ML19260C523

Forwards Response to NRC Request for Info Re Implementation of NUREG-0578 Recommendations.Two Oversize Drawings Encl

ML19260C523
Person / Time
Site:	Oyster Creek
Issue date:	01/04/1980
From:	Finfrock I JERSEY CENTRAL POWER & LIGHT CO.
To:	Eisenhut D Office of Nuclear Reactor Regulation
References
RTR-NUREG-0578, RTR-NUREG-578 NUDOCS 8001070340
Download: ML19260C523 (64)

Text

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Jersey Central Power & Ught Cornpany

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g Madison Avenue at Punch Bowl Road e

Morristown, New Jersey 07960 (201)455-8200 January 4, 1980 Mr. Darrell G. Eisenhut United States Nuclear Regulatory Commission Washington, DC 20555

Subject:

NUREG 0578 Implementation Oyster Creek Nuclear Generating Station Docket No. 50-219

References:

a.

D. G. Eisenhut's September 13, 1979 letter to all operating nuclear power plants b.

H. R. Denton's October 30, 1979 letter to all operating nuclear power plants c.

I. R. Finfrock, Jr.'s October 19, 1979 letter to D. G. Eisenhut d.

I. R. Finfrock, Jr.'s November 21, 1979 letter to H. R. Denton e.

T. D. Keenan's (Chairman, BWR Owners' Group)

December 14, 1979 letter to D. G. Eisenhut f.

D. G. Eisenhut's November 14, 1979 letter to T. D. Keenan (Chairman, BWR Owners' Group)

References a and b require that operating nuclear power plants submit certain information regarding the implementation of the NUREG 0578 recommendations.

This letter and its enclosures are intended to satisfy the necessary require-ments and to clarify certain Jersey Central Power & Light Company (JCP&L) positions which have been modified since our previous responses (References c and d).

1.

Item 2.1.2 - Relief and Safety Valve Test Program JCP&L will participate with the BWR Owners' Group in conducting a test program which will be presented to the NRC prior to January 31, 1980 (Reference e).

2.

Item 2.1.3 - Safety and Relief Valve Position Indication Based upon the NRC's feedback on this item (Reference f), JCP&L has elected to monitor the position of the five relief valves and sixteen safety valves with a system using accelerometers as the sensor. The sensors and preamplifiers, which will be installed in the containment this outage (January-March 1980), will not be environmentally or seismically qualiNed; however, it is anticipated that they will be properly qualified by mid-summer 1980.

The lack of qualification 1694 198 34'O 8001070 A

Jersey Central Power & Light Company is a Member Of the General Pubhc * )tihties System

Mr. Darrell G. Eisenhut Page 2 January 4, 1980 is not considered a safety hazard since there are other indications that the valves are open. Each valve has an RTD at its discharge point, and drywell pressure and temperature would respond to an open relief valve. All equipment installed in the drywell will be seismically supported to ensure that it will not disable another piece of equipment.

3.

Item 2.1.4 - Containment Isolation A listing of essential and nonessential systems which penetrate the containment is provided in Enclosure 1.

4.

Item 2.1.6.a - Leak Reduction Program A summary description of Oyster Creek's program to reduce leakage from systems outside containment is provided in Enclosure 2.

5.

Item 2.1.6.b - Plant Shielding A report which assesses the post-accident dose rates at Oyster Creek is included as Enclosure 3.

The impact of these dose rates on post-accident plant operation outside the control room is being assessed.

JCP&L will have this review complete by March 1,1980. A meeting with the NRC's Systematic Evaluation Program (SEP) staff will be requested to review any major modifications and major shielding additions required to assure integration with SEP topics; such as, turbine missiles, seismic capabilities, pipe whip, etc., where possible.

6.

Item 2.1.8.a - Post-Accident Sampling a summary descript'on of the changes which are being designed to enhance post-accident sampling are provided in Enclosure 4.

7.

Item 2.1.8.b - Interim Methods for Estimatina Radioactive Releases A brief description of the methods used to estimate release rates for noble gases, radio-iodines, and particulates in the event that existing instrumentation goes offscale it provided in Enclosure 5.

8.

Item 2.2.2.b - On-site Technical Support Center _

The interim and long-range plans for the Technical Support Center are provided in Enclosures 6A and 6B, respectively.

1694 199

Mr. Darrell G. Eisenhut Page 3 January 4, 1980 9.

RCS Venting A description of the modifications planned to augment the present capability to vent the reactor coolant system is provided in.

We trust that this letter is responsive to your requirements at this time.

If additional clarification is required, please advise.

Very truly yours, 06Y/

Ivan R. Finfro

,J.

Vice President pk Enclosures 1694 200

ENCLOSURE 1 January 4, 1980 CONTAINMENT ISOLATION ESSENTIAL SYSTEMS:

Those systems at the Oyster Creek Nuclear Station that are required to mitigate the consequences of postulated accidents.

0"-ESSENTIAL SYSTEMS

Those systems at the Oyster Creek Nuclear Station that are not required to mitigate the consequences of postulated accidents.

The following systems penetrate primary containment and are essential:

a.

Emergency Condenser System b.

Control Rod Drive System c.

Core Spray System d.

Containment Spray System e.

Torus to Reactor Building Vacuum Breaker System The following systems penetrate primary containment and either perform non-essential functions within the containment or are not required in the initial response to an accident:

a.

Main Steam Systen b.

Feedwater System c.

Reactor Building Closed Cooling Water System d.

Instrument Air System e.

Demineralized Water System f.

Reactor Cleanup System g.

Shutdown Cooling System h.

Liquid Poison System 1.

Drywell Equipment Drain Tank

j. Drywell Sump k.

Drywell & Torus Atmosphere Control Systems 1.

Reactor Recirculation Loop Sample System m.

Reactor Head Cooling System n.

Nitrogen System o.

Traversing In-Core Probe System Each of the systems fall into the following categories:

a.

Systems administratively isolated during normal operation.

b.

Systems presently provided with isolation provisions.

c.

Systems requiring modifications in order to provide isolation provisions. This will be accomplished during the 1980 refueling outage scheduled to begin January 5, 1980.

1694 201

ENCLOSURE 2 January 4,1980 JERSEY CENTRAL POWER & LIGHT COMPANY REPORT TO THE NUCLEAR REGULATORY COMMISSION IN RESPONSE TO NUREG-0578, SECTION 2.1.6.a DESCRIBING OYSTER CREEK NUCLEAR GENERATING STATION'S LEAK REDUCTION PROGRAM 1694 202

INDEX Page I.

INTRODUCTION................................................

1 II. PROGRAM SC0PE...............................................

1 III. 1980 REFUELING OUTAGE LEAK REDUCTION PROGRAM................ 2 A.

Leakage Identification and Measurement Methods.......... 2 B.

Leakage Reduction Measures..............................

4 IV. LONG TERM L EAKAGE REDUCTION PR0 GRAM......................... 4 A.

Preventative Maintenance Program........................

4 B.

System Design Review....................................

5 C.

System Design Change Implementation..................... 5 1694 203

I.

INTRODUCTION NUREG-0578, Section 2.1.6.a. discusses the integrity of systems outside the primary containment that will or may have to function during a serious transient or accident with large radioactive inventories in the fluids they possess. The " Lessons Learned Task Force" has determined that a more positive control and knowledge of the leakage rates of these systems is needed to provide the operating staff with the maximum usable equipmee and to restrict or control the release of radioactive materials to the environment.

Oyster Creek has in existence two ongoing programs to minimize and control leakages of radioactive fluids from systems which have the potential of containing highly radioactive fluids. The first of these existing programs is the containment leakage test program in accordance with 10CFR50 Appendix J.

The " Appendix J" tests performed at Oyster Creek are designed to identify and measure leakages from the primary containment and systems in normal operation connected to the primary system.

Excessive leakages found are repaired and retested so that total leakage is within allowable limits as specified in the Technical Specifications. The second existing program at Oyster Creek is that of regular operator equipment inspection tours.

Operators make routine inspections of all accessible equipment. As part of this inspection, any observed leakage is noted and job orders are initiated.

Identified leaks are repaired by the maintenance staff as promptly as feasible in order to minimize the burden on radwaste pro-cessing systems and reduce the probability of airborne contamination in the environment.

The Oyster Creek NUREG-0578 Leak Reduction Program will augment the existing programs by implementing all practical leak reduction measures for systems that could carry radioactive fluid outside the containment and to the extent practical, measure actual leakage rates with each system in operation. This program will include a preventative maintenance program to reduce and maintain leakage to as-low-as practical levels.

This report will describe Oyster Creek's NUREG-0578 Leak Reduction Program and schedule of implementation. Generally, all practical leak reduction measures will be implemented during the upcoming refueling outage comencing in January. All feasible design changes will be implemented by the end of the refueling outage scheduled to commence on April 15, 1981 subject to the availability of procurement of long lead items.

