ML18082A670

Forwards Addl Info Re Full Power License Requirements Identified at 800619 Meeting.Info Supplements

ML18082A670
Person / Time
Site:	Salem
Issue date:	07/01/1980
From:	Mittl R Public Service Enterprise Group
To:	Schwencer A Office of Nuclear Reactor Regulation
References
NUDOCS 8007030229
Download: ML18082A670 (65)

Text

• -

0 PS~G f Public Service Electric and Gas Company 80 Park Place Newark, N.J. 07101 Phone 201/430-7000 July 1, 1980 Director of Nuclear Reactor Regulation

u. s. Nuclear Regulatory Commission Washington, D.C.

20555 Attention:

Mr. A. Schwencer, Acting Chief Licensing Branch No. 3 Division of Licensing Gentlemen:

FULL POWER LICENSE REQUIREMENTS NO. 2 UNIT SALEM NUCLEAR GENERATING STATION DOCKET NO. 50-311 PSE&G hereby submits, in the enclosure to this letter, addi-tional information related to the full power license re-quirements identified at a meeting with members of the NRC staff on June 19, 1980.

This information supplements our letter dated June 27, 1980.

Should you have any questions in this regard, please do not hesitate to contact us.

Encl.

The Energy People B0070302Zl}

r-:r

~yf Jil-rs, R. L. Mittl General Manager -

Licensing and Environment Engineering and Construction 95-0942

FULL POWER REQUIREMENTS NO. 2 UNIT SALEM NUCLEAR GENERATING STATION TMI-2 RELATED ITEMS.

II.B.2 PLANT SHIELDING Provide (1) a radiation and shielding design review that identifies the location of vital areas and equipment in which personnel occupancy may be unduly limited or safety equipment may be unduly degraded by radiation during oper-ations following an accident resulting in a degraded core, and (2) a description of the types of corrective actions needed to assure adequate access to vital areas and protec-t ion of safety equipment.

(See NUREG-05 7 8, Section 2.1. 6b, and letters of September 27 and November 9, 1979.)

Response

This information was provided in PSE&G letter, R. L. Mittl too.. D. Parr, dated January 4, 1980.

Additional informa-tion is provided in the revised pages to be inserted in the enclosure to the January 4, 1980 letter (Attachment 1).

II.B.3 POSTACCIDENT SAMPLING Provide (1) a design and operational review of the capability to promptly obtain and perform radioisotopic and chemical analyses of reactoi coolant and containment atmos-phere samples under degraded core accident conditions with-out excessive exposure, (2) a description of the types of corrective actions needed to provide this capability, and (3) procedures for obtaining and analyzing these samples with the existing,equipment.

(See NUREG-0578, Section 2.l.8a and letter of September 27 and November 9, 1979).

Response

This information was provided in PSE&G letter, R. L. Mittl to 0. D. Parr, dated January 4, 1980.

Additional informa-M P80 64 09/l

/

/

' tion is provided in the revised pages to be inserted in the January 4, 1980 letter (Attachment 1).

II.E.4.2 CONTAINMENT ISOLATION DEPENDABILITY Provide (1) containment isolation on diverse signals, such as containment pressure or ECCS actuation, (2) automatic isolation of nonessential systems (including the bases for specifying the nonessential systems), (3) no automatic re-opening of containment isolation valves when the isolation signal is reset.

(See NUREG-0578, Section 2.1.4, and let-ters of September 27 and November 9, 1979.)

Response

This information was provided in PSE&G letter, R. L. Mittl to O. D. Parr, dated January 4, 1980.

Additional informa-tion is provided in the revised pages to be inserted in the enclosure to the January 4, 1980 letter (Attachment 1).

III.A.1.1 - UPGRADE EMERGENCY PREPAREDNESS Provide an emergency response plan in substantial compliance with NUREG-0654, "Criteria for Preparation and Evaluation of Radiological Emergency *Response Plans and Preparedness in Support of Nuclear Power Plants" (which may be modified after May 13, 1980 based on public comments) except that only a description of and completion schedule for the means for providing prompt notification to the population (App.

3), the staffing for emergencies in addition to that already required (Table B.l), and an upgraded meteorological program

(]\\pp. 2) need be provided.

NRC will give substantial weight to FEMA findings on offsite plans in judging the adequacy against NUREG-0654.

Perform an emergehcy response exercise to test the integrated capability and a major portion of the basic elements existing within emergency preparedness plans and Organizations.

M P80 64 09/2

• Response Additional information will be provid~d subsequent to a forthcoming meeting with the NRC staff.

III.D.1.1 PRIMARY COOLANT SOURCES OUTSIDE CONTAINMENT Reduce leakage fr9m systems outside containment that would or could contain highly radioactive fluids during a serious transient or accident to as-low-as-practical levels, measure actual leak rates and establish a program to maintain leak-age at as-low-as-practical levels and monitor leak rates.

(See NUREG-0578, Section 2.l.6a, and letters of September 27 and November 9, 1979.)

Response

This information was provided in PSE&G letter, R. L. Mittl to o. D. Parr 1 dated January 4, 1980.

Additional informa-tion is provided in revised pages to be inserted in the enclosure to the January 4, 1980 letter (Attachment 1).

III.D.3.4 CONTROL ROOM HABITABILITY Identify and evaluate potential hazards in the vicinity of the site as described in SRP Sections 2.2.1, 2.2.2, and 2.2.3, confirm that operators in the control room are ade-quately protected from these hazards* and the release of radioactive gases as described in SRP Section 6.4, and, if necessary, provide the schedule for modifications to achieve compliance with SRP Section 6.4.

Response

This information is provided in Attachment 2.

M P80 64 09/3

• . FULL POWER PLANT SPECIFIC ITEMS (NON-TMI)

1.

IE Bulletin 79-07.

Response

It is expected that a letter will be transmitted by July 7, 1980, summarizing total number of supports evaluated, results of the evaluation and number of modifications made.

M P80 64 09/9

Containment Isolation Provisions for PWRs and BWRs (Section 2.1. 4 NRC Position

1.

All containment isolation system designs shall comply with the recommendations of SRP 6.2.4; i.e., that there be diversity in the parameters sensed for the initiation of containment isolation.

2.

All plants shall give careful reconsideration to the definition of essential and non-essential systems, shall identify each system determined to be essential, shall identify each system determined to be non-essential, shall describe the basis for selection of each essential system, shall modify their containment isolation designs accordingly, and shall report the results of the re-evaluation to the NRC.

3.

