includes:
(1) the corrective steps which have been taken and the results achieved; (2) corrective steps which will be taken to avoid
.further violations; and (3) the date when full compliance will be achieved.
Prior to discussing-the efforts being taken to improve our system to evaluate, take corrective action, report a condition adverse to safety, we would like to describe the action taken with respect to the technical problem associated with this NOV. Attached is a copy of Licensee Event Report (LER)84-026 which was submitted on November 20,.1984. This document discussed both the evaluation and acceptability of the existing CS-ESW delta pressure condition, and the future corrective actions we have planned. The i
LER addresses and satisfies the technical concerns of the negative delta
. pressure condition regarding dose assessment and as such represents our
' response to the violation for corrective steps which have been taken and the results achieved.
It is important to emphasize that although the dose assessment concluded that doses caused by possible CS heat exchanger leakage l
is negligible, efforts are continuing to detennine and correct the negative l
delta pressure condition.
~ In response to the corrective steps which will be taken to avoid further L
violations, and when full compliance will be achieved, a'ctions are in effect j
or are being planned, to address that portion of the violation which states l
.that ".... measures were not established to promptly evaluate, take L
. corrective action, and report a condition....".
l l
- Your. letter of March 14, 1985, which transmitted this NOV, requested that we expand our reply to this violation to address the continuing problem of tracking and implementing identified problems. Our response to NRC Region I f-inspection 84-28, dated March 15, 1985, provided the supplementary information you requested on our commitment tracking system.
In that response we stated the tracking system identifies and lists commitments, and provides summary reports to personnel responsible for implementation and to their respective management.
It is not, however, a control system which ensures the action will be completed. As would be expected, appropriate i
I management action; based on information supplied througn the tracking system report, is.necessary for completion of a commitment.
In this.
. instance, the CS-ESW negative delta pressure was tracked as an open NRC unresolved item from 1978.
Please note that the tracking system described in procedure LP-002 " Regulatory Correspondence Management and Commitment Control" did not exist during 1978.
This item was extracted from the management tracking system in effect during 1978.
The tracking system can be credited in part for the resolution of this item. _ It was through this system that QA Site Audit S-0C-82-14 identified and issued their Audit Finding Nonconfonnance whicn stated " corrective action to evaluate and close out NRC unresolved item 78-19-02 had not been implemented."
In order to assess if NRC concerns are being actively addressed, open NRC items are being compared with tnose items on our commitment tracking system.
Each open item will be evaluated to ascertain if resolutions are properly planned and scheduled.
It is our intention to complete this review by August 30, 1985.
In conjunction with this review, documentation for completed items is being forwarded to the resident inspectors.
A meeting was held by the Vice President of the Technical Functions division to investigate why corrective action had not been implemented in a timely fashion. During this meeting, the Vice President directed that a sequence of events and formal critique be prepared for this occurrence. The critique will De reviewed to determine if any additional measures should be implemented.
Your cover letter also identified another NRC unresolved item relating to the implementation of the required reactor coolant leakage reduction program and stated that this is an additional example of failure to meet docketed commitments. The item was regarded as complete by the responsible implementing manager, however the inspector pointed out that a portion of the total program was' not being implemented. _ During the exit meeting for inspection 50-219/85-01 this item was discussed and an additional measure of assurance was established. For those NRC committed to or required actions that the licensing action item system considers complete and for which verification cannot be established via the action item system's files, the completed actions would.be forwarded to the site Quality Assurance Department for eventual verification.
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Violatien B Technical Specification 6.8.1 requires, in part, that written procedures be established, implemented, and maintained.
Paragraph 3.8.7.2 of Station Procedure 113, Conduct of Installed Instrument Surveillance, Calibration, and Maintenance, requires that a trip signal be inserted into a channel undergoing surveillance if its sensing instrument is valved out of service for greater than one hour.
Contrary to the above, on January 10, 1985, a reactor water level triple low level sensor was valved out of service for 71 minutes and no trip signal was inserted.