II. PROGRAM SCOPE All systems which penetrate the primary containment were considered for inclusion in the augmented Leak Reduction Program. A listing of the systems included and not included in the program are provided below:

A.

Systems Included in the Leak Reduction Program The portions of the following systems which are outside the primary containment will be covered by the augmented Leak Reduction Program:

1694 204 1

1.

Shutdown Cooling 2.

Isolation Condenser 3.

Core Spray 4.

Containment Spray 5.

Reactor Water Cleanup 6.

Standby Gas Treatment 7.

Reactor Coolant Sampling 8.

Reactor Coolant System Instrumentation 9.

Drywell Equipment Drain

10. Drywell Floor Drain

11. Reactor Building Equipment Drain

12. Reactor Building Floor Drain B.

Systems Not Included in the Augmented Leak Reduction Program are:

1.

Reactor Head Cooling 2.

Liquid Poison 3.

Feedwater 4.

Instrument Air 5.

Demineralized Water 6.

Closed Cooling Water 7.

Main Steam 8.

Control Rod Drive 9.

Nitrogen

10. Traversing In-core Probe

11. Drywell and Torus Atmosphere Control and Vacuum Relief III.

1980 REFUELING OUTAGE LEAK REDUCTION PROGRAM During the refueling outage commencing January 5, 1980, Oyster Creek will identify and measure leakage from the systems specified in Section II of this report and will implement all practical leakage reduction measures.

A.

Leakage Identification and Measurement Methods 1.

Leakage Identification 1694 205 a

All valves, pumps, flanges, fittings, relief valve discharge lines, and drain and vent lines on all accessible portions of the systems specified in Section II, including instrument loops will be visually inspected for evidence of leakage to the environment. Where practical, insulation will be removed to expose the component directly. The isolation valves of all auxiliary systems which interface with these systems will, to the extent practical, be inspected for through leakage. For those systems within the scope of this program, a review of maintenance history records shall be made to identify chronic leakage problems which may not be occurring at the time of this leakage inspection..

Records shall be maintained specifically to document leaks identified on these systems during the outage. The records shall identify system, component, leakage path, leakage rate, and measurement method. These records will be maintained and used as part of the long-term Leak Reduction Program to identify chronic and generic leakage problems.

2.

External Leakage Measurement External leakage is leakage of the system process fluid through a component part to the surrounding environment or to an open drain.

Leakage may be measured by one of the following techniques:

a.

A visual inspection for zero leakage by the absence of evidence of liquid leakage such as wetness, stains on the component, or liquid on the floor.

b.

Accumulating all external leakage into a container and measuring volume increase over a period of time.

c.

Whenever accumulating the leakage is not practical, a visual leakage rate estimate may be made, eg,10 drops per minute.

d.

When it is known or can be reasonably assumed that isolation valves of a system are leak tight, external leakage can be measured by local leak rate testing methods, i.e., measurement of fluid or gas makeup or pressure decay of fluid or gas of known volume, e.

A qualitative determination of leakage can be made by tracer gas leak detection or mass spectrometer.

NOTE:

In areas of high radiation, personnel exposure is to be kept as low as reasonably achievable.

In these areas, estimates of leakage may be made.

3.

Internal Leakage Measurement Internal leakage is leakage of the system process fluid through a valve or valves to or from other parts of the system or to or from other systems. Leakage of heat exchanger tubes is internal leakage.

Various methods will be used to detect and measure system internal leakage where practical at all system isolation valves and heat exchangers.

1694 206 3

a.

If local test taps are available, an LLRT-type test may be performed on the upstream side of the valve with the downstream side capable of accepting leakage, i.e., drained and vented or at reduced pressure.

b.

If an LLRT-type test is not practical, the downstream side of the valve may be drained and vented and through leakage collected from a drain into a container and volume increase measured over a period of time or pressure decay of the upstream side measured.

c.

Where it is not practical to directly measure internal leakage, an examination of plant records of interfacing systems for evidence of internal leakage may be made.

d.

A qualitative @ termination of leakage may be made by listening to valves with a stethoscope for the characteristic high frequency sound of valve seat leakage.

B.

Leakage Reduction Measures All practical leak reduction measures will be taken during the Oyster Creek 1980 refueling outage to reduce the leakage from the systems within the scope of this program identified to have significant leakage. These measures may include tightening packing followers, repacking valves, installing new pump seals, plugging heat exchanger tubes, replacing gaskets, lapping valve seats, capping leaking test lines, seal welding, etc. The technique utilized will depend on the specific requirements of the leakage problem and the availability of spare parts. Upon completion of the leak repair, leak testing will be repeated as described in Section IV.A.

For leakages which cannot be repaired during the outag7, aparopriate measures will be taken to control and contain the leak unti: it can be repaired.

IV. LONG-TERM LEAKAGE REDUCTION PROGRAM The long-term leak reduction program consists of two parts. The first part places emphasis on the problems associated with leakage through components into the environment or into other systems. The second part will be a design review of plant systems for leakage potential.

If deficiencies are found, modifications of the plant system will be considered.

A.

Preventative Maintenance Program The preventative maintenance program to minimize system leakage to the environment will include the following:

1.

Regularly scheduled system walkdowns through all accessible areas by operators to identify and report leakages on a piece part basis.

These leakages will be logged and categorized as to urgency of repair.

1694 207 4

2.

A defined leakage test program to include component systen and integrated leak rate testing at a frequency not to exceed each refueling outage.

3.

Periodic review of leakage logs to identify generic and chronic problems and the development of practical solutions to these problems such as improved packing, seal injection, leakage collection systems, etc. The solutions may also include changing to components with greater leakage control such as bellows seal valves or canned rotor pumps.

B.

System Design Review A thorough review of plant systems will be made to identify points at which highly radioactive fluids which might be present in a system following an incident could be transferred to other systems or areas of the plant thereby decreasing accessibility or operability during the recovery operation. The following changes will be considered as part of this design review:

1.

Automatic actions to prevent highly radioactive fluids from entering other areas of the plant.

2.

Automatic actions to prevent higiily radioactive fluids from escaping to the outside plant environment.

3.

System design changes to better control and manage highly radioactive fluids including improved gasket and seal materials.

4.

Procedural changes to better control and manage highly radioactive fluids using existing systems.

C.

System Design Change Implementation The system design changes which are deemed essential as a result of the system design review will be implemented as soon as practical. All feasible design changes will be implemented by the end of the refueling outage scheduled to commence on April 15, 1981 subject to the availability of procurement of long lead items.

1694 208 5

4 ENCLO5URE 3 January 4, 1980 CALCULATION OF POTENTIAL POST-ACCIDENT DOSE RATES AT THE OYSTER CREEK NUCLEAR GENERATING STATION FOR NUREG-0578 By: Kenneth R. Goddard Michael W. Laggart Jersey Central Power S Light Company Morristown, New Jersey December, 1979 1694 209

Calculation of Potential Post Accident [7se Rates at the Oyster Creek Nuclear Generating Station for NUREG-0578 CONTENTS 1.

Introduction 2.

Calculation of Volumetric Source Terms 3.

Calculation of Direct Radiation Dose 4.

Airborne Radiation 5.

Control Room 6.

Sumary and Conclusions 7.

References TABLES 1.

Energy Groups 2.

Time Points for Calculations 3.

Mass Attenuation Coefficients 4.

Macroscopic Cross Sections 5.

Fluence Rate to Dose Conversion Factors 6.

Isotopos Explicitly Considered in Analysis 7.

Volumetric Source Term - Containment Atmosphere 8.

Volumetric Source Term - Containment Water (excluding Noble Gases) 9.

Volumetric Source Term - Containment Water (including Noble Gases) 10.

Reference Pipe Characteristics 11.

Infinite Dose Factors 12.

Reactor Building Post Accident Dose Rates FIGURES 1.

Reactor Building 23'-6" elevation 2.

Reactor Building 51'-3" elevation 3.

Reactor Buil. ding 75'-3" elevation 4.

Reactor Building 95'-3" elevation 5.

E-W Section Through Control Room 6.

N-S Section Through Control Room 1694 210

1.

Introduction Section 2.1.6.b of NUREG-0578 requires a design review of plant shielding for post-accident operations. Additional guidance was provided in a letter dated October 30, 1979.

Oyster Creek, in common with other BWR plants, is designed to mitigate the effects of major design basis events without requiring access outside of the control room.

Because of this design feature, no effort was ende to assure accessibility of any part of the reactor building following an accident involving significant fuel damage.