All non-essential systems shall be-automatically iso-lated by the containment isolation signal.*

4.

The design of control systems for automatic containment i.solation valves shall be such that resetting the isola-tion signal will not result in the automatic. reopening of containment isolation valves.

Reopening of contain-ment isolation valves shall require deliberate operator action.

Response

1.

The containment isolation system complies with the requirements for isolation initiation by diverse para-meters as described in Section 5.4 of the FSAR.

A num-ber of isolation ~ignals are provided for valv~ closure.

Each signal is indicative of certain operating condi-tions and is generated by diverse input parameters.

The isolation signals arid their input parameters are as follows:

Containment Isolation -

Phase A

a.

Manual* Actuation M P79 54 01/16 Salem 1 & 2 JAN 1 1980

b.

High Containment Pressure

c.

Low Pressurizer Pressure

d.

High Differential Pressure Between St~am Lines

e.

High Steam Line Flow Coincident with Low Steam Line Pressure or Low-Low Tavg.

Containment Isolation -

Phase B

a.

Manual Actuation

b.

High-High Containment Pressure Containment Ventilation Isolation

a.

Manual Actuation

b.

High Containment Pressure

c.

Low Pressurizer Pressure

d.

High Differential Pressure Between Steam Lines

e.

High Steam Line Flow Coincident with Low Steam Line Pressure or Low-Low Tavg.

f.

High Containment Radiation - Particulate

g.

High Containment Radiation -.Iodine

h.

High Containment Radiation -

Gaseous Main Steam Line Isolation

a.

Manual Actuation

b.

High-High Containment Pressure

c.

High Steam Line Flow Coincident with Low Steam Line pressure or Low-Low Tavg.

M P79 54 01/17 Salem 1 & 2 JAN 1 1980

Feedwater Isolation

a.

Manual Actuation

b.

High Containment Pressure

c.

Low Pressurizer Pressure

d.

High Differential Pressure Between Steam Lines

e.

High Steam Line Flow Coincident with.Low Steam Line Pressure or Low-Low Tavg.

f.

High-High Steam Generato.r Water Level

g.

Reactor Trip Coincident with Low Tavg.

2.

The containment isolation* system isolates those system which are not required for the mitigation of accidents specified in Section 14 of the FSAR.

A review.of Salem design has demonstrated conformance with these require-ments.

The valves and systems isolated by the various isolation signals are indicated in Table 5.4-1 and Figures 5.4-1 through 5.4-27 of the FSAR.

All lines penetrating the containment are shown in these figures along with their isolation provisions.

All non-essential systems are.

either automatically isolated upon a containment isola-tion signal, or provided with non-return check valves, or closed during power operation and under administrative control.

Essential systems are not isolated since they M P79 54 01/18 Salem 1 & 2 JAN 1 1980

are required to perform functions needed to maintain the plant in a safe condition following an accident.

These essential systems are as follows:

Residual Heat Removal - part of Safety Injection Safety Injection Contain*ment Fan Coolers - Service Water Steam Supply to Auxiliary Feedwater Pump Turbine Main Steam Atmospheric Relief Auxiliary Feedwater Charging - Portion for Safety Injection The Westinghouse Owners Group has prepared a report entitled "Classification of Lines Penetrating Containment and a Review of Containment Isolation Logic and Philosophy".

This report has been reviewed for applicability to Salem Units 1 and 2.

Salem conforms with the established essential/non-essential categories and recommended isolation provisions.

No further changes are required in the Salem design other than those previously noted in response to item 2.1.4.4.

3.

As stated previously, all non-essential systems are either isolated upon containment isolation signals, or provided with non-return check valves, or closed during power operation and under administrative control.

M P79 54 01/19 Salem 1 & 2 JUL 1 1980

4.

A review of the containment isolation valve control systems has been performed to verify that the valves remain closed upon resetting of the isolation signal until the operator takes deliberate action to reposi-tion them.

As a result of the review, design changes have been initiated to modify the control circuitry in two areas.

The results of this review, including a description of the two areas where modifications were deemed warranted, were submitted on July 13, 1979 in response to IE Bulletins79-06A.

Implementation of the design changes will be completed in accordance with the Category A implementation schedule.

M P79 54 01/20 Salem 1 & 2 i 1980

Integrity of Systems Outsiqe Containment likely to Contain Radioactive Materials for PWRs and BWRs (Section 2.1.6.a)

NRC Position Applicants and licensees shall immediately implement a pro-gram to reduce leakage from systems outside containment that would or could contain highly radioactive fluids during a serious transient or accident to as-low-as-practical levels.

This program shall include the following:

1.

Immediate Leak Reduction

a.

Implement all practical leak reduction measures for all systems that could carry radioactive fluid out-side of containment.

b.

Measure actual leakage rates with system in opera-tion and report them to the NRC.

2.

Continuing Leak Reduction Establish and implement a program of preventive mainten-ance to reduce leakage to as-low-as-practical levels.

This program shall include periodic integrated leak tests at a frequency not to exceed refueling cycle intervals.

CLARIFICATION Licensees shall, by January 1, 1980, provide a summary des-cription of their program to reduce leakage from systems outside containment that would or could contain highly radioactive fluids during a serious transient or accident.

Examples of such systems are given on page A-26 of NUREG-0578.

Other examples include the Reactor Core Isola-tion Cooling and Reactor Water Cleanup (Letdown function)

Systems for BWRs.

Include a list of systems which are ex-cluded from this program.. Testing of gaseous systems should include helium leak detection or equivalent testing M P80 61 01/1 Salem 1 & 2 JUL 1 1980

methods.

Consider in your program to reduce leakage poten-tial release paths due to design and operator deficiencies as discussed in our letter to you regarding North Anna and Related Incidents dated October 17, 1979.

RESPONSE

PSE&G has established and implemented a comprehensive leak reduction program, which is intended to maintain the leakage rates as-low-as-practical, for the following systems outside containment:

1.

The RHR System, in its entirety.

2.

The Safety Injection System - all those portions which have direct contact with the containment building.

3.

Containment Spray System -

only the portion which would have direct contact with the RHR system from the isolation valve to the containment building.

4.

CVC System -

the operational portion of the system which includes the letdown heat exchanger lines, seal water heat exchangers, centrifugal charging pumps, and lines to and from the volume control tank.

M PSO 61 01/2 Salem 1 & 2 JUL 1 1980

5.

Waste Gas System -

the entirety of the system except for the lines to and from the nitrogen supply portion of the system.

6.

Liquid Radwaste System -

the waste evaporator and the waste holdup tanks.

