ML20214K722

Advises That Due to Unavailability of Software & Hardware, Util Proposes to Phase Implementation of Radiological Effluent Tech Specs,Per 841022 Rev 1 to Tech Spec Change Request 69.Revised Pages to Proposed Tech Specs Encl

ML20214K722
Person / Time
Site:	Oyster Creek
Issue date:	08/13/1986
From:	Fiedler P GENERAL PUBLIC UTILITIES CORP.
To:	Zwolinski J Office of Nuclear Reactor Regulation
References
NUDOCS 8608220064
Download: ML20214K722 (14)

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' Cw% Qf ol-GPU Nuclear Corpcration NggIgf Post Office Box 388 Route 9 South Forked River.New Jersey 08731-0388 609 971-4000 Writer's Direct Dial Number:

August 13, 1986 Mr. John A. Zwolinski, Director BWR Project Directorate No.1 Division of Boiling Water Reactor Licensing U.S. Nuclear Regulatory Commission Washington, DC 20555

Subject:

Oyster Creek Nuclear Generating Station Docket No. 50-219 Radiological Effluent Technical Specifications (RETS)

By letter dated October 22, 1984, GPU Nuclear submitted Technical Specification Change Request (TSCR) No. 69, Revision 1, which contains the proposed RETS for Oyster Creek. On July 31 and August 1,1985, meetings were held between members of my staff and Mr. Jack Donohew, the NRC Project Manager for Oyster Creek, to discuss implementing various aspects of the RETS.

Implementation problem areas involve the development of computer software required to utilize the offsite dose projection methodology contaiaed in the Offsite Dose Calculation Manual (0DCM) and the unavailability of some monitoring equipment to monitor certain effluent pathways as required by the RETS.

Due to the currently unavailable software and hardware, we propose to phase the implementation of the RETS. A phased implementation would allow near term implementation of those portions of the RETS which do not require computer software development or hardware availability and allow implementation of the remaining specifications as hardware modifications and software are completed. The problem areas, including our proposed implementation schedules, are provided below:

A. Software In order to utilize offsite dose assessment methodology contained in the ODCM, computer software needs to be developed. Software development had been deferred pending NRC staff concurrence with the proposed specifications. With the staff approval of our proposed RETS, software development has been underway with completion scheduled before the end of the current refueling outage. Therefore, we intend to implement those sections in the RETS which are affected by 0DCM computer software availability prior to restart from the current (Cycle llR) refueling outage, which is scheduled for approximately mid-October 1986. The -

affected RETS sections, to be implemented prior to restart from the 11R 4t< /dCU outage, are as follows: ' I 8608220064 860813 g g [#1[ IdI8 DR ADOCKOSOOg9 q GPU Nuclear Corporation is a subsidiary of the General Public Utilities Corporation

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1. 3.6.J and 4.6.J i

2. 3.6.K and 4.6.K -y

3. 3.6.L a

4. 3.6.M and 4.6.M l

5. 3.6.N and 4.6.N 4 B. Radioactive Liquid Effluent Monitoring Instrumentation 9_

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The three liquid effluent radioactivity monitors listed in proposed RETS

< - Table.3.15.1 are currently inoperable. These conitors are the liquid i radwaste effluent discharge monitor, ,th,e Turbine Building Sump No.1-5 radioactivity monitor and the service wa',er system discharge line -

radioactivity monitor. The radwaste lit,0id effluent discharge monitor has "

been out of service since 1981. Plant personnel have utilized sampling l and analysis of the batch discharges made via this pathway as allowed by ,

our current Technical Specifications (3.6.B.2). We believe adequate justification exists to delete the requirement for the radwaste liquid a effluent monitor. A Technical Specification Change Request to propose the i deletion of the radwaste liquid effluent discharge monitor is planned for j sutmittal by September 30, 1986. 3

~ . q Installation of the radioactivity monitor for discharges from Turbine Building Sump No.1-5 has been recently completed. We expect this monitor [

to be ready for service prior to restart from the current (Cycle llR) -

reft;eling outage. f J

The third liquid effluent radioactivity monitor is the Service Water  ;

System discharge line monitor. This monitor, which is original plant i

equipment, requires some equipment upgrsde and further testing to ensure [

that the monitor will be operable under all background conditions and to -

verify its calibration. Although the Service Water radioactivity monitor is considered inoperable with respect to conformance with RETS