In order to assess the accessibility of areas outside of the control room, and to review the habitability of the control room itself, calculations were performed of potential dose rates due to direct radiation from contaminated piping and equipment, and an estimate was made of potential levels of airborne radioactivity in the reactor building. Scattered radiation was not calculated, nor was radiation from the contaminated water in the torus. Both could considerably increase general radiation levels in areas not otherwise exposed to direct radiation from piping and equipment. The reactor building 23 foot elevation would be most affected.

As stated in I. R. Finfrock's letter of October 19, 1979, this review focuses on personnel safety and does not address the design of new ghielding, the redesig~n of existing shieldwalls, and the identification of equipment requiring improved environmental qualification, which will be completed at a 1 ster time.

16:\\ 2\\\\

2.

Calculation of Volumetric Source Terms The GE Source Term Information (Reference 3) was used as basic input for the source term calculation.

It is in units of curies per thermal megawatt, multiplied by 1 for noble gases (Kr and Ze), 0.5 for halogens (I and Br), and 0.01 for all other fission products. Since the GE data is for 452 isotopes, the amount of data entry and calculation required was reduced by considering explitly in this analysis only the 142 most important isotopes (Table 6). The GE source term for each of these isotopes for seven points in time following shutdown (Table 2) was multiplied by 1930 MWt (the Oyster Creek power level).

The gamma spectrum for each of the 142 isotopes was taken from the Radiological Health Handbook (Reference 2). Where gamma intensities were not available from the Handbook, they were conservatively estimated. The gamma rays were grouped into eight energy groups for calculational purposes (Table 1).

Three volumetric source terms were calculated. Each is a table cf gammas per second per cubic centimeter for each energy group and each point in time. The containment air source term contains 100% of the noble gases and 25% of the halogens diluted in a volume of 306,000 cubic feet (the combined free volume of the drywell and torus). The first containment water source term contains 50% of the halogens, no noble gases, and 1 % of the remaining fission products distributed in 89,000 cubic feet (the combined volume of water in the torus and the reactor coolant system). The second containment water source term contains 100% of the noble gases, 50% of the halogens, and 1% of the remaining fission products distributed in the same volume of water.

Since only 142 of the 452 isotopes were used in this derivation, the source terms were increased to account for the remaining isotopes.

The required correction factors were found separately for each group of isotopes (noble gases, halogens, other) and for each point in time by forming the ratio of the total curies in the GE data to the subtotal of curies of the 142 isotopes. The corrections were generally small, except at time 0 when considerable quantities of very short lived isotopes are present.

The final volumetric source terms used in the analysis are presented in Tables 7, 8, and 9.

1694 212

3.

Calculation of Direct Radiation Dose The direct radiation dose from piping and equipment containing fluids with the volumetric source terms derived in Section 2 was calculated by the following steps:

A.

For a range of typical pipe si:es (Table 10) an " Infinite Dose Factor" (IDF) was derived where Dose Rate (R/hr) = IDF/ Distance from Pipe (feet)

The IDF allow calculation of the dose' rate at any distance from an infinite pipe of given size and specified source term. The IDFs were calculated using the equation in Reference i page 360:

DOSE FROM AN INFINITE CYLINDRICAL PIPE g=l, F (e, b )

(Rockwell, Page 360)

O

)

2 07 = gamma flux at dose point, gammas /sec-cm Sy = volumetric source strength, gammas /sec-cm-)

Ro = radius of source, cm a = perpendicular distance from dose point to source, em

= self absorpt. ion distance, em (Rockwell, pages 361-363)

Q = half-angle subtended by source = 90*

b2 = bi + us bl = UwTw B = buildup factor Tw = pipe wall thickness, em Uw = macroscopic cross-section of pipe wall, cm-1 (Table, 4 )

Us = macroscopic cross-section of source, cm-1 (Table 4 )

F (e, b ) = function from Rockwell, pages 385-390 2

In this calculation, (a + :) was replaced by (a), which is conservative and is close to correct for distances such that a >= 10 Ro,which was also an assumption in the calculation of 2.

This substitution is what makes the concept of an IDF possible, since 9Iis then a linear function of a.

The dose is derived from the flux using the factors in Table 5 su=m ed over the S energy groups.

S I

@ /DCi R/hr D

=

i=1 1694 213

The IDF therefore equals aD, and is a function of pipe si::e, the time after the accident, and the volumetric source term chosen. The IDFs are given in Table 11.

B.

For each of the chosen points in the reactor building (Figures 1 through 4), the pipes within sight of that point containing radioactive fluids were identified. The piping was divided into linear segments and the length, distance and orientation of each segment with respect to the dose point were measured and recorded. Using the method of Reference 1 pages 111-115 a geometry factor was found for each segment which represents the ratio of the dose from that segment to the dose from an infinite pipe of which that segment is a portion. That is:

= D GF Dp y

Using the IDF:

IDF/A D

=

7 Where A is the perpendicular distance from the dose point to the infinite pipe.

The doses from all segments within sight of each dose point were added to generate the results shown on Table 12.

The containment spray heat exchange.rs and emergency condensers were modeled as large D pes with reduced vol'unetric source i

terms to account for the proper ratio of radioactive to non-radioactive water and for the actual radioactive inventory in each heat exchanger.

1694 214

4.

Airborne Radiation With the containment atmosphere source term given in Table 7, airborne radiation in the reactor building could be quite significant, depending on the amount of leakage from the primary containment.

Assuming a 1% per day leak rate, the technical specification limit, a reactor building free volume of 1.8 million cubic feet, and a standby gas treatment system flow rate of 2600 cubic feet per minute (one of two trains operating), reactor building air source terms can be calculated.

Under these assumptions, reactor building air concentrations of I-131 could exceed 1 mc/cc within one day of the accident, and would remain above 0.1 mc/cc even 30 days after the accident. The magnitude of this result is due to the high source term (25% of the core iodine in the primary containment atmosphere) and the large leak rate which is assumed to continue even after the accident when the primary contain-ment would presumably be depressuri:ed, and perhaps even sub-atmospheric.

Similar assumptions would also lead to very high noble gas concentrations in the reactor building. The calculated concentrations in mc/cc are as follows:

Time I-131 I-133 Xe-133 Xe-133m Xe-131m Kr-85 Kr-85m 10 min'

.0192

.0313

.148 5.1E-3 4.6E-4 7.84E-4

.0166 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />

.111

.176

.852

.0293, 2.66E-3 4.54E-3

.0842 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />

.651

.842 5.01

.165

.0159

.0273

.168 1 day 1.08

.869 8.26

.247

.0273

.0479

.0237 10 days

.567 7.95E-4 3.06

.0206

.0225

.0546 4.486E-17 30 days

.101 1.05E-10.221 4.98E-5 8.91E-3

.05<

0 If these concentrations were present, entry into the reactor building would not be possible, primarily due to the I-131.

5.

Control Room The control room at Oyster Creek is located in the turbine building at a floor elevation of 46'-6" separated from the reactor building by the office building. Figure 5 shows an E-W section through the control room and related structures. The source of potential direct exposure closest to the control room are the two Main Steam lines and the two Feedwater lines, which are located below and to the south of the control room.

i694 215

The Main Steam and Feedwater lines contain isolation valves intended to prevent the escape of post-accident radioactive fluids.

The 18 " feedwater lines each contain two check valves. The 24" Main Steam lines each contain two globe valves which close cutomatically on low reactor pressure, low-low reactor water level, high main steam flow, high main steam line tunnel temperature, and high main steam line radiation. No actions would be taken by the operators to allow the Main Steam and Feedwater systems to be contaminated by post-accident fluids, since allowing such fluids outside of the secondary containment could result in unacceptable offsite releases.

A calculation was done to determine the potential effect of contaminating the Main Steam and Feedwater lines on the habitability of the control room. Figure 6 shows the geometric relationship of the lines to the control room. The source term used for the steam lines was the source term of Table 7 multiplied by the ratio 306,000/S147.

The source term used for the Feedwater lines was the source term of Table 8 multiplied by 89,000/7000. This simulates a worst case where the entire inventory of the damaged core is retained in the reactor coolant system volume. The potential dose at the closest point in the control room from all four contaminated lines as a function of time following the accident is as follows:

Time Dose Rate, mr/hr 10 min.

49.4 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> IS.S 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 1.6 1 day

.07 10 days

.008 30 days

.003 This calculation contained a number of conservatisms in addition to the source terms such as assuming that the steam and feedwater lines parallel the entire south side of the control room, whereas the feed lines actually turn south away from the control room. Shielding by the foot thick control room floor was not considered.

Given the conservatisms in this calculation, it is clear that the Shin Steam and Feedwater lines present no threat to the habitability of the control.

1694 216

6.