7.

Sampling Syste'm - all sample lines which have direct contact with the primary systems.

The systems and portions of systems excluded from the pro-gram are:

the Boric Acid Recovery portion of the eve System, because Boric Acid Recovery will not be used during an accident; the Primary Water Recovery portion of the eve System, because it is not anticipated that this portion of the eve System will be utilized during an accident; the liquid radwaste and gaseous radwaste systems contain portions which will not be utilized during an accident.

In the Liquid Radwaste System, the entirety of the Liquid Radwaste System except for the waste evaporator and the waste holdup tanks is excluded from the program.

The Gas System nitrogen supply portion is excluded since this portion of the system will not be radioactive.

M PBO 61 01/3

-17a-Salem 1 & 2 JUL 1 1980

I

. f The Containment Spray System contains portions which will not be directly in contact with the radioactive fluid imme-diately after an accident.

Most of the Containment Spray System is in this category.

During an accident the Contain-ment Spray System utilizes water from the refueling water storage tank.

Once this source is depleted the pumps are shut down.

After the pumps are shut down, supply water to the Containment Spray System is taken from the RHR System.

Thus, only th~ section of the system directly in contact with the RHR System will contain radioactive fluid.

The program consists of integrated leak tests of these sys-tems.

The baseline study and the actual leak rates with the systems in operation will be submitted as soon as practical, but no later than prior to entering Mode 4 following the next refueling on each unit.

At intervals not to exceed each refueling outage, an operat-ing pressure leak test will be performed on portions of the Safety Injection System, Residual Heat Removal System, Chem-ical and Volume Control System, Reactor Coolant Sampling, M P80 61 01/4

-17b-Salem l & 2 JUL 1 1980

Liquid Radwaste, Gaseous Radwaste, and Containment Spray Systems.

The pressurized systems will be visually inspected for leakage into the building environment.

Any observed leakage will be corrected to the extent reasonably practi-cal.

Where feasible, leakage from liquid containing systems will be determined by counting the number of drops from each system.

Where a system is not normally operating, such as in the RHR and the Safety Injection System, the leakage will be determined by pressurizing the applicable portions of each system to operating pressure and a walk-through of the system will be made to determine the amount of leakage from the system.

In the Gaseous Radwaste System, the leak-age will be determined by operating the waste gas compres-sors in the recirculation mode while maintaining the system pressure using a regulated nitrogen source.

The entire Gaseous Radwaste System, including the gas decay tanks and their relief valves, will be pressurized and leak tested.

The system make-up rate will be determined with a gas flow-meter at the supply regulator.

All nonwelded connections in the system will be checked with a soap-and-water solution to locate any leaks and appropriate means will be taken to eliminate these leaks.

M P80 61 01/5

-17c-Salem 1 & 2 JUL 1 1980

The operating leak reduction program consists of a daily review of radioactive liquid waste processing, calculated containment leakage, and leakage observations.

The daily leakage observations include those portions of the pre-viously identified systems which ar~ not shielded, enclosed, or otherwise controlled.

Using this information, unidenti-fied leakage baseline data is accumlated.

Leakage quanti-ties which exceed the baseline by a factor of two or more will be inves~igated to determine the source of leakage.

Another feature of the program is the use of strategically located air monitors in the Auxiliary Building.

These monitors have the capability to detect low leakage volumes of primary coolant at normal activity levels.

Automatic monitoring is supplemented with smear surveys in controlled areas which contain major components of the listed systems.

Utilizing these monitoring programs, investigations and cor-rective actions will be implemented to reduce system leakage outside of the containment.

When sources of leakage are identified, appropriate corrective measures to reduce the leakage will be initiated.

Should the correction of the leakage require reduction in operating mode, the corrections will be made during the next scheduled shutdown.

M P80 61 01/6

-17d-Salem 1 & 2 JUL 1 1980

PSE&G is taking other steps to eliminate unnecessary leakage into the Auxiliary Building.

A design review of the systems which may contain radioactive fluid immediately after an accident is being made to determine whether currently capped valve leakof f connections can be connected or hardpiped to the Liquid Radwaste System.

This review will be completed by August 1, 1980.

PSE&G has reviewed the ~esign of those systems which may contain radioactive fluid immediately following an accident, to assure design adequacy and elimination of as much leakage as practicable.

This review was performed in accordance with IE Circular No. 79-21, "Prevention of Unplanned Re-leases of Radioactivity."

The results of this review indi-cate that the system designs are acceptable.

M PBO 61 01/7

-17e-Salem 1 & 2 JUL 1 1980

Design Review of Plant Shielding and Environmental Qualif i-cation of Equipment for Spaces/Systems Which May be Used in Post-Accident Operations (Section 2.1.6.b)

NRC Position With the assumption of a post-accident release of radioac-tivity equivalent to that described in Regulatory Guides 1.3 and 1.4, each licensee shall perform a radiation and shield-ing design review of the spaces around systems that may, as a result of an accident, contain highly radioactive mate-rials.

The design review should identify the location of vital areas and equipment, such as the control room, rad-waste control stations, emergency power supplies, motor con-trol centers, and instrument areas, in which personnel occu-pancy may be unduly limited or safety equipment may be duly degraded by the radiation fields during post-accident opera-tions of these systems.

Each licensee shall provide for adequate access to vital areas and protection of safety equipment by design changes, increased permanent or temporary shielding, or post-acident procedural controls.

The design review shall determine which types of corrective actions are needed for vital areas throughout the facility.

CLARIFICATION Any area which will or may require occupancy to permit an operator to aid in the mitigation of or recovery from an accident is designated as a vital area.

In order to assure that personnel can perform necessary post-accident opera-tions in the vital areas, we are providing the following guidance to be used by licensees to evaluate the adequacy of radiation protection to the operators:

1.

Source Term The minimum radioactive source term should be equiv-alent to the source terms recommended, in Regulatory Guides 1.3, 1.4, 1.7 and Standard Review Plant 15.6.5. with appropriate decay times based on plant design.

M P80 61 01/8 Salem 1 & 2 JUL 1 1980

a.

Liquid Containing Systems:

100% of the core equilibrium noble gas inventory, 50% of the core equilibrium halogen inventory, and 1% of all others are assumed to be mixed in the reactor coolant and liquids injected by HPCI and LPCI.

b.

Gas Containing Systems:

100% of the core equi-librium noble gas inventory and 25% of the core equilibrium halogen activity are assumed to be mixed in the containment atmosphere.