This is a Severity Level IV violation (Supplement I).
Response
We concur with the violation, as stated.
The event which resulted in this violation was reported in detail in Licensee Event Report 50-219/85-001. The primary cause of the event was personnel error, on the part of both control room personnel and instrument technicians.
The instrument technicians logged the time out of-service for the sensor but were not aware of the one hour Technical Specification limit. Control room personnel were aware that the sensor was out of service; however, they did not log the time out-of-service. As a contributing factor, the surveillance procedure used by the instrument technicians did not specifically mention the one hour Technical Specification limit.
The subject Oyster Creek Technical Specification states, "When necessary to conduct tests and calibrations, one channel may be made inoperable for up to one hour per month without tripping its trip system."
We consider the time limitation of this specification too restrictive. A comparison of Oyster Creek Technical Specifications with BWR Standard Technical Specifications (NUREG 0123, Revision 3) was performed and we conclude that the added flexibility for instrument surveillance out-of-service times prescribed by the Standard Technical Specifications'are desirable for Oyster Creek.
i Standard Technical Specifications allow up to two hours for instrument calibration and surveillance without inserting a trip into a sensor's associated trip system.
In addition, there is no per month stipulation as in Oyster Creek Technical Specifications. We feel there is adequate justification to change the applicaDie Technical Specifications and a request for change will be submitted.
The following corrective steps have been taken:
1.
Control room personnel have been instructed to refamiliarize themselves with the protective instrumentation requirements in Technical Specification Table 3.1.1.
2.
Control room operators have been instructed to closely monitor instruments having out-of-service time limits during surveillance testing. Times will be tracked by operators as close as practicable. Control room personnel have been directed to ensure the one hour limitation of Technical Specification table 3.1.1 is met, either by returning the instrument to service or having the instrument channel placed in the tripped condition, providing that such action does not cause the total trip function to occur.
5 3.. Lead control room operators have been instructed to log all surveillance tests in the control room log,-recording the start time, completion time and test results.
4.
An " Equipment-Out-of-Service Sheet" has been added to each applicable surveillance which requires instrument technicians to check out-of-service times prior to the' start of surveillance activity.
5.
A memorandum concerning the subject of keeping Technical Specification time-limited equipment out-of-service beyond the specified times has been reissued as required reading to instrument technicians and supervisors.
6.
The department responsible for perfonning surveillances is tracking surveillance completion dates to ensure surveillances are not executed twice within a 30 day period if out-of-service times are significant.
7.
~A copy of the Notice of Violation together with instructional memoranda are currently being distributed to all licensed operators and instrument technicians and their supervisors as required reading.
Corrective steps which will be taken to avoid further violations include the i
following:
Oyster Creek Procedure' ll6, which administers the conduct of the
- 1..
surveillance program, will be revised to formally incorporate the addition of the " Equipment-Out-of-Service Sheet" to each applicable o
surveillance.
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2.
All instrument surveillance test procedures are reviewed against a
. quality assurance procedure review check list specifically written l
for surveillance test procedures. The check list requires that the surveillance procedure include documentation of out-of-service and return-to-service times. Existing instrument surveillance L
procedures have been reviewed to verify that out-of-service and return-to-service times are documented.
l L
Full compliance has been achieved with the introduction of the l'
" Equipment-Out-of-Service-Sheet." The revision to Procedure 116 is expected to be approved and fully implemented by May 10, 1985.
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Surveillance test results since November 10, 1977 indicate that the delta pressure between the Emergency Service Water (tube side) and Containment Spray (shell side) is negative tube-to-shell in the Containment Spray heat exchangers. This condition is contrary to that described in the Facility Description and Safety Analysis Report. A negative tube-to-shell delta -
pressure would allow leakage from the Containment Spray (CS) System to enter the Emergency Service Water (ESW) System at the CS heat exchanger and, thence, to the environment. The cause of the negative delta pressure is believed to be a decrease in ESW pump performance and increased pressure drop in ESW piping. Radiological consequences during nomal operation and accident conditions have been analyzed and the contribution to offsite dose is considered negligible.