Summary and Conclusions Oyster Creek is designed, as are other BWR plants, so that no access outside of the control room is required to mitigate the effects of major design basis events. The location of the control room, which incorporates considerable shielding and its own ventilation system, assures its accessibility following any postulated accident.

The potential post-accident dose rates due to direct radiation and the potentisi airborne radioactivity le els in the reactor building are so high that access to the reactor buitding cannot be assured follow-ing an accident resulting in releases equivalent to those required by NUREG-0578.

Because of this, various items of equipment located in the reactor building, such as motor control centers, instrument racks, and the sampling sink, may be inaccessible for a number of days following the accident. Of these, sampling is a function which is of such importance that it must be relocated outside the reactor building in order to assure its accessibility.

If any other vital functions are identified, appropriate actions will be taken to assure that the function will be maintained. Emergency procedures will address the possibility that the reactor building may be inaccessible following an accident.

e 1694 217

7 References 1.

Reactor Shielding Design Manual, T. Rockwell, (ed.), McGraw-Hill, 1956.

2.

Radiological Health Handbook, U.S. Department of Health, Education, and Welfare, Revised Edition January 1970.

3.

Radiation Source Term Information for NUREG-0578 Implementation, General Electric Company, November 1979 (Contains results of computer run identified as SNUMB=

70075, dated 11/9/79.)

e 1694 218

TABLE 1 ENERGY GROUPS MEV 1.

.001.1 2.

.10001.4 3.

40001.8 4.

.80001-1.0 5.

1.00001-1.5 6.

1.50001-2.0 7.

2.00001-3.0 8.

3.00001 6.0 TABLE 2 TDIE POINTS FOR CALCULATIONS (AFTER SHUTD0hN) 1.

0 2.

10 minutes 3.

I hour 4.

S hours 5.

I day 6.

10 days 7.

30 days 1694 219

Table 3 MASS ATTENUATION COEFFICIENTS, em /gm (RADIATION HEALTH HANDBOOK, PAGES 137-139)

Material HO Concrete ENERGY RANGES, MeV Fe 2

.001.1

.370

.171

.179

.1. 4

.094

.106

.0963

.4.S

.0669

.0786

.0709

.8-1.0

.0599

.0707

.0637 1.0-1.5

.0488

.057S

.0519 1.5-2.0

.0425

.0494

.0448 2.0-3.0

.0362

.0397

.0365 3.0 6.0

.0314

.0303

.0290 Table 4 MACROSCOPIC CROSS SECTIONS, Cm" Material ENERGY RANGES,Mev Fe HO Concrete 2

.001.1 2.886

,.171

.429

.1.4

.733

.106

.231 4.S

.522

.0786

.170

.8-1.0

.467

.0707

.153 1.0-1.5

.381

.0575

.124 1.5-2.0

.331

.0494

.107 2.0-3.0

.282

.0397

.0876

3. 0-6. 0

.245

.0303

.0696 Calculated based on densities of Fe 7.8 gn/cm HO 1

2 Concrete 2.4 1694 220

Table 5 FLUENCE RATE TO DOSE CONVERSION FACTOR GAmtA5/cm2-sec to give 1 R/hr Energy Ranges, Mev DC

.001.1 5.4 E6

.1.4 1.3 E6

.4.8 6.7ES

.8-1.0 5.5ES 1.0-1.5 4.0E5 1.5-2.0 3.2E5 2.0-3.0 2.4E5 3.0-5.0 1.75E5 1694 221

9 e

TABLE 6 ISOTOPES EXPLICITLY CONSIDERED IN ANALYSIS ISOTOPE ISOTOPE ISOTOPE ISOTOPE ISOTOPE ISOTOPE ISOTOPE ISOTOPE AG-109M CE-146 KR-87 ND-147 RB-88 SB-128M SR-91 1E-134 AG-111 CS-134 KR-88 ND-149 RB-89 SB-129 SR-92 XE-133 AG-111M CS-137 KR-89 ND-151 RB-90 SB-130 SR-93 XE-133M BA-137M CS-138 LA-140 PD-109 RH-193M SB-131 SR-94 XE-135 BA-139 CS-139 LA-141 PD-111 RH-104 SB-132 TC-101 XE-135M BA-140 CS-140 LA-142 PM-147 RH-194M SB-133 TC-102 XE-137 BA-141 EU-155 LA-143 PM-148 RH-105 SE-81 TC-104 XE-138 BA-142 I-128 MO-101 PM-149 RH-105M SE-83 TC-105 Y-90 BR-82 I-139 MO-102 PM-151 RH-196M SE-83M TC-99M Y-91 BR-82M I-131 MO-104 PM-152 RH-107 SE-84 TE-127 Y-91M BR-83 I-132 MO-99 PM-154 RH-108 SM-153 TE-129 Y-92 BR-84 I-133 NS-100 PR-142 RH-109 SM-155 TE-129M Y-93 BR-84M I-134 NB-95 PR-143 RU-103 SN-127 TE-131 Y-94 BR-85 I-135 NB-97 PR-144 RU-105 SN-128 TE-131M Y-95 CE-141 1-136 NB-97M PR-145 RU-106 SH-129 TE-132 Y-96 CE-143 KR-83M NB-98 PR-146 RU-107 SN-130 TE-133 2R-95 CE-144 KR-85 NB-98M PR-147 RU-108 SR-89 TE-133M' 2R-97 CE-145 KR-85M NB-99 PR-148 SB-127 SR-90 CN wO Jh N

N N

TABLE 7 VOLLAtETRIC SOURCE TERM - C0fffADetENT AB10SPilERE ENERGY GROUP - MEV Tiue 0.1

.1.4

.4.8

.8-1.0 1.0-1.5 1.5-2.0 2.0-3.0 3.0-6.0 0

4.6248E8 9.3631E8 1.3085E9 4.5536E8 3.0474E8 4.1976E8 4.4136E8 4.6855E6 10 min.

3.101E8 4.589E8 6.1829E8 2.6204E8 1.6156E8 1.67E8 2.0103E8 2.2676E6 I hour 2.846E8 1.4932E8 3.8638E8 1.5293E8 1.1409E8 5.8935E7 8.536E7 4.4543E5 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 1.8944E8 1.7409E8 1.2876E8 1.2122E7 4.5654E7 1.6949E7 1.0647E7 3.836El 1 day 1.6004E8 1.0726E8 5.75E7 1.5077E6 9.5713E6 3.0456E6 1.952E5 3.566E-8 10 days 5.2243E7 2.2544E7 2.7262E6 4.6282E4 5.9197E4 5.4481E3 1.1929E-18 0

30 days 3.9469E6 4.0097E6 4.024E5 6.4756E2 8.242E2 7.6528El 0

0 TABLE 8 VOLLEIETRIC SOURCE TEPJi - CONTAllefENT WATER (EXCLUDING NOBLE GASES)

ENERGY GROUP - MEV Time 0.1

.1.4

.4.8

.8-1.0 1.0-1.5 1.5-2.0 2.0-3.0 3.0-6.0 0

5.4288E7 9.5952E8 3.8338E9 2.38778E9 1.82994 E9 5.0292E8 1.8713E8 4.92E7 10 min.

2.448E7 5.9346E8 2.7032E9 1.5293E9 1.4369E9 3.8175E8 1.6675E8 3.5914E7 I hour 2.0686E7 4.6304E8 2.1693E9 8.8188E8 1.0337E9 3.4025E8 8.3421E7 4.3143E6 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 1.8249E7 4.0141E8 8.0694E8 6.84E7 3.1903E8 1.2794E8 2.4327E6 1.6183E4 1 day 1.6272E7 3.7303E8 4.245E8 1.5063E7 6.7366E7 3.4661E7 5.3893E5 1.1603El 10 days 6.9719E6 1.6636E8 6.4362E7 3.1078E6 6.0761ES 9.0457E6 3.5281E5 0

30 days 1.7581E6 3.3488E7 3.7496E7 9.4642E5 9.,9544E4 3.099E6 1.6568E5 0

Source term units are gammas /cm -sec.

C 4

N N

u

TABLE 9 V01.UMETRIC SollRCE TERM - CONTAINHEffl' WATER (INC111 DING NOBLE GASES)

ENERGY GR0 lip - HEV Time 0.1

.1.4

.4.8

.8-1.0 1.0-1.5 1.5-2.0 2.0-3,0 3.0-6.0 0

1.636E9 3.8E9 6.5431E9 2.7834E9 2.Oll6E9 1.7162E9 1.6259E9 4.92E7 10 min

1. 0847E9 1.932E9 3.6172E9 1.6957E9 1.4512E9 8.0187E8 8.2352E8 3.5914E7 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 9.9329E8 1.127E9 2.5171E9 9.9862E8 1.0337E9 4.2052E8 3.7342E8 4.3143E6 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 6.64E8 8.226E8 8.9354E8 8.4311E7 3.1903E8 1.3722E8 3.9021E7 1.6183E4 1 day 5.6127E9 5.7462E8 4.4605E8 1.5354E7 6.7366E7 3.4838E7 1.2097E6 1.le03El 10 days 1,8421E8 1.6688E8 6.4397E7 3.1078E6 6.0761ES 9.0457E6 3.5281ES 0

30 days 1.4912E7 3.3518E7 3.7532E7 9.4642E5 9.9544E4 3.099E6 1.6568E5 0

Source term units are gammas /cm -sec 4

N N

4

TABLE 10 REFERENCE PIPE CHARACTERISTICS Nominal Actual Actual Way Pipe Si:e (in.)