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requirements, it will remain in service until the upgrade is complete; however, the alarm setpoint cannot be established in accordance with the ]

ODCM due to problems with sensitivity and background radiation. An alarm i setpoint will be established for the present monitor which will minimize --

the receipt of spurious alarms and provide an indication that effluent _

radioactivity has increased to an unacceptable level during the time ,

between effluent sampling and analysis. The upgrade of this monitor is -

currently planned to be completed during the next (Cycle 12R) refueling il

(- 1 outage. 3 -

l The sections in the RETS which are affected by inoperable liquid effluent 9 monitoring instrumentation are as follows:

1. 3.15. A.1 and Table 3.15.1, Item 1.a, b and c (equipment i operability) j 3.lS.A.2 and 3 (equipment operability)

2. 1 3, 4.6.I.2 (equipment operability) J

4. Table 4.15.1, Items 1.a, b and c (equipment surveillance) y iii E

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C. ~ Gaseous Effluent Instrumentation

Hardware and software problems currently exist which preclude the ability of our radioactive gaseous effluent =cnitoring in

tru entation to i completely satisfy the equipment operability requirements in the proposed RETS for monitoring the stack and turbine building effluent pathways.

This monitoring system, which has been referred to as RAGEMS, was designed in 1977/1978 to meet Appendix I to 10 CFR 50. It was modified to meet post-accident monitoring and sampling requirements in NUREG-0737.

Originally, RAGEMS was to provide on-line analysis to identify the principal noble gas radionuclides; however, this capability _is not functional . Attachment I contains a proposed reworded basis for specification 3.15.B to reflect the system's capabilities as they have l

evolved since our October 22, 1984 submittal and to remove references to RAGEMS.

At present, ventilation fan capacity curves are used to estimate stack effluent flowrate. Flow measurement will be provided during operating cycle 11.

The instrumentation for continuous monitoring of the turbine building ventilation exhaust effluent is scheduled for modifications during the next operating cycle (Cycle 11). Approximately one half of the turbine building ventilation flow is exhausted to the main stack, which is monitored. The other half is exhausted from the turbine building vents.

2 This flow normally comes from the feedwater and condensate pump area and the turbine lube oil bay. The turbine building operating floor vent will be administrative 1y closed during power operation. If this vent is needed due to operational concerns, or if it is necessary to operate this vent to

-support the modification / testing of the turbine building gaseous effluent monitoring system during the next operating cycle, appropriate administrative controls will be applied.

At the August 1,1985 meeting, it was stated that modifications to provide stack effluent flow measurement and continuous monitoring of the turbine building vent gaseous effluent would be accomplished during our current

11R outage, if management review resulted in concurrence. After further review of this addition to the planned workscope of the llR outage, it was decided that available resources and lead time cannot~ support the

scheduling of this work during the 11R outage without significantly impacting outage duration. The llR outage is currently scheduled for a i' six month duration. ,

The specifications which are affected by currently unavailable radioactive gaseous effluent monitoring instrumentation are as follows:

1. 3.15.B.1 (equipment availability)

2. 3.15.B.2, 3 (equipment availability)

3. Table 3.15.2, Items 2.d and 3.are (cquipment availability)

4. Table 4.15.2, Items 2.d and 3.are (equipment surveillance)

In addition to the above, Attachment 2 contains six revised pages for the  !

proposed RETS reflecting minor editorial changes as follows:

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1) The word " dose" was inadvertently omitted in the fourth paragraph in the basis section for specification 3.6.E.

2) Action 121 in the Table 3.15.2 notations should be deleted as it is not applicable to Table 3.15.2.

3) ODCM is incorrectly spelled in specification 3.15.A.2.

4) Installation date should be removed from Item 1.C in Table 3.15.1.

5) Table 3.15.1, versus 3.1.5.1, is the correct reference in specification 4.6.I.2.

6) The phrase "after the holdup line" should be added to Note e of Table 3.1.1. These words were inadvertently omitted and will make Item I.1 of Table 3.1.1 consistent with specification 3.6.E and its associated basis.

Revised pages 3.6-10, 3.15-1, 3.15-3, 3.15-6, 4.6-2 and 3.1-12 are attached.

All. specifications, other than those previously discussed above, in the proposed RETS can be implemented at Oyster Creek within one month of issuance .

by NRC. This time is necessary in order for new procedures ard procedure revisions to be put in place and to allow for personnel training on these changes to be effected.