Post-LOCA containment heat removal characteristics of the ESW/CS System are not affected. Corrective action will be determined pending the outcome of ongoing ESW System evaluation.
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The earliest record identifying this condition is a' surveillance test data sheet dated November 10, 1977.
IDENTIFICATION OF OCCURRENCE The Emergency Service Water (ESW) side of the Containment Spray (CS) heat exchangers operates at a lower pressure than the CS side.
This condition is reportable in accordance with 10CFR50.73 (a)(2)(fi)(B).
CONDITIONS PRIOR TO OCCURRENCE The reactor was in various operating and shutdown modes prior to and subsequent to the identification of this condition.
i OESCRIPTION OF OCCURRENCE The ESW System delivers cooling water to the tube side of the CS heat exchangers from the ultimate heat sink (Barnegat Bay via intake canal).
The CS System circulates demineralized water from the pressure suppression chamber (Torus) through the shell side of the CS heat exchangers to the Drywell for post-LOCA containment cooling.
Installed dalta pressure instrumentation provides ' indication of tube-to-shell side delta pressure in the CS heat exchangers.
Since November 10, 1977, the tube-to-shell delta pressure has been recorded as less than zero on surveillance test data sheets.
This condition is contrary to the system design basis as stated in Section VI-7.2 of the Facility Description and Safety Analysis Report (FDSAR) which states:
"Each of the four service water pumps will also have a capacity of about 3000 gpm but will deliver a higher head than the containment spray pump to maintain a higher pressure on the service water side of the heat exchangers so that any leakage will be into the containment side of the heat exchangers." Data resulting from pre-operational testing shows that positive tube-to-shell delta pressure was achieved as designed.
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I APPARENT CAUSE OF OCCURRENCE The apparent cause of this condition is believed to be gradual cegradation of ESW System performance since initial plant operation.
Some biofouling found in ESW piping resultin'g in increased pressure drop in ESW piping coupled witn a decrease in ESW pump perfomance cnaracteristics result in decreased ESW pressure at the CS neat exchangers.
ANALYSIS OF OCCURRENCE AND SAFETY ASSESSMENT The function of the CS and ESW Systems is to remove heat from the primary containment following a LOCA to control pressure and temperature and maintain them at acceptable levels in order to: 1) reduce the driving force for containment leakage; 2) maintain the structural integrity of containment.
The CS/ESW System transfers post-accident decay heat to the ultimate heat sink and provides a physical boundary to the release of post-accident fission products.
The CS/ESW System comprises two redundant loops each containing two heat exchangers, two ESW pumps and two CS pumps.
The safety function is achieved with operation of one CS pump, one ESW pump and two heat exchangers piped in parallel in one of the two loops.
The consequence of higher CS (shell side) pressure than ESW (tube side) pressure is that snould a leak develop in the heat exchanger then Torus water would leak into the ESW System. Radioactive material contained in Torus water would then ce discharged to the oischarge canal througn the leak.
Tne capability of the CS/ESW System to perfom its cooling safety function is not affected by the existing system condition. Adequate cooling water flow is provided by ESW pumps. As determined by calculation the minimum required flow is equal to 2370 gpm. Surveillance testing continually demonstrates ESW flow to be approximately 3000 gpm or greater. Should heat exchanger leakage occur, it will be minimal and will not affect CS cooling effectiveness.
The capability of the CS/ESW System to provide a boundary to the release of post-accident fission products has been affected by the negative tube-to-shell CS heat exchanger delta pressure. Leakage in the heat exchangers will not be contained within the CS System.
Leakage can escape the system boundary into the ESW System and, ultimately, to the discharge canal.
An evaluation has been perfomed to estimate the offsite dose due to leakage from the CS System during a loss of coolant accident (LOCA).