I.D.

Thickness (In) 1 (sch. 40)

1. 04 9

.133 2

2.067

.154 4

4.026

.237 6

6.065

.286 8

7.981

.322 10 10.020

.365 12 11.938

.406 14 13.126

.437 16 15.000

.500 (Marks' Standard Handbook for Mechanical Engineers, Eighth Edition, 1978, Pages 8-156 through 8-159) i694 225

TABLE 11 INFINITE DOSE FACTORS CONTAINMENT AIR NOMINAL PIPE SIZE, INO!!!S TIME I

2 4

6 8

10 12 14 16 0

289.45 1097.1 3948.2 8441.5 14245 21973 29292 35480 42357 10 min.

137.46 520.08 1874.7 3999.6 6747.3 10392 13856 16747 20006 I hour 73.57 275.87 995.66 2116.5 3559.5 5468.5 7289.3 8756.9 10481 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 22.404 83.491 291.09 616.89 1029.4 1575.5 2112.1 2507.4 2982.3 1 day 8.4182 31.09 105.17 221.9 3'63.74 555.32 741.04 874.19 1018.1 10 days

.99288 3.6844 11.42 23.567 38.041 56.832 76.498 89.495 99.17 30 days

.14945

.57131 1.7915 3.7414 6.1075 9.2146 12.48 14.57 16.278 CONTAINMliNT WATER ( EXCLUDING NOBLE GASES)

TIMli 1

2 4

6 8

10 12 14 16 0

662.35 2285.7 7296.7 14233 21400 29415 37023 41477 46574 10 min.

477.8 1653.4 5282.9 10295 15536 21376 26978 30252 34084 I hour 340.43 1175.9 3730.9 7273.7 10956 15038 18945 21261 23844 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 104.18 359.9 1121 2172.3 3265.3 4449.3 5583.5 6279.8 6953.9 1 day 43.3 149.5 457.27 879.16 1304.3 1744.5 2145.2 2420.4 2597.6 10 days 8.9015 30.835 91.116 172.59 254.03 330.18 404.25 457.56 474.56 30 days 3.2735 11.271 34.087 65.864 96.671 128.56 155.49 177.8 186.01 CONTAINMENT WATER (INCLUDING NOBLE GASES)

TIMil 1

2 4

6 8

10 12 14 16 O

1294.7 4556.7 14483 28372

.43261 59750 75055 86409 96999 C7N 10 min.

726.65 2553.2 817'2.5 15887 24218 33402 42053 4810' 54168

'I)

I hour 439.23 1534.7 4870 9479.3 14384

!.9740 24826 28182 31649 J""

8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 128.63 446.13 1377.6 2650.6 3987.1 5405.5 6750.7 7615.8 8396.8 1 day 51.142 175.97 528.61 1003.9 1483.5 1968.6 2406.6 2704.7 2879.4 rs) 10 days 9.4702 32.483 94.412 177.14 259.29 340.66 409.27 462.05 478.01

)f 30 days 3.3172 11.398 34.348 66.234 97.108 129.03 155.94 178.22 186.36

TABLE 12 REACTOR BUILDING POST ACCIDENT DOSE RATES DIRECT RADIATION, R/hr Time After Location Shutdown 23-A(1) 23-A(2) 23-B(1) 23-B(2) 23-C(1) 0 12595 24983 1886 3652 1593 10 min 9200 14121-1378 2079 1157 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 6450 8359 966 1238 815 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 1899 2267 284 337 242 1 day 726 805 109 120 96 10 days 137 138 21 21 19 30 days 53 53 8

8 7

23-C(2) 51-A(1) 51-A(2) 51-B(1) 51-B(2) 51-C(5) 51sC (6) 0 3201 1419 2955 3174 6433 55 1374 10 min 1795 1035 1645 2305 3597 39 772 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 1065 727 964 1623 2129 28 468 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 294 215 260 481 585 9

141 1 day 108 83 92 190 214 3.7 59 10 days 19 16 16 37 37 0.8 11 30 days 7

6 6

14 14 0.3 3.8 75-A(3) 75-A(4) 75-B(3) 75-B(4) 75-C (1) 0 738 21504 3559 82472 91 10 min 470 11704 2383 47871 66 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 309 6930 1611 29693 46 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 91 2011 477 8682 14 1 day 35 737 186 3264 5.5 10 days 6

107 34 524 1.1 30 days 2.2 33 13 176 0.4 75-C(2) 95-A(3) 95-A(4) 0 183 771 38749 10 min 103 396 18778 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 61 226 10108 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 17 66 2934 1 day 6

24 1041 10 days 1.1 3.2 117 30 days 0.4 0.9 24 1694 227

TABLE 12 (Cont.)

NOTES (1) Uses source term from Table 8 (no noble gases in containment water *).

(2) Uses source term from Table 9 (all noble gases in containment water).

(3) Uses source term from Table 8 for Emergency Condenser condensate lines and from Table 7 for E.C. steam lines. This simulates a situation where the ECs were used after the core spray had mixed the RCS and torus water.

(4) Uses source term from Table 8 multiplied by the ratio (89,000/7000) for the condensate lines, and Table 7 multiplied by (306,000/5147) for the steam lines. This simulates a worst case where the entire inventory of the damaged core is retained in the RCS volume.

(5) Uses source term from Table 8 (no noble gases in containment water).

(6) Uses source term from Table 9 multiplied by 89,000/7000 (most conservative case, assumes all radioactive material is retained in the RCS water).

• Containment water is the mixture of RCS and torus water.

1694 228

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ENCLOSURE 4 January 4, 1980 RESPONSE TO NUREG-0578 Section 2.1.8.a Improved Post Accident Sampling Capability

1.0 INTRODUCTION

The existing reactor coolant and primary containment atmosphere sampling systems and radiological and chemical analysis facil-itics at the Oyster Creek Nuclear Station have been reviewed for compliance with NUREG-0578 section 2.1.8.a and the NRC letter of clarification on the subject dated October 30, 1979. As a result of this review, system and plant modifications are required to ensure compliance with the NRC position as stated in NUREG-0578. These modifications which are detailed below are scheduled for implementation prior to January 1, 1981. 2.0 PROPOSED MODIFICATIONS Design basis accident calculations were performed using the source terms derived by our analysis of section 2.1.6.b of NUREG-0578. As a result of these calculations, the reactor building, in which the reactor coolant and drywell atmosphere sampling stations are presently located, was rendered inaccessible due to high radiation levels for at least several days following the accident. This requires that a new sampling station be pro-vided at a location outside the reactor buil'uing. The original counting room of the plant, which is now vacant, is presently being considered as the location for the new sampling system. The relation of the counting room relative to the reactor building and other plant structures is shown in Figures 1 In addition to the new sampling station, in-line gas analy;ers will be provided to measure the concentration of dissolved hy-drogen and oxygen in liquid samples. A sample preparation and dilution station will also be provided in the new sampling room. The extent to which remote handling devices are used at this station will depend in large part on the sample size. A 10 milliliter liquid sample size was selected for design review purposes. The sample size, however, is being evaluated to de-termine if it is in fact the minimum acceptable. The existing radiological spectrum analysis equipment ;t the plant is capable of analyzing the isotopes required by NUREG-0578. The existing equipment has a sensitivity of 100jyci/ml which can accommodate a properly diluted sample of reactor coolant and drywell atmosphere. The scope of work for the modification is shown conceptually in Figure 2. The new post-accident sampling station for reactor coolant and containment atmosphere will include the following design fea-tures. i694 235