For gas containing lines connected to the primary system (e.g., BWR steam lines) the concentration of radioactivity shall be determined assuming the activity is contained in the gas space in the primary coolant system.

2.

Dose Rate Criteria The dose rate for personnel in a vital area should be such that the guidelines of GDC 19 should not be exceeded during the course of the accident.

GDC 19 limits the dose to an operator to 5 Rem whole body or its equivalent to any part of the body.

When determining the dose to an operator, care must be taken to determine the necessary occupancy time in a specific area.

For example, areas requiring contin-uous occupancy will require much lower dose rates th~n areas where minimal occupancy is required.

Therefore, allowable dose rates will be based upon expected occupancy, as well as the radioactive source terms and shielding.

However, in order to provide a general design objective, we are providing the following dose rate criteria with alternatives to be documented on a case-by-case basis.

The recommended dose rates are average rates in the area.

Local hot spots may exceed the dose rate guidelines provided occupancy is not required at the location of the hot spot.

These doses are design objectives and are not to be used to limit access in the event of an accident.

M P80 61 01/9

-18a-Salem 1 & 2 JUL 1 1980

RESPONSE

a.

Areas Requiring Continuous Occupancy:

<lSmr/hr.

These areas will require full time occupancy during the course of the accident.

The Control Room and onsite technical support center are areas where continuous occupancy will be required.

The dose rate for these areas is based on the control room occupancy factors con-tained in SPR 6.4.

b.

Areas Requiring Infrequent Access:

GDC 19.

These areas may require access on a regular basis, but not continuous occupancy.

Shielding should be provided to allow access at a fre-quency and duration estimated by the licensee.

The plant Radiochemical/Chemical Analysis Labor-atory, radwaste panel, motor control center, instrumentation locations, and reactor coolant and containment gas sample stations are examples where occupancy may be needed of ten but not continuously.

The Salem station radiation shielding design basis includes TID-14844 assumptions for certain areas of the plant.

Con-siderable shielding is installed throughout the plant that would allow access to many other areas of the plant for shorter than normal periods of time after an accident even though the shielding in these areas was not designed specifically for post-LOCA sources.

M P80 61 01/10

-18b-Salem 1 & 2 JUL 1 1980

I The Primary Coolant Source terms used for this review are based on a release of 100% of the noble gases, 50% of the Halogens and 10% of the Cesiums, Bariums, and Strontiums.

These release fractions are considered representative of Regulatory Guide 1.4.

Calculations are based on 24 hrs.

decay.

Calculations are focused on areas in the Auxiliary Building and Penetration Areas.

Dose rates in the containment are being calculated for locations where necessary (post-LOCA) pieces of equipment and instruments are located.

The systems and areas reviewed include:

RHR System Safety Injection CVCS Demineralizer Area Charging Pump Compartments Reactor Coolant Filter Seal Water Filter Area Chemistry Lab M P80 61 01/11

-!Sc-Salem 1 & 2 JUL 1 1980

I Primary Sample Lab Fuel Handling Building Spent Fuel Pool Heat Exchanger Area Liquid Radwaste Control Room Technical Support Center Diesel Generator Compartments Diesel Oil Supply Tank Compartments

- Electrical Relay & Switchgear Rooms Gaseous Radwaste Valve Stations Liquid.Radwaste Valve Stations Component Cooling Accessibility to systems and areas:

Residual Heat Removal System -

Elev. 45' and 55 1 Aux. Bldg.

1)

The RHR pump compartments on elevation 45 1 in the Auxil-iary Building would have a genera~ area dose rate within the compartments of approximately 30,000 R/hr.

2)

The dose rate in the adjacent RHR compartment will be approximately 30 mr/hr.

This compartment is accessible while the other RHR system is operating.

M P80 61 01/12

-18d-Salem l & 2 JUL 1 1980

t

3)

The dose rates on elevation 55' from the operating RHR system below are approximately 8 R/hr.

This dose rate drops off by a factor of 2 after one week decay.

Lead sheet placed on the floor will further reduce the dose rate such that limited access is afforded to this area.

Permanent shielding in this area is required on an exposed portion of 14" RHR suction pipe.

Six inches of lead will be installed to shield this pipe.

This shielding will be installed by January 1, 1981.

4)

Access to either of the RHR pump compartments can be accomplished by draining and flushing each respective system.

Safety Injection System

1)

The Safety injection pump compartment is inaccessible while operating.

2)

Dose rates in adjacent areas, such as the Spent Fuel Pool Heat Exchanger area and Component Cooling Heat Ex-changer compartments are approximately 600 mr/hr at con-tact with the pipe chase and pump compartment shield walls.

This dose rate drops off substantially several M PSO 61 01/13

-18e-Salem 1 & 2 JUL 1 1980

feet from the walls.

Limited access is afforded to these areas ana no additional permanent shielding is planned.

Charging Pump Compartments

1)

Dose rates in the vicinity of these pumps are estimated to be 5000 R/hr, thus precluding access while the pumps are operating.

2)

The dose rate-through the wall separating the pump com-partments is approximately 5 R/hr.

3)

The dose rate outside the Chargin.g Pump compartments is approximately 200 mr/hr; therefore, access to the compo-nents in the general area is available.

4)

Charging pump valve compartment dose rates may be unacceptably high due to short lengths of exposed pipe and valves.

Further analysis of potential accidents which may require operation of the valves in these com-partments will be performed to determine whether perma-nent lead shielding should be installed.

Any shielding determined to be necessary as a result of this analysis will be installed by January 1, 1981.

M P80 61 01/14

-18f-Salem 1 & 2 JUL 1 1980

Chemical and Volume Control - Demineralizer Area

1)

Dose rates from the demineralizers would not have a sig-nificant effect on access.

Resins are changed upon either high radiation level or high pressure drop.

2)

The dose rates from piping and valves located behind valve aisle shield walls would be the major source of radiation and result in levels of approximately 1 R/hr in the operating aisles.

This would be reduced by decay and will afford sufficient access to the area for limited valve operations.

Reactor Coolant and Seal Water Filters

1)

The dose rates from these filters do not present a prob-lem, since the elements are replaced at predetermined radiation levels rather than high pressure drop.

Post-accident radiation levels in this area will not preclude access for filter changing.

Each filter is located in an individual shielded compartment.

M P80 61 01/15

-18g-Salem l* & 2 JUL 1 1980

Primary Sample Lab

1)

Dose rates from Primary Sample System tubing that would be used to draw a Reactor Coolant sample are calculated to be approximately 50 R/hr at contact and 2 R/hr in the general area of the lab.