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Several assumptions were made and parameters chosen in evaluating this hypothetical leakage and the subsequent offsite dose.
The source term utilized was derived from the core inventory given in the Wash-1400' Reactor Safety Study, Appendix 6,8 curies. Calculation of Reactor Accident Conseque The figure utilized was 3.98E This activity was assumed to be all Iodine-131, even though the particulate activity was factored into the source term.
Due to tne release being unpressurized liquid, no consideration was given to noble gas activity, nor was consideration given to the various phqses of iodine activity and their associated characteristics.
All of the 3.98E8 curies was assumed to enter into the torus and be diluteo by Torus water (82,000 cubic feet). This is considered conservative in that all of the activity would not enter into tne Torus and the Torus water volume assumed is the minimum level allowed during plant operation. A further measure of conservatism was included in that no credit was taken for water in the CS System piping and heat exchangers.
It was assumed that the CS System was in service for one week after the LOCA.
Table i below:
The infonaation is further detailed on Table 1 Activity Released to Torus (Wash-1400) 3.98E8 Curies l
Torus Water Level 82,000 cubic feet 2.32E9 ML Activity of Torus Water l.72 E6 uCi/ML Leak Rate 0.03 gallons /nour 1.9E4 ML/ week Total Activity Released 3.27E3 Curies Dilution (Discharge Canal Flow) 2138.9 cubic feet /second
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Oyster Creek, Unit 1 0 1510101012 !Il9 8 (4 012I6 0l0 015 0F 0 17 The information in Table 1 was entered into a dose analysis computer program. This program computes dose for several pathways. The three pathways utilized for the dose calculation we.re: 1) exposure from snoreline deposition; 2) ingestion of salt water sport fish and; 3) ingestion of salt water invertebrates. These are the most likely pathways of offsite exposure from a sM1_1 into the discharge canal. Dose due to each exposure pathway was calcul_ated and then summed for eacn age group. The program is in accordance with Regulatory Guide 1.109.
As the dose model used considered exposure factors appropriate for evaluations during nomal plant operations (per Regulatory Guide 1.109) potential radiation dose values are, therefore, conservative. There was no credit.taken for actions which would be perfonned to limit exposure from the pathways identified. Plant procedures are in effect that specify actions to be taken to minimize offsite dose during an occurrence such as this hypothetical leak. Emergency Plan Implementing Procedure (EPIP) No. I will be implemented upon receipt of. a high radiation alann at the discharge of the heat exchanger (ESW side).
S1nce the radiation monitors are area radiation monitoring devices, their indications may not be entirely reliable. However, sampling will also be performed in the discharge canal in accordance with EPIP-ll to determine appropriate actions to minimize exposure to radioactive ef fluents. These actions would include reconsnendations to restrict fish and invertebrate ingestion if such protective actions are warranted and the isolation of the leaking containment spray heat exchanger based on ESW radiocctivity levels as determined via sampling. Limits of the sampling are based on 10CFR20 limits, which are more conservative tnan 10CFR100 limits.
Notification of civil authorities and the activation of the Environmental Assessment Command Center will be implemented by EPIP-3 and EPIP-35.
Protective action recommendations will be made per EPIP-2 to appropriate civil authorities to ensure that actions such as fish /invertibrate ingestion are implemented to ensure that any effect to the health and safety of the public is minimized.
These controls are enacted to minimize the ingestion pathway and will occur if radioactivity is detected in the canal.
Therefore, including ingestio,n pathway doses is inappropriate for this evaluation.
This is consistent with Regulatory Guide 1.3, which does not include the ingestion pathway doses post-LOCA.
Integral to the evaluation of the acceptability of the present system condition is the integrity of the tube and shell sides of the CS heat exchangers.
The original heat exchangers were replaced with ones containing 90-10 copper-nickel alloy tubes during the 1978-79 refueling outage.
These tubes developed pinhole leaks in approximately one year and were replaced with titanium tubes in 1980. The heat exchanger presently is constructed of titanium tubes with an aluminum-bronze tube sheet. The corrosion resistance of these materials against sea water is excellent. Therefore, the probability of any gross failure of tubing is small.