If you should have any questions, please contact Mr. Paul F. Czaya at (201) 299-2542.

Very truly yours, I

Vice President and Director 3yster Creek i PBF:gpa 2467f/0207A Attachments l

cc: Dr. Thomas E. Murley, Administrator Region I U.S. Nuclear Regulatory Commission 631 Park Avenue King of Prussia, PA 19406 Mr. Jack N. Donohew, Jr. , Project Manager U.S. Nuclear Regulatory Commission 7920 Norfolk Avenue Bethesda, MD 20014 NRC Resident Inspector Oyster Creek Nuclear Generating Station l

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i ATTACHMENT I l

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3.15-7 Basis: 3.15.A The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in licuid effluents during actual or potential releases of liquid effluents. The use of this instrumentation is consistent with the requirements of General Design Criteria 60-and 64 of Appendix A to 10 CFR Part 50.

Radioactivity monitors on the liquid radwaste effluent line and in the Turbine Building Sump No. 1-5 initiate .

a trip to stop the effluent discharge pump when the trip lsetpoint-is exceeded. The reactor service water ,

- system discharge line radioactivity monitor initiates

- an alarm in the reactor control room when the alarm setpoint is exceeded.

The alarm / trip setpoint for each of these instruments is calculated and adjusted in accordance with the i

methodology.and parameters in the 00CM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20.106.

Basis: 3.15.8 The radioactive gaseous effluent instrumentation is i

provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during releases of gaseous effluents. The alarm / trip setpoint for each of the nolle gas monitors is calculated and adjusted in accordance with the methodology and parameters in the 00CM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20.106. The instrumentation in  ;

Table 3.15.2 also includes provisions for monitoring f- hydrogen below the explosive level.in the offgas system downstream from the recombiner. The

• operability and use of this instrumentation is consistent with the requirements of General Design Criteria 60 and 64 of Appendix A to 10 CFR Part 50.

The offgas hydrogen monitor and the radioactive gas monitors for the condenser air ejector offgas, the stack effluent, and the offgas building exhaust l ventilation have alarms which report in the reactor '

control room. The offgas hydrogen monitor initiates a bypass of the Augmented Offgas System in the event the setpoint is exceeded.

The Stack and the Turbine Building exhaust ventilation effluent air are monitored by a radioactive gaseous effluent monitoring system. It can sample effluent for radioactive particulates, iodine, and noble

! gases. It can measure the gross concentration of radioactive noble gases. A grab sample of the '

effluent air will be taken at least once per month and analyzed for the principal noble gas radionuclides (Reference Table 4.6.2).

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3.15-8 The gross gamma activity concentration of noble gas in Stack effluent is displayed in the reactor control room. That channel also causes an alarm in the reactor control room in the event a high activity concentration setpoint.is exceeded. Low flow of sampled Stack effluent would also cause an alarm in the reactor control room. Although flow data may be collected by a computer the. sample flow and the sampled stream flow (Stack 1and Turbine Building vent) can also be observed at a display located near the monitoring instrument (in which case the channel continues to serve its essential function and remains OPERABLE.) If the noble gas activity concentration display and the associated alarm become inoperable in the reactor control room, then OCNGS will perform the appropriate. action according to Table 3.15.2.

Purging the drywell to purify its atmosphere may discharge most of the air and gases in a brief time.

Hence, the drywell is purged only when the radioactive noble gas monitor in the stack monitoring system is operating in order to ensure measurement of radioactive gases discharged.

Frequently, the drywell is vented to control its pressure. But since the release rate is comparatively small, the effluent is monitored as usual and the extra requirement in Table 3.15.2 Action 124 that is applied during purging is not imposed during drywell venting.

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3.1-12 TABLE 3.1.1-(Cont'd)-

• Action. required when minimum conditions for operation'are not satisfied.

Also permissible to trip inoperable trip system. 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.

See Specification 2.3 for Limiting Safety System Settings.

NOTES:

a. Permissible to bypass, with control rod block, for reactor protection system reset in refuel mode.

- b. Permissible to bypass below 800 psia in refuel and startup modes.

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c. One (1) APRM in each operable trip system may be bypassed or inoperable provided the requirements of Specification 3.1.C and 3.10.C are satisfied. . Two APRM's in the same quadrant shall not be concurrently bypassed except-as noted below or permitted by note.