1) The ability to draw pressurized and unpressurized reactor coolant samples. 2) The ability to purge the liquid and atmosphere sample lines to ensure a representative sample. 3) A means to contain the spread of high activity post-accident purge water and sample sink drainage. 4) Sample under negative containment atmosphere pressure. 5) Flushing of liquid sample lines with demineralized water and atmosphere sample 31nes with an appropriate supply of clean air. 6) Shielding of the sample station, sample lines and sample containers to ensure that the exposure to personnel is within 5 Rem whole body and 18 3/4 Rem extremity doses specified in NUREG-0578. The sample sink hood will be maintained at a negative pressure to prevent airborne contamination of the area. The exhaust air from the station will be monitored and treated, if neces-sary, before release. All piping modifications downstream of the second containment isolation valve will be non-seismic and designed to Quality Group D. 3.0 CLARIFICATIONS AND QUALIFICATIONS oron concentrations will not be analyzed since BWR's do not c rely on dissolved boron for shutdown. Chloride concentrations will not be analyzed since no undemineralized water penetrates the primary containment. However, a method of measuring conduc-tivity will be provided. Source terms for shielding, dilution and sample size are based on t he assumption that all the noble gases resulting from the accilent have escaped from the liquid to the containment atmos-phere. 1694 236

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ENCLOSURE 5 January 4, 1980 INTERIM METHODS FOR ESTIMATING RADI0 ACTIVE RELEASES INTRODUCTION NUREG 0578, recommendation 2.1.8b, requires that methods be developed to estimate offsite release rates during an accident in the event that installed instrumenta-tion goes off-scale. This repc-t sumarizes the methods which will be employed until new wide-range instrumentation is installed during the latter part of 1980. BACKGROUND During normal operations, tne exhaust from the reactor building, the old radwaste building, and the majority of the turbine building is routed to the ventilation stack. The flow rates for each of these buildings are 56,500 cfm,14,500 cfm, and 91,000 cfm, respectively. During the summer, the turbine building's four roof ventilators are turned on which exhaust 36,000 cfm and reduce the stack flow rate by an equal amount. The turbine building also has two local ventilation systems for the feedpump room (68,000 cfm), and the turbine lube oil / reheater protection areas (10,000 cfm) which exhaust to the environment. The new radwaste building has its own local ventilation system (34,000 cfm) which also exhausts directly to the environment. During an accident, the reactor building's supply and exhaust fans are shut down; and the building is exhausted with the Standby Gas Treatment System wl...h discharges to the ventilation stack at a flow rate of 270G cfm. . Additionally, the local turbine building exhaust fans and roof ventilators could be shutdown if there were indications that significant amounts of radioactivity were being released. All other ventilation alignments remain unchanged. At present, the ventilation stack and the exhaust from the new radwaste building have installed instrumentation for monitoring offsite releases. INTERIM METHODS The following is a summary of the methods which would be used to estimate the release rates of noble gas and iodine /particulates in plant effluents: A. Stack Monitorino Noble gas activity is the only parameter that is monitored, recorded, and has alarm set points at the present time. In the event of an accident in which the instrumentation goes off scale, sample taps are provided for obtaining grab samples of the diluted stack effluent at the stack monitoring system. Under normal operating conditions, an 1800 cc marinelli beaker is used to obtain a sample at that location. It has a sufficient count rate to be counted using normal techniques on the multichannel analyzer. The stack monitor at normal operating conditions reads approximately 103 cpm. 1694 239 1

The stack monitor readout maximuu is 106 cpm. At that level a 1.8 cc sample taken at the stack monitor locat on would give a comparable count rate. Sampling techniques presently uset are capable of obtaining small gas samples in the 1 cc range. Sample dilution techniques are also used at the plant. These are capable of further diluting samples which are too hot to count. By the use of sample dilution and other available counting geometrics, accurate estimation of noble gas effluent release rates can be obtained at order of magnitude levels over the maxinum monitor range. 1. Noble Gas Effluents (a) Tracor Northern TN-ll computer based Geli multichannel analyzer system for gama isotopic analysis. The size of grab samples must be adjusted to fit within the count rate limits of the instrument. Calibration over the spectral range of the instrument is performed daily using NBS mixed isotope calibration sources. (b) The stack monitoring / sampling location is in the package heating boiler house adjacent to the base of the stack. Sample is drawn through an isokinetic probe in the stack designed to give a representative sample. The sample is removed to a count room located remotely from the area. Counting instrument backgrounds are routinely taken which correct for external radiation in the area. (c) If 10 cpm is not exceeded, a control room readout is used. If 10 cpm is exceeded, the grab sample method described above will be used. (d) The capability of obtaining release rate estimates every fifteen minutes exists. (e) The installed stack gas monitoring equipment receives power from the plant's vital instrument bus which has two sources of power. The off-line counting equipment is supplied with normal convenience outlets. 2. Radio-iodine and Particulate Effluents (a) The instrument to be used for sample analysis is a Tracor Northern TN-ll computer based Geli multichannel analyzer system. Normal count room techniques are used for calibra-tion, background correction, and sample analysis. The count room location is remote enough to preclude high background levels. (b) The monitoring / sampling location is in the package heating boiler adjacent to the base of the stack. The sample is drawn through an isokinetic probe in the stack. 1694 240 2

(c) The sampling system contains one particulate filter and one charcoal cartridge. These are removed by unscrewing the sample holder and placing the fi'ter and cartridoe into a shielded (if necessary) container. The sample collection period is adjusted to give a concentration on the filt:r and cartridge in the range to be counted on the TN-ll system. It would, therefore, not be anticipated to have excessive radiation levels from the samples. (d) Prior to counting,the charcoal cartridge would be flushed with air in a radioactive fume hood to remove entrained noble gases. Buildup of noble gas decay products from iodine isotopes on the charcoal would not interfere with the analysis. (e) Sample collection equipment

• is powered from the plant's vital instrument bus. The off-line counting equipment is supplied by normal convenience outlets.

B. New Radwaste Ventilation Monitoring The ventilation duct monitor alarm points are set at a level to preclude exceeding 10CFR20 limits at the site boundaries. When the alarm setpoints are reached, the ventilation system is isolated according to plant procedures. In the event the system was not isolated and the monitor range was exceeded, samples for noble gas, parciculates, and radio-iodines would be obtained and analyzed in a manner similar to that described in the stack effluent analysis. C. Miscellaneous Turbine Building Exhausts During an accident, the turbine building would be isolated from the reactor coolant system since there are no reactor plant auxiliary systems located in the turbine building. Therefore, it is unlikely that the effluents would contain any more radioactivity than normally. Furthermore, the local ventila-tion systems could be shut down so that the entire building is exhausted to the ventilation stack. However, if the local exhaust fans were running, it is possible to draw samples of the exhaust air and analyze it as described in "A", above. PROCEDURES Procedures for conducting all aspects of measurement analysis will be prepared or revised, as appropriate, to include the following considerations: A. Minimizing radiation exposure. B. Calculational methods. C. Dissemination of results. All new and existing equipment are included in the plant's equipment calibration program. Calibration techniques are as specified by the manufacturer. 3 1694 24I

ENCLOSURE 6A January 4, 1980 INTERIM PLAN FOR THE ONSITE TECHNICAL SUPPORT CENTER The interim plan for meeting the requirements of the Onsite Technical Support Center (TSC) in confomance with the requirements of NRC NUREG-0578, as further clarified by the ec letter of October 30, 1979, from the Director of the Office of Nuclear Reactor Regulation is presented herein. The plan presented herein responds to the reauirements of Items lA through 1F, Section 2.2.2.b of Enclosure 1 of the NRC letter of October 30,1979, cited above. A. NRC Position / Clarification Establish a Technical Support Center P';) and provide a complete description. JCP&L Resoonse The TSC will be located in the Conference Room of the Auxilicry Office Building. This area, along with the surrounding offices within the , Auxiliary Office Building, will provide sufficient space for all personnel necessary to cerfom the accident assessment function. This area also meets the requirement of being located within the plant security boundary. B. NRC Position / Clarification Provide plans and procedures for engineering /manacement support and staffing of the TSC. JCPAL Response The temocrary TSC will be approximately 500 square feet, (Conference Room only) and can, therefore, accomodate 10-15 people. Five of these people will be NRC personnel. The remaining staff of the TSC is defined in the Emergency Plan. C. NRC Position / Clarification Install dedicated communication between the TSC and the Control Room, near site emergency operations center, and the NRC. 1694 242

ENCLOSURE CA Page 2 of 2 JCP&L Response Telephones will be installed in the TSC to allow communication to the Control Room, the nearsite erergency operation center and the NPC; as well as, a telephone will be installed in the Control Room for use in communicating to the TSC. D. NRC Position / Clarification Provide monitoring (either portable or oemanent) for both direct radiation and airborne radioactive contaninants. The monitors shall provide warning when the radiation levels in the support center are reaching potentially dangerous levels. The licensee should designate action levels to define when protective measures should be taken (such as using brea*,hing apparatus and potassium fodide tablets, or evacuation to the Control Room). JCP&L Response Portable monitors will be provided upon activation of the TSC. These monitors will detect both direct and airborne radiation and will be temporarily provided by the plant Health Physics Department. The plant Health Physics Supervisor will also detemine the action levels to determine when and what protective measures will be taken. The maximum action to be taken will be evacuation from the TSC to the Control Room. E. NRC Position / Clarification Assimilate or ensure access to Technical Data, including the licensee's best effort to have direct display of plant parameters, necessary for assessment in the TSC. JCP&L Resoonse The temporary TSC will be located in the vicinity of the Document Control Center which contains all the technical literature for the plant (such ass drawings, technical manuals, procedures, etc). The mechanism which will be utilized to obtain the plant parameters will be via the dedicated telephone system between the Control Room and the TSC. F. NRC Position / Clarification Develop procedures for nerfaming this accident assessment from th'e Control Room should the TSC become uninhabitable. JCP&L Response The Emergency Plan addresses staffing of the Control Room should the TSC become uninhabitable. 1694 243