These dose rates would be higher at T = o.

Permanent shielding will be installed on those sample lines to be used for post-accident sam-ples.

This shielding will be installed by January 1, 1981.

Counting Room

• l)

Direct dose rates in the counting room are not signifi-cantly affected by accident radiation source terms due to the location of the counting room.

Fuel Handling Building

1)

Dose rates in the Fuel Handling Building due to direct radiation from the Containment will not be significantly affected.

The only exception to this is streaming from the elevation 130' Containment Personnel Hatch and through the doorway into the Fuel Handling Building at M P80 61 01/16

-18h-Salem l & 2 JUL 1 1980

f.

elevation 130'.

This would be minimized by placing tem-porary block shielding in front of this doorway.

2)

The dose rates at the Spent Fuel Pool Heat Exchanger and Pump area in the Auxiliary Building are estimated to be approximately 600 mr/hr, thus affording limited access to this area.

Areas to Which Access May be Required Following an Accident For the purpose of this study, the areas discussed below are considered vital areas.

Accessibility is based on direct radiation levels due to contained radiation sources.

Control Room The Control Room is located on elevation 122' and is suffi-ciently shielded from systems containing highly radioactive fluids.

The radiation levels in the Control Room due to direct dose rates from the systems that may be required to operate after an accident are in the millirem range.

M P80 61 01/17

-18i-Salem 1 & 2 JUL 1 1980

Technical Support Center The Technical Support Center is located in the Clean Facili-ties Building and the doses due to the systems that will be operating in the Auxiliary Building are negligible.

The doses to individuals in this building over the course of an accident are due to the cloud dose from releases from the plant, and with the available shielding, result in a whole body dose of less than 3 rem.

Areas in the Auxiliary Building That do not Contain Highly Radioactive Sources of Radiation but may Require Access These areas include:

Diesel generator compartments Diesel oil supply tank compartmen*ts Electrical relay and switchgear rooms Preliminary analysis shows that sufficient shielding exists between these areas and adjacent compartments that contain radiation sources such that access to these areas is not M P80 61 01/18

-18j-Salem 1 & 2 JUL 1 1980

precluded.

When access requirements to these areas are established, doses to personnel will be estimated.

These rooms are, in most cases, located several compartments away from the compartments containing the systems which would be used after an accident.

Access to Areas in the Auxiliary Building Which May Contain Highly Radioactive Sources The hydrogen purge controls and containment isolation valve reset controls are operated from the Control Room.

Access to other areas of the Auxiiiary Building related to this equipment is unnecessary.

Chemistry Lab The chemistry lab is located on elevated 100' -

Auxiliary Building.

It is sufficiently shielded such that the major contribution to the dos~ rate in the lab is due to streaming from the containment personnel hatch.

The shielding between the chemistry lab and the hatch reduces the radiation level in the chemistry lab from this source to less than 20 mr/hr.

Gaseous Radwaste Control Center The valve operating station for the gas decay tanks is accessible.but will require additional shielding of valves M P80 61 01/19

-18k-Salem 1 & 2 JUL 1 1~0

I and small bore piping.

The amount of shielding required is presently being evaluated.

This shielding will be installed by January 1, 1981.

Liquid Radwaste Control Station (Valve Areas)

The valve station is shielded from the tank compartments and access to the area is possibl~.

Dose rates in this area from the tank itself would be approximately 2.5 R/hr.

Re-mote operators will be installed on valves located in the valve compartment since, high radiation levels from the asso-ciated piping will preclude access to the valve compartment.

Component Cooling Pump and Auxiliary Feedwater Pump-and Valve Areas These areas are located on elevation 84 1 The dose rates

~-

. from shielded sources adjacent to this area are approxi-mate~y 1800 mr/hr (maximum).

This does not preclude access to this area.

M PSO 61 01/20

-181-Salem 1 & 2 JUL 1 1980

Piping Layout for Systems Containing Highly Radioactive Fluids The piping layouts for the systems described above have been reviewed.

Most of this piping is shielded.

Piping that is not presently shielded and would affect access to an area due to high dose rates from that piping will be shielded by January 1, 19 81.

Equipment Qualification - Capability to Withstand Possible Radiation Doses Following an Accident All vital pieces of equipment in the RHR, HPSI, LPSI, and CVC systems have been identified.

Integrated doses after 120 days were calculated for the following pieces of equip-ment:

RHR pumps, Safety Injection pumps and Centrifugal Charging pumps.

The direct dose to each o~these pieces of equipment is 2.3 x 107 rem after a period of 120 days.

These doses were calculated using the TID source terms over a 30 day elapsed period of time.

The Containment Spray pumps are also considered vital equip-ment, but these pumps will not be exposed to any radioactive M P80 61 01/21

-18m-Salem 1 & 2 JUL 1 1980

fluid.

During the injection phase, the pumps take suction from the RWST.

During the recirculation mode, the RHR pumps supply the spray headers and the containment spray pumps are not run.

The capability of this equipment to withstand the radiation exposures is being evaluated.

This evaluation and any necessary modifications will be completed by January 1, 1981.

Additionally, radiation exposures for all Class lE electrical equipment is being evaluated (in conjunction with IE Bulletih 79-0lB).

Other Considerations

~

Local and Portable Shielding Temporary shielding such as lead bricks, lead blankets, and lead sheet is available at the station for use where small quantities of shielding may be required to shield local hot spots.

M P80 61 01/22

-18n-Salem 1 & 2 JUL 1 1980

It is intended, however, to install permanent shielding where possible to reduce the amount of temporary shielding required.

Source Terms The source terms used for this study were based on one-day decay.

The calculated dose rates would be reduced by a factor of approximately 26 at 30 days after the start of an accident.

For systems such as primary sampling, where access is required at one hour after an accident, dose rates would be a factor of 10 higher than the one-day decay re-sults.

In this instance, more shielding may be added as necessary.

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-180-Salem 1 & 2

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Improved Post-Accident Sampling Capability (Section 2.1.8.a)

NRC Position A design and operational review of the reactor coolant and containment atmosphere sampling systems shall be performed to determine the capability of personnel to promptly obtain (less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />) a sample under accident conditions withou~ incurring a radiation exposure to any individual in excess of 3 and 18 3/4 Rems to the whole body or extremities, respectively.

Accident conditions should assume a Regulatory Guide 1.3 or 1.4 release of fission products.

If the review indicates that personnel could not promptly and safely obtain the samples, additional design features or shielding should be provided to meet the criteria.