2 Any one APRM may be removed from service for up to one hour for test.or calibration without inserting trips in its trip system only if the remaining operable APRM's meet the requirements of Specification 3.1.B.1 and no control rods are moved outward during the calibration or test.

During this short period, the requirements of Specifications 3.1.B.2, 3.1.C and 3.10.C need not be met.

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d. The (IRM) shall be inserted and operable until the APRM's are operable and reading at least 2/150 full scale.

e. Offgas system isolation trip set at f 2.1/E C1/sec where E = average gamma energy from noble gas in offgas after the holdup line (Mev). Air ejector i isolation valve closure time delay shall not exceed 15 minutes.

f. Unless SRM chambers are fully inserted.

g. Not applicable when IRM on lowest range.

h. One instrument channel in each trip system may be inoperable provided the circuit which it operates in the trip system is placed in a simulated tripped condition. If repairs cannot be completed within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> the reactor shall be placed in the cold shutdown condition. If more than one instrument channel in any trip system becomes inoperable the reactor shall be placed in the cold shutdown condition. Relief valve controllers shall not be bypassed for more than 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> (total time for all controllers) in any 30-day period and only one relief valve controller may be bypassed at a time.

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4 3.6-10 Basis: 3.6.E_ Some radioactive material.is released from the plant under controlled conditions as part of the normal operation of-the. facility. . Other radioactive material not normally-

, intended for release could be inac'vertently released. in the event of an accident. .Therefore, limits in 10 CFR Part 20 apply to releases during normal operation and limits in 10 CFR Part 100 apply to accidental releases.

t Radioactive gases from the reactor pass through the steam lines to the turbine and then to the main condenser where they are extracted by.the air ejector, passed through i

holdup piping and released via the plant stack _ preferably-after treatment in the Augmented.0ffgas System.

Radioactive materials release limits for the plant stack have been calculated using meterological data from a 400 ft. tower at the plant site. The analysis of these on-site meterological data shows that a release of radioactive gases after 30 minutes holdup'in the offgas system of 0.3 C1/sec., wculd not result in a whole body radiation dose exceeding the 10 CFR 20 v'alue of 0.5 rem per year.

y The Holland plume rise model with no correction factor was used in the calculation of the effect of momentum and buoyancy of a continuously emitted plume.

Independent dose calculations for several locations offsite-were made by the AEC staff from onsite meteorological data developed by the licensee and diffusion assumptions appropriate to the site. The procedure followed is described in Section 7-5.2.5 of " Meteorology and Atomic Energy - 1968", equation 7.63 being used. The results of.

these calculations were equivalent to those generated by the licensee provided the average gamma energy per i

disintegration for the assumed noble gas mixture with a 30 minute holdup is 0.7 MeV per disintegration. Based on these calculations, a maximum release rate limit of gross activity, except for iodines and particulates with half

lives longer than eight days, in the amount of 0.21/E i

curies per second will not result in offsite annual doses in excess of the limits specified in 10 CFR Part 20. The E determination need consider only the average gamma energy per disintegration since the controlling whole body dose isl due to the cloud passage over the receptor and not cloud submersion, in which the beta dose could be additive.

The above discussion does not take into consideration the reduction in release rate afforded by operation of the Augmented Offgas System.

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o 3.15-1 3.15 . Radioactive Effluent Monitoring Instrumentation Applicability: Applies to instrumentation whose function is to monitor aqueous and airborne radioactive offluents from the Station.

Objective: To assure that instrumentation to monitor radioactive effluents is OPERABLE when effluent is discharged or that means of measuring effluent is provided.

Specification 3.15.A Liquid Effluent Instrumentation

1. The radioactive liquid effluent monitoring channels listed in Table 3.15.1 shall be OPERABLE with their alarm / trip setpoints set to initiate alarm / trip in the event the limit of Specification 3.6.I.1 is exceeded.

2. The alarm or trip setpoint of these channels shall be determined and set in accordance with the method described in the ODCM. I

3. When a radioactive liquid effluent monitoring instrumentation channel alarm / trip setpoint is less conservative than required by the above specification, without delay suspend the release l of radioactive liquid effluents monitored by the l affected channel, or declare the channel inoperable, or change the setpoint so it is acceptably conservative, or provide for manual

initiation of the alarm / trip function (s).