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the heat exchangers have been leak tested before restart from the recent refueling outage. The criteria for allowable leakage during the test was based on a small fraction of the exposure limits of 10CFR100 as outlined below.
The dose evaluation model used in calculating doses is based on~ normal operations when ingestion of fish and invertibrates is not restricted.
However, since dose cue to ingestion will not occur during an accident because of the plant controls discussed previously, the only dose factor will be direct shoreline exposure.
As noted in Table 1, the leak rate model used a 0.03 gpn CS water leak rate into ESW as being representative of a leak due to galvanic corrosion between the tube and tubesneet from all four heat exchangers.
To be more conservative and to specify a leakage criterion that can be accurately measured, a leak rate of 1 gpm was used as the input to the dose calculation.
This increases doses Dy a factor of approximately 2000.
Using the 1 gpm leakage and recalculating direct exposure dose the following table results:
Table 2 Offsite Doses Based on CS System Leakage Scaled to a Leak Rate of 1 gpm Direct Shoreline 10CFR100 i
Exposure (Rem)
Limits (Rem)
Whole Body 0.024 25 Tnyroid 0.024 300 As stated in " Safety Evaluation of SEP Topic XV-19,' Radiological Consequences of a Loss of Coolant Accident", the potential contribution to the LOCA dose of less t,han 0.1 rem is considered to be negligible.
Therefore, it is concluded that the doses caused by CS heat exchanger leakage is negligible.
Thus,1 gpm was the criterion for leakage permitted during the CS heat exchanger leak test although a higher limit could be justified through this same logic.
This leak rate limit applies to the total of all four heat exchangers and still provides a factor of 4 margin to what has previously been identified as an insignificant dose. The CS System is the only system L
operating after a LOCA which has this potential for liquid release to the environment.
The existing condition will not create significant environmental impact s
during normal operations.
surveillance testing during normal operations.The CS heat exchanger is in operation the CS and ESW systems is performed once per month. Surveillance testing of both During the surveillance g..
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2 test, both CS and ESW systems operate simultaneously with one pump operating in each system. One loop is tested at a time. The pumps run for about 15 to 20 minutes, during which. time they staoflize and the readings for flow rates and pressures are taken.
If a CS heat exchanger develops a leak during testing, leakage to the environment could occur.
However, since the duration of the test is very short and the Torus water chemistry and radiation level is on the order of 10-4 uCf/ml (liquid isotopic concentrations) the contribution from leakage to offsite releases will be approximately nine orders of magnitude less than the values calculated for the accident scenario, and therefore will be negligible.
CORRECTIVE ACTION During the recent refueling outage steps were taken to assess the operational condition of the ESW System as follows:
1)
An examination of selected portions of ESW piping was performed.
Some biofouling was noted near the ESW pumps at the intake structure, but the extent of the biofouling was considered to not nave a significant effect on flow rate.
ESW piping examined in the turbine and reactor buildings was found to be clean.
ESW piping is chlorinated on an intermitent basis to minimize further biofouling.
2)
One of the four ESW pumps was refurbished. A minor improvement in performance was noted, however, indicated performance characteristics are still celow the design pump curve.
3)
As it is felt that current ESW flow instrumentation might provide a conservative indication of actual ESW flow rate, an annubar flow measuring device was installed in one ESW loop to verify the accuracy of portable ultrasonic flow instrumentation. However, problems were
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encountered in obtaining a reliable measurement with the annubar.
Efforts will continue to resolve these hardware problems.
4)
Two of four restricting orifices in ESW discharge lines from the CS heat exchangers were examined and found to be in good condition.
Their correct sizing was verified.
Further corrective action will be based upon the ongoing evaluation of ESW System performance.
The CS heat exchangers were checked for leaks prior to entering the plant's post-outage startup phase and were found to have no leakage.
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