A design and operational review of the radiological spectrum analysis facilities shall be performed to determine the capability to promptly quantify (less than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) certain radioisotopes that are indicators of the degree of core damage.

Such radionuclides are noble gases (which indicate cladding failure), iodines and cesiums (which indicate high fuel temperatures), and non-volatile isotopes (which indicate fuel melting).

The initial reactor coolant spectrum should correspond to a Regulatory Guide 1.3 or 1.4*

release.

The review should also consider the effects of direct radiation from piping and components in the auxiliary building and possible contamination and direct radiation from airborne effluents.

If the review indicates that the analyses reqµired cannot be performed in a prompt manner with existing equipment, then design modifications or equipment procurement shall be undertaken.to meet the criteria.

In addition to the radiological analyses, certain chemical analyses are necessary for monitoring reactor conditions.

Procedures shall be provided to perform boron and chloride chemical analyses assuming a highly radioactive initial sample (Regulatory Guide 1.3 or 1.4 source term).

Both analyses shall be capable of being completed promptly; i.e., the boron sample analysis within an hour and the chloride sample analysis within a shift.

P79 131 01/10 Salem 1 & 2 JAN 1..1.0.Rf':

Response

A design and operational review of the containment atmosphere and reactor coolant sampling systems has been performed to determine the capability of personnel to promptly and safely obtain a sample under post-accident conditions within the time and exposure constraints identified above.

A piping Arrangement Drawing, an Instrument Schematic and Controls Logic Diagrams are attached.

(Drawing Nos. SK-12879, 207510-B-9491, 248251-B-9803 through 248254-B-9803) ~

The containment components of the Radiation Monitoring System are utilized to acquire a containment air grab sample.

This portion of the system was designed for normal operating conditions.

Therefore, under the relatively higher pressures and temperatures which could be expected during an accident, the existing air sampling pump would fail and a representative sample of containment atmosphere would thus be unobtainable.

Acquisition of a post-LOCA sample is furthe~ precluded by the fact that the grab sample location point is in the ~lectrical penetration area, an area which becomes inaccessible during an accident due to radiation streaming through the surrounding penetrations.

In addition, the Radiation Monitoring System containment isolation valves close upon a containment isolation signal.

P79 131 02 Salem 1 & 2 JAN 1 1980 J

The Reactor Coolant Sampling System utilizes Reactor Coolant System pressure to acquire a sample.

Should an unpressurized condition exist after an accident, the head available is insufficient to draw a sample to the primary sampling laboratory.

Also, the post-accident containment flood level is such that the sample system components inside the containment isolation valves would be under water.

As with containment air sampling, the Reactor Coolant Sampling System containment isolation valves close upon a containment isolation signal.

Proposed Modifications As a result of this review, modifications will be made to enable containment atmo~phere and reactor coolant sampling to be 9erformed in an expeditious (within one hour) and safe (within allowable dose criteria to personnel) manner in the event of an accident.

New design features will be added for use during accident conditions only while continuing to use the existing features for normal operations.

The new features will be designed for a containment environment of 50 psig and 350°F and reactor coolant conditions of 2485 psig and 650°F.

P79 131 03

-23a-Salem 1 & 2 JAN 1 1980

These safety related additions will be designed to Nuclear Class II, Seismic Category I criteria.

The post-LOCA containment atmosphere sampling system will consist of.two independent, electrically separated loops for each unit while the post-LOCA Reactor Coolant Sampling System will have two electrically separated lines and equipment for each unit up to the primary sampling laboratory at which time they will be tied into the existing sample lines and equipment.

The design will maintain physical and electrical separation as much as possible throughout the systems.

In the post-LOCA containment air sampling system design, each unit.will have redundant air supply and return lines.

The existing inside containment supply and return lines from the Radiation Monitoring System will be utilized by teeing upstream of the inside containment isolation valves (IVC7,9,ll,13 and 2VC7,9,ll,13).

The samples will thus be drawn from and returned to elevation 145' inside containment.

Upon teeing into the two supply and two return lines, each new pair of supply and return lines will be run through ~eparate electrical penetrations.

In the post-LOCA Reactor Coolant Sampling System design, new sampling lines will tee into the existing lines off the #11 and #13, (#21, and #23) hot legs, upstream of P79 131 04

-23b-Salem l & 2 JAN 1 1980

valves 11 and 138832 (21 and 238832).

The new lines will be run through separate electrical penetrations.

Each of the new sampling lines inside containment will be provided with a 150-foot delay coil which will allow for decay of some of the short-lived isotopes prior to the sample reaching the primary sampling laboratory.

Each of the six new lines (four containment air and two reactor coolant) will have normally closed/fail closed, air-operated isolation valves inside and outside containment.

To meet channel separation criteria, the the isolation valves will have different vital power supplies to their solenoid coils.

The outside containment isolation valves will be supplied with redundant control air lines.

Backup pressurized air accumulators will be provided for the inside containment isolation valves.

Each accumulator will be sized for approximately 1000 cycles of operation (more than a one-month supply, based on the conservative assumption that a sample will be taken once an hour dµring the initial month following an accident).

All air accumulators and isolation valves inside containment will b~ mounted on platforms above the containment flood level.

For post-LOCA containment air sampling, there will be one sampling location in each unit, each with the ability to P79 131 05

-23c-Salem 1 & 2 JAN 1:, 19_, 8_0 __

draw a sample from either unit.

The Unit 1 location will be in the Auxiliary Building on elevation.84' in the spent fuel pool heat exchanger compartment while the Unit 2 location will be in the Auxiliary Building on elevation 100'.

The Unit 1 location will have two radioactive gas processing pumps for drawing Unit 1 samples, two sample stations for acquiring Unit 1 samples, two sample stations for acquiring Unit 2 samples, a panel for Unit 1, and a panel for Unit 2.

The Unit 2 location will have the same items with the exception that the two pumps will be used for drawing Unit 2 samples.

The 0.5 cfrn pumps will be of the dual containment stainless steel bellows type.

The dual containment feature contains an inter-barrier leak test port which will provide for an early indication of degradation.

Chilled water will be provided to the pumps to cool the motors.

The pumps will have the same vital power supply as the containment isolation valves in each respective loop.

Check ~alves in the pump discharge line will prevent containment air from reaching the pump through the return line to the containment.

Each of the sample stations will be provided with permanent in-line, stainless steel sample vessels.

The panels will be provided with valve position indication along with valve and pump controls.

In addition, the panels will have P79 131 06

.. -23d-Salem 1 & 2 JAN t-l980

phone jacks for communication with the Control Room.