4. When less than the minimum number of radioactive liquid effluent monitoring instrumentation j channels are OPERABLE, take the ACTION shown in Table 3.15.1 Make every reasonable effort to restore the instrument to OPERABLE status within l

30 days and, if unsuccessful, explain in the L next Semiannual Radioactive Effluent Release Report why the inoperability was not corrected .

in a timely manner, f The provisions of Specifications 3.0. A, 3.0.B. and 6.9.2 are not applicable.

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3.15-3 TABLE 3.15.1 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION Minimum" Instrument Channels Applicability Action Operable I. GROSS R'ADI0 ACTIVITY MONITORS

a. Liquid Radwaste Effluent Line 1 b 110

b. Reactor Building Service Water System Effluent Line 1 b 112

c. Turbine Building Sump No. 1-5 1 b 114' i

2. FLOW MEASUREMENT DEVICES

a. Liquid Radwaste Effluent Line 1 b 113

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+ -TABLE 3.15.2 NOTATIONS 3.15-6

-a. Channels shall be OPERASLE and in service as indicated except that a channel may be.taken out of service for the purpose of a check, calibration, test, maintenance or sample media change without declaring the channel to be inoperable.

-b. During releases _via this pathway.

c.- During Augmented Offgas Treatment System' Operation.

d. .One hydrogen and one temperature sensor.

e.- Monitor / sampler or an alternate shall be OPERABLE to monitor / sample-Stack effluent whenever the drywell is being purged. g ACTION 122 With no channel 0PERABLE, effluent releases via this pathway may continue provided the flow rate is estimated

'whenever the exhaust fan combination in this system is changed.

ACTION 123 .With no channel 0PERABLE, effluent releases via this pathway may_ continue provided a grab sample,is taken at least once per 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> and is analyzed for gross-

. radioactivity within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter.or provided an.

alternate monitoring system with local display is; utilized.

ACTION 124 With no channel OPERABLE, effluent releases via this pathway may continue provided a grab-sample is taken at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and analyzed for gross radioactivity within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or provided an alternate monitoring system with. local display is utilized. Drywell purge is permitted only when the radioactive noble gas monitor is operating.

ACTION 125 With one channel 0PERABLE, operation of the main condenser offgas treatment system may_ continue provided a recombiner temperature sensing instrument is operable. . When only one of.the types of instruments, 1.e., hydrogen monitor or temperature monitor, is operable, the offgas treatment.

system may be operated provided a gas sample is collected at least once per day and is analyzed for hydrogen within four hours. In the event neither a hydrogen monitor nor a recombiner temperature sensing instrument is operable when required, the Offgas Treatment System may be operated provided a gas sturple is collected at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and analyzed within the following 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

ACTION 126 With no channel 0PERABLE, effluent releases via this pathway may continue provided the required sampling'is initiated with auxiliary sampling equipment as soon as reasonable after discovery of inoperable primary sampler (s).

ACTION 128 With no channel OPERABLE, effluent releases via the sampled pathway may continue provided the sampler air flow is estimated and recorded at least once per day.

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1 4.6-2 4.6.EL Main-Condenser ~0ffgas Radioactivity

1. The gross radioactivity.in fission gases

discharged from the main condenser. air ejector shall_be measured by! sampling and analyzing the-gases a.. .at least once.per month, and

b. When the reactor is operating at more than 40 percent of rated power, within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after an--increase in the fission gas release via the air ejector of more than 50 percent, as indicated by the Condenser Air Ejector Offgas Radioactivity Monitor after factoring out increase (s) due to change (s) in thermal power level.

4.6.F_ Condenser Offgas Hydrogen Concentration The concentration of_ hydrogen in offgases downstream of the recombiner in the Offgas System shall be monitored with hydrogen monitoring instrumentation as described in Table 3.15.2.

4.6.G Not Used 4.6.H Not Used 4.6.I Radioactivity Concentration in-Liquid Effluent

1. Radioactive liquid wastes shall be sampled

-and analyzed according to the sampling and analysis program in Table 4.6-1.

Alternately, pre-release analysis of batch (es) of radioactive liquid waste may be by gross beta or gamma counting provided a maximum concentration limit of 1 X 10 7 A Ci/ml_in the discharge canal at the Route 9 bridge is applied.

2. The alarm or trip setpoint of each

. radioactivity monitoring channel in Table 3.15.1 shall be deterrrined on the basis of I sampling and analyses results obtained according to Table 4.6.1 and setpoint method in the ODCM and set to alarm or trip before exceeding the limits of Specification 3.6.I.