ENCLOSURE 6B January 4, 1980 LONG RANGE PLAN FOR PERMANENT ONSITE TECHNICAL SUPPORT CENTER The plan for meeting all the requirements of the onsite Technical Support Center (TSC) in conformance with the requirements of NRC NUREG 0578, as further clarified by the NRC letter of October 30, 1979 from the Director of the Office of Nuclear Reactor Regulation, is presented herein. The plan presented for the TSC takes no exception in intent to that spe-cified by NUREG 0578, Section 2.2.2.b. Subsequent engineering effort will define in detail design criteria, precurement specifications, facility modifications and other efforts required to complete implementation of this plan by June 1, 1981. The completion date is based on making the necessary interface tie-ins to the plant during its 1981 refueling outage. The plan presented herein responds to the requirement of Item 1G, Sectton 2.2.2.t* of Enclosure 1 of the NRC letter of October 30, 1979 cited above. The fonnat of the plan presented is structured to respond to the format of Items 2-10 of Section 2.2.2.b of Enclosure 1 to that letter where the NRC position and clarifications thereto are presented. LOCATION (Section 2.2.2.b, Item 2) NRC POSITION / CLARIFICATION It is reconinended that the TSC be located in close proximity to the control room to ease communications and access to technical information during an emergency. The center should be located onsite, i.e., within the plant security boundary. The greater the distance from the CR. the more 1694 244

ENCLOSURE 6B Page 2 of 11 sophisticated and complete should be the corrnunications and availability of technical information. Consideration should be given to providing key TSC personnel with a means for gaining access to the Control Room.

RESPONSE

Two candidate locations for the TSC for the Oyster Creek Nuclear Generating Station are presently under review and assessment. The preferred candidate TSC would be located in the Turbine Building. The alternate location under consideration is in the Office Building which is adjacent to both the Turbine and Reactor Buildings (Attachment 1: JCP&L Dwg.19508, Building Complex Plan). Both of these candidate locations are within the plant security boundarv anti in existing structures. If upon detailed review, neither of these areas are found suitable, another area within plant boundary will be selected for the TSC function. The preferred TSC location at the Turbine Building is at Elevation 63'-9" and directly above the existing Control Room. This area is shown on Drawing 4511; Third Floor Plan, Office Building (Attachment 2), and is presently designated as the. Mechanical Equipment Room No.1. Large areas within this location are unused and suitable for adaptation to TSC personnel and equipment requirements. Access to the Control Room is provided by an existing stairway joining these two areas. This candidate location represents the closest available proximity to the Control Room that could be totally dedicated to the TSC function. The alternate candidate location is the allocation of the currently occupied third floor area of the Office Building. This would be bounded 3 C, ( 4,0 ), (0,0 ) and, in addition, by Drawing 4511 coorriinates (0 0 ). (0 0 C 4 A 3A the present Eating and lieeting Room adjacent to it. This location is at the same elevation as the Control Room and in proximity to it (approximately 100'). Access to the Control Room is by an existing connecting corridor. 1694 245

ENCLOSURE 6B Page 3 of 11 PHYSICAL SIZE & STAFFI'!G (Sect'on 2.2.2.b, Item 3) NRC POSITION / CLARIFICATION The TSC should be larce enouch to house 25 persons, necessary engineering data and infomation displays (TV monitors, recorders, etc.). Each licensee should specify staffing levels and disciplines reporting to the TSC for emergencies of varying severity.

RESPONSE

The TSC location shall provide a work area for a staff of twenty-five (25) people, engineering data, infomation displays, communication and computer data acquisition facilities and radiation monitoring equipment. The preferred TSC location in the Turbine Building has an approximate useable floor area of 1500 square feet. Adequacy of this area size will be detemined upon subseouent soace allocation requirements of necessary equipment in addition tc pmvision for the personnel working area. This candidate location will be completely dedicated to the TSC function both during normal plant operation as well as emercency conditions so that efficient utilization is possible of available space. The alternate candtdate TSC area in the Office Buildino shall consist of two categories a dedicated core TSC area, and a multiple use TSC area. The dedicated core area shall be reserved for TSC function both during emergency conditions ano during normal plant operations. This area shall contain the engineering data, information displays, comunication and computer data acquisition facilities, and radiation monitoring equipment. This shall be located in what is now the Third Floor Eating and Meetino Room and which will be 1694 246

ENCLOSURE 6B Page 4 of 11 available for use upon scheduled completion of the new Maintenance Building in 1981. Available floor space is approximately 400 square feet. The alternate TSC location area that will be for multiple use is at third floor Office Coordinates (0,0c), (0,0 )., (0, O ), (0, Oa)- 1 1 a 4 c 4 This area shall be for exclusive use of TSC working personnel during a declared emergency. At other times it will be occupied by presently assigned plant personnel. It shall at all times contain desks, chairs, tables, and phones suitable for use by TSC personnel. Available gross office floor space is approximately 2,500 square feet;however,this is partitioned and furnished for normal plant operating personnel use. The staff levels and disciplines reporting to the TSC during emergencies of varying severity shall be in accordance with plant procedures. ACTIVATION (Section 2.2.2.b, Item 4) NRC POSITION / CLARIFICATION The center should be activated in accordance with the " Alert"tievel as defined in the NRC document " Draft Emergency Action Level Guidelines, NUREG-0610" dated September,1979, and currently out for public coment. Instrumentation in the TSC should be capable of providing displays of vital plant parameters from the time the accident began (t = 0 defined as either reactor or turbine trip). The Shift Technical Advisor should be consulted on the " Notification of Unusual Event" however, the activation of the TSC is discretionary for that class of event.

RESPONSE

The TSC shall be activated in accordance with clant precedures in con-formance with NUREG-0610 quidelines and as noted above. s Selected instrumentation records, computer printouts, and other displays 1694 747

ENCLOSURE 6B Page 5 of 11 of vital plant parameters shall be automatically initiated upon either a reactor or turbine trip. This shall include, but not necessarily be limited to; strip charts of hard wired instrumentation, listing or interrogation of plant parameter levels accessed by the existing com-puter, and a sequence event recorder. This plant parameter data shall be either continuously recorded, or at time intervals adequIte to provide trending as well as status information. INSTRUMENTATION (Section 2.2.2.b, Item 5) NRC POSITION / CLARIFICATION The instrumentation to be located in the TSC need not meet safety-grade requirements,but should be qualitatively comparable (as regards accuracy and reliability) to that in the control room. The TSC should have the capability to access and display plant parameters independent from actions in the control room. Careful consideration should be given to the design of the interface of the TSC instrumentation to assure that addition of the TSC will not result in any degradation of the control room or other plant functions.

RESPONSE

Instrumentation that is located in the TSC shall be qualitatively comparable to that in the control room. The majority of this instrumentation will con-sist of acquisition of analog or digital data records from that available in the Control Room. Instrumentation interfaces,shall be designed to assure no degradation of the control room or other plant functions. Selected analog data shall be recorded in a continuous manner on charts upon activation of a reactor or turbine trip. Digital data will be recorded by a line printer in the TSC with capability to address selected parameters, as well as duplicating the complete plant historical and 1694 248

ENCLOSURE 6B Page 6 of 11 current status records available to the control room. Digital data plotting capability shall be provided to trend selected plant parameters. A preliminary schematic of data instrumentation is shown in Attachment 3. INSTRUMENTATION POWER SUPPLY (Section 2.2.2.b Item 6) NRC POSITION / CLARIFICATION The power supply to the TSC instrumentation need not meet safety-grade requirements, but should be reliable and of a quality compatible with the TSC_ instrumentation requirements. To insure cootinuity.ot.infonnation at the TSC, the power supply provided should be continuous or:e the TSC is activated. Consideration should be given to avoid loss of stored data (e.g., plant computer) due to momentary loss of power or switching tran-sients. If the power supply is provided from a plant safety-related power source, careful attention should be given to assure that the capa-bility and reliability of the safety-related power source is not degraded as a resul.t of this modification.