Hoods will be provide~ at the sample stations to capture any gases released when obaining the sample and exhaust them to the plant ventilation system.

Area radiation monitors will be installed in both of the sampling locations with Control Room indication.

In acquiring a post-LOCA containment air sample, the sampling location in the affect~d unit will be utilized.

To prevent flow to the sampling location in the other unit normally closed solenoid valves will be provided in the supply and return lines.

If the primary sampling location is unavailable, these normally closed valves will be*

opened to allow the sample to be taken in the other. unit:

• Af~er obtaining a containment air sample, a 100 psi nitrogen purge will be used to purge any airborne particulate in the tubing back into containment.

A check valve in the nitrogen line will prevent contamination of the nitrogen supply~

For reactor coolant sa~pl*ing, the sampling locations for both units will be in the primary sampling laboratory located in the Auxiliary Building of Unit l on elevation 110'.

The existing equipment in the primary sampling laboratory will be used to process and analyze the sample, the equipment having been designed to handle fluids to P79 131 07

-23e-Salem l & 2

.JAN i-1~80

2485 psi and 650°F.

At the lab, new sample lines will tee into the existing reactor coolant sa~ple liries.

Sample pressure and temperature will be reduced to within reasonable l"imits prior to reaching the sample vessel br sample sink.

The existing hoods at the sample sink are connected to the building exhaust system.

New panels, one for each unit, wili be installed in the lab.

The new panels will provide valve position indications along with valve and pump controls.

In addition, the panels will have phone jacks so that Control Room communication can be maintained while personnel are acquiring samples.

The area radiation monitors in th~ lab provide Control Room indication and alarms.

Each unit will be supplied with two reactor coolant sampling pumps located in the boric acid evaporator compartment for each unit.

The 0.25 gpm pumps will be provided with the same vital power supply as their corresponding containment isolation valves.

These pumps will be used when the reactor coolant pumps are not operating and the system pressure is insufficient to provide a sample to the lab.

A sensor will be installed on the suction side of the pump.

When the containment isol~tion valves are opened a timer will start to ensure enough time expires to allow the sample to reach the P79 131 08

-23f-Salem 1 & 2 JAN

pumps.

If pressure is sensed, the pumps will be bypassed.

After the sample is taken, the pump will be shut off and containment isolation reestablished.

The lines will be flushed with primary water to the drain header.

To minimize radiation effects, tubing will be routed through normally or potentially radioactive compartments wherever possible.* Additionally, high energy break analyses will be performed.

Tubing in the containment, electrical and mechanical penetration areas, and the pipe alley will not be shielded since these areas are already inaccessible in a post-LOCA condition.

A small portion of the tubing will be routed through the switch gear rooms.

Radiation shielding and tube rupture protection will be provided in the switchgear rooms.

Molded lead shielding will be provided to hou~e.the tubing runs through the Spent Fuel Pool Heat Exchanger Compartments, Safety Injection Compattments, Unit 2 Component Cooling Compartment, Unit 2 Sampling location room on elevation 100'~ boric acid evaporator compartments, and the primary sampling laboratory.

All P79 131 09

-23g-Salem 1 & 2

. Ill M

_J

tubing which must cross from one unit to the other will also be provided with molded lead shielding and will pass through a penetration on elevation 114' of the Auxiliary Building.

Supports for all the tubing runs will be designed to Seismic Category I criteria.

Sample vessels will be wrapped in lead.

Also, to reduce contact exposure to extremities, all manual valves will be provided with long extension stems.

A study of the primary sampling laboratory has been per-formed to further assess additional shielding provisions which may be required to reduce background levels of radia-tion and exposure to personnel to as low as reasonably achievable.

Interim Methods Containment Atmosphere Sampling:

In order to provide the necessary interim method for obtain-ing a containment atmosphere sample under accident condi-tions, modifications have been made to the installed sample point for containment air sampling to enable placement of the sample rig for collecting the particulate, iodine and gaseous sample in a remote location out of the high radia-tion field.

To accomplish this, the existing sample and return lines were lenghtened.

No shielding of the sample M P80 38 10/l

-23h-Salem 1 & 2 JUL 1 1900

lines is necessary except at the sampling rig.

Properly r

shielded containers for the samples, radiation monitoring instruments, airborne and surface contamination control have been provided.

The analysis of the particulate and iodine samples will be performed with existing analysis equipment in accordance with current procedures, which have been modified to accept the high activity samples.

The iodine can be analyzed with portable dual channel sodium iodide analyzers, using distance or collimation to reduce the apparent activity the detector sees to a proper level.

Shielding of the sample

• will limit personnel exposure.

The analysis of the gaseous sample for hydrogen and noble gases will be done using current procedures employing only microliter size samples.

The modifications to equipment and procedures were completed on May 1, 1980.

Reactor Coolant Sampling:

In order to provide an interim means of obtaining a ~eactor coolant sample under accident conditions, an off-line sampling arrangement has been provided.

A sample from the RCS will be obtained as follows (see attached figures).

M P80 38 1.0/2

-23i-Salem l & 2 JUL 1 1980

All normal sampling lines will be isolated due to the accident.

Valves 1, 2 and 3 will be opened and valve 4 closed.

The normal path from the RCS sample cooler to the sample hood in the primary sample room will be isolated to avoid treating high radiation levels in that room.

The Control Robm will be notified to reset the isolation signal, thereby allowing the sampling personnel to remotely open the necessary sample valves to establish flow through the cooler and into the off-line sample arrangement.

The flow will be monitored by initially viewing the flowmeter, and then by monitoring a radiation meter with a remote detector viewing the shielded collector drum.

After an appropriate sample purge period, (approx. 4 minutes of purge), the sampling personnel close remote operated sample valves and terminate flow to cooler.

Valve number 2 is then closed and valve 4 may be opened to drain required sample volume (1/2 ml) from the sample line into shielded sample bottle.

If additional split samples are necessary, additional aliquots may be taken.

Valve 3 is closed to prevent backflow from the collector.

)

Radiological considerations are incorporated.

Valves 2, 3, and 4 have reach rods through the lead shielding.

The small M PSO 38 10/3

-23j-Salem 1 & 2 JUL 1 1980

I sample aliquot can be withdrawn on a slide through a port in the lead shielding, the collector bottle is heavily shielded and remote from the sampling personnel and is vented through a charcoal filter to the plant vent.

Viewing of the sampling operation will be done by the sampling personnel using mirrors.

Tubing is small bore stainless steel to minimize radiation source strength.

Existing techniques and procedures for boron and gamma analysis will be used.

Boron analysis is performed using an automatic piston burette and pH meter, and will be locally shielded in the laboratory.

Appropriate microliter samples can be taken from the aliquot and diluted as necessary for gamma analysis.

M P80 38 10/4 1/4

-23k-Salem 1 & 2 JUL 1 1980

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Increased Range of Radiation Monitors (Section 2.1.8.b)

NRC Position The requirements associated with this recommendation should be considered as advanced implementation of certain requirements to be included in a revision to Regulatory Guide 1.97, "Instrumentation to Follow the Course of an Accident," which has already been initiated, and in other Regulatory Guides, which will be promulgated in the near-term.

1.

Noble gas effluent monitors shall be installed with an extended range designed to function during accident conditions as well as during normal operating conditions; multiple monitors are considered to be necessary to cover the ranges of interest.

a.

Noble gas effluent monitors with an upper range of 105 uCi/cc (Xe-133) are considered to be practical and should be installed in all operating plants.

b.

Noble gas effluent monitoring shall be provided for the total range of. concentration extending from a minimum of 10-7 uCi/cc (Xe-133).

Multiple monitors are considered to be necessary to cover the ranges of interest.

The range capacity of individual monit"&rs shall overlap by a factor of ten.

2.

Since iodine gaseous effluent monitors for the accident condition are not considered to be practical at this time, capability for effluent monitoring of radioiodines for the accident condition shall be provided with sampling conducted by adsorption on charcoal or other media, followed by onsite laboratory analysis.

3.

In-containment radiation level monitors with a maximum range of 108 rad/hr shall be installed.

A minimum of two such monitors that are physically separated shall be provided.

Monitors shall be designed and qualified to function in an accident environment.

M P79 54 01/32 Salem 1 & 2 1 19PI/

• • ---**** ~-.

Response

1.

The plant vent gaseous monitors have the following detection range capab i1 it ies:

Unit 1:

sx10-6 to sx10-l uCi/cc Xe-133 Unit 2:

lx10-6 to lxlo2 uCi/cc Xe-133 Design changes have previously been initiated, and equipment purchased to upgrade the detection range capability of Unit 1 to that of Unit 2.

Further design modifications are presently being evaluated to provide the gaseous monitors with a detection range capability of io-7 to 105 uCi/cc Xe -

133.

The modified system will utilize multiple monitors with the required overlap to rn~et the above cr~teria.

An alternate consideration is the use of a de*tector with

~-

a range of 104 uCi/cc if the containment exhaust is diluted by at least a factor of 10.

These modifications will be completed by January 1~ 1981.

In the interim, a single thin wall GM tube will be positioned on top of the auxiliary building approximately 150 feet away from each unit's vent.

In the avent of a major discharge of activity, the count rates displayed by the existing radiation monitors will 1ndicate which unit is releasing the high activity.

If the radiation f icld M P79 54 01/33 Salem 1 & 2 MAR 2* 8 1980

~............ -*-** *-*--*. --------'---------------

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t produced is high enough, then radiation levels at the GM tube will be measurable.

Curves will be used to relate the direct radiation dose rates to the activity dis-charged.

Dose rates in mR/hr are converted into micro-curies per cc using standard volume source calculations.

The properties of the instrument are provided below:

Instrument -

Halogen q.uencb~d,. GM tube.

Range -

0.1 mR/hr to io4 mR/hr.

Energy Dependency -

I: 20% from 100 keV to.2.s MeV.

Calibration Frequency -

Every 18 months.

Location -

150 1 away from each unit vent with the GM tube located in a shielded lead cavei Background Correction - Existing equipment is capable of adequately monitoring low range requirements.

Cosmic ray background radiation is small enough compared to the dose rates being measured not to be a significant problem.

Two inches of lead shielding around the detector ~ill minimize detector response to other background radiation sources.

Access To Readings ** Readings will be available contin-uously and will be recorded in the Unit 1 Control Room.

~

Power Source - Vital power will be provided.

M P79 54 01/33.l

-25a-Salem l & 2 HAR 2 8 l980

Calculational Methods and Information Dissemination The calculational method used for converting instrument readings to release rates are provided in the instruc-t ions for interpreting monitor response.

Calculations using the ORIGEN code were used to provide spectrum dis-tribution and average'energy values.

Analysis indicates that within a few hours after reactor shut-down, Xenon 133 becomes the predominant isotope.

During an accident the channel readings will be recorded and the information will be posted on an emergency status board in the on-si te Technical Support Center.

Since the channel measures the production of ion pairs produced in the tube, calibration is performed by determining the GM tube plateau..

This interim modification was operational on April 4, 1980.

The procedures for performing this analysis have been incor-porated into the appropriate station procedures.

Interim measures will be employed in the event of high noble gas release rates from the atmospheric steam relief valves.

To quantify the noble gas release rate, calculations have been made which relate the release rate to the dose rate at a prescribed distance from the relief valve steam plume.

A thin-window, hand-held instrument will be used to take read-ings during steam relief and correlate the dose rate and dis-tance to the released activity by reference to a curve.

This guidance has been incorporated into the appropriate station procedures.

M P79 54 01/33.2

-25b-Salem l &*2 JUL 1 1980

2.

The Salem design provides for iodine sampling by adsorption on charcoal cartridges, followed by onsite laboratory analysis.

3.

The containment high range monitors presently have the following maximum detection ranges:

unit 1:

104 R/hr.

Unit 2:

107 R/hr.

M P79 54 01/33.4

-25c-Salem 1 & 2 JUL 1 1980

One monitor is provided for each unit.

The Unit 2 monitor has undergone environmental qualification to demonstrate proper operation in an accident environment.

In addition, this monitor has been calibrated in a special test ~acility to verify proper readings in high radiation fields.

In order to meet the requirement for monitors with a range of 108 R/hr, we are investigating the possibility of shielding the existing.Unit 2 monitor.

An alternate consideration is the use of the existing 107 R/hr (gamma) monitor.

An additional monitor with similar range capability will be installed to meet redundancy requirements.

Installation of new monitors for both units will be completed by January 1, 1981.

M P79 54 01/34 Salem 1 & 2

ATTACHMENT 2

ATTACHMENT 2

• III.D.3.4 - Control Room Habitability An analysis of control room habitability as an result of an accidental release of radioactivity was performed in re-sponse to FSAR Question 14.16.

The analysis demonstrated that potential exposure levels to the control room operators are below