RESPONSE

The TSC instrumentation power supply shall be normally provided by the off-site plant power source. In addition an independent, regulated power supply, shall be provided and activated upon a reactor or turbine trip. TECHNICAL DATA (Section 2.2.2.b, Item 7) NRC POSITIOh/ CLARIFICATION Each licensee should establish the technical data requirements for the TSC, keeping in mind tha accident assessment function that has been established for those persons reporting to the TSC during an emergency. As a minimum, data (historical in addition to current status) should be available to permit the assessment of: 1694 249

ENCLOSURE 6B Page 7 of 11 Plant Safety Systems Parameters for: - Reactor Coolant System - Secondary System (PHPs) - ECCS Systems - Feedwater & Makeup Systems - Containment In-Plant Radiolooical Parameters for: - Reactor Coolant System - Containment - Effluent Treatment - Release Paths Offsite Radiological - Meteorology - Offsite Radiation Levels

RESPONSE

Technical data required for accident assessment is presently under review. Table 7-1 is a listing of reactor parameters which are input to the Xerox and Prime computers. These will be made available at the TSC. Listed therein are the signals, quantity of data points, signal range, scan frequency, units and panel location (in main Control Roon). In addition, the main Control Room annunciator panel parameters (750 points) will be available to the TSC. These will be selectively displayed, via the seouence-event recorder. A continuous strip chart recorder will be installed in the TSC tn monitor reactor water level. 1694 250 ~

ENCLOSURE 6B Page 8 of 11 Records that pertain to the as-built conditions and layouts of structures, systems, and components shall be stored and filed at the TSC. Selected baseline documentation, such as General Arrangements of Piping and Instrument Drawings, shall be available in full size and critical ones displayed on walls in the TSC area. Other records shall be available at the TSC in microfilm form with capability for reproduction or viewing there. Subsidiary documentation shall be accessible from the onsite Document Control Center. DATA TRANSMISSION (Section 2.2.2.b. Item 8) NRC POSITION / CLARIFICATION In addition to providing a data transmission link between the TSC and the Control Room, each licensee should review current technoloay as regards transmission of those paraneters identified for TSC display. Although there is not a requirement at the present time, each licensee should. investigate the capa,bility to transmit plant data offsite to the Emergency Operations Center, the NRC, the reactor vendor, etc.

RESPONSE

Essential data will be transmitted by direct wiring between the Control Room area and the TSC. This type of transmission applies both to analog and digital data formats. The feasibility of digitizing analog data after receipt at the TSC will be investigated. STRUCTURAL INTEGRITY (Section 2.2.2.b, Item 9) NRC POSITION / CLARIFICATION A. The TSC need not be designed to seismic Category I requirements. The center should be well built in accordance with sound engineering 1694 251

ENCLOSURE 6B Page 9 of 11 practice with due consideration to the effects of natural phenomena that may occur at the site. B. Since the center need not be designed to the same stringent requirements as the Control Room, each licensee should prepare a backup plan for responding to an emergency from the ccntrol room.

RESPONSE

A. The two candidate locations for the TSC occupy areas of currently existing buildings. The preferred candidate TSC area directy above the control room meets the structural integrity requirements of the Turbine Building. The alternate TSC location is in the Office Building. This has been built in conformance with local and state code require-ments for a structure of this type. B. The TSC function is based on acquisition of essential data transmitted from the control r,oom. In the event that this TSC function is to be performed from the control room area itself, all essential data is available there. Offsite comunication will utilize the control room facilities. A to be detennined limited work area will be assigned for designated TSC personnel, such as the existing kitchenette or shift supervisors office, in accordance with emergency procedures. HABITABILITY (Section 2.2.2.b, Item 10) NRC POSITION / CLARIFICATION The license should provide protection for the technical support center personnel from radiological hazards including direct radiation and air-borne contaminants as per General Design Criterion 19 and SRP 6.4. A. License should assure that personnel inside the technical support center (TSC) will not receive doses ir. excess of those specified in GDC 19 and SRP 6.4 (i.e., 5 Rem whole body and 30 Rem to the thyroid for the duration of the accident). Major sources of radiation should be considered.

ENCLOSURE 6B h _e 10 of 11 8. Permanent monitoring systems should be provided to continuously indicste radiation dose rates and airborne radioactivity concentra-tions inside the TSC. The monitoring systems should include local alarms to warn personnel of adverse conditions. Procedures must be provided which will specify appropriate protective actions to be taken in the event that high dose rates or airborne radioactive concentrations exist. C. Pemanent ventilation systems which include particulate and charcoal filters should be provided. The ventilation systems need not be qualified as ESF systems. The design and testing guidance of Reg-ulatory Guide 1.52 should be followed except that the systems do not have to be redundant, seismic, instrumented in the control room or automatically activated. In addition, the HEPA filters need not be tested as specified in Regulatory Guide 1.52 and the HEPA's do not have to meet the QA requirements of Appendix B to 10 CFR 50.

However, spare parts should be readily available and procedures in place for replacing failed components during an accident. The systems should be designed to operate from the emergency power supply.

D. Dose reduction measures such as breathing apparatus and potassium iodide tablets can not be used as a design basis for the TSC in lieu of ventilation systems with chan:oal filters. However, potassium iodide and breathing apparatus should be available.

RESPONSE

A. The candidate TSC locations selected minimize potential of radiological hazards. Locations utilize both distance and existing shielding ~ provided by the Reactor Building external walls. The preferred can-didate location directly above the control room provides the same degree of radiological protection as the control room itself to direct radiation. jfg4 }}}

ENCLOSURE 6B Page 11 of 11 B. A. permanent monitoring system shall be provided to continuously indicate radiation dose rates and airborne concentrations inside the TSC. This will include local alarms to warn of adverse conditions. The system shall be activated upon occupancy of the TSC. A plant procedure shall be provided to specify appropriate actions in the event of dose rates or airborne radioactive concentrations exceeding specified limits. C. The TSC shall be provided with a ventilation system to enable the center to be habitable to the same degree as the control room for post-ulated accident conditions. Equipment spare parts lists and instruction manuals will be reviewed to determine those repairs that can be effected in a safe and ticely manner durina plant energency conditions. D. Potassium iodide and breathing apparatus shall be additionally pro-vided,and located at the TSC. 1694 254

TABLE 7.1

SUMMARY

OF INPUT SIGNALS TO PLANT COMPUTERS SIGNAL SCAN ENG PANEL SIGNAL QNTY RANGE FREQ UNITS LOCATION High Level (136) LPRM 124 0-1VOC 60 Sec 0-125 Watts /CH2 3R & SR APRM 8 0-IVDC 1 Sec 0-150% Power 3R & SR Tip Level 4 0-IVDC 1 Sec 0-125 Watts / Cit 2 4R Low Level (18) Feedwater Flow 1 200-1000MV 1 Sec 0-8x106 lbs/hr 9R Feedwater Temp 1 200-1000MV 1 Sec 0-400*F 10R Reactor Pressure 1 200-1000MV 1 Sec 950-1050 psig SF Reactor Pressure 1 200-1000MV 1 Sec 0-1600 psig 9R Reactor Core 1 200-1000MV 1 Sec 0-30 psid if Pressure Drop Recirculation Flow 5 200-1000MV 1 Sec (0-20)x104 SR gpm Recirculation Inlet Temp 5 200-1000MV 1 Sec 0-600*F 10R Reactor Water Level 2 200-1000MV 1 Sec 0-8 Feet 9R Reactor Steam Flow 1 200-1000MV 1 Sec 0-8x106 lbs/hr 9R m W 4 N LD LT1

TABLE 7.1

SUMMARY

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9 ENCLOSURE 6B ATTACH!1ENT 3 SEQUEllCE OF ANNUNCIATOR r- - EVENTS RECORDER PANELS I I ~ l l I I l l PRINTER XEROX-PRIME I PLOTTER COMPUTER 1 I I l 1 i i i i i KEYBOARD-I/O L AN0/0R DEDICATED TELEPHONE - TECHNICAL SUPPORT CENTER REACTOR WATER LEVEL CONTINUOUS RECORDER 1694 261

~~ ~ ENCLOSURE 7 January 4, 1980 RCS VENTING JCP&L's position with respect to this item is as outlined previously" in our submittal dated October 19,1979. That submittal made commitments to upgrade the venting arrangement for the isolation condensor steam line high points and exhaust the gases (during a casualty only) to the con-tainment instead of the main steam system. The attached sketch illustrates the conceptual design for modifying this system. The four solenoid valves will be remotely operable from the control room and shall be safety grade class IE. The lines will discharge to the torus suppression chamber via the existing clean-up systen relief valve discharge line. 1694 262

4'iNE1 3WNE3 V/4 5 V/4-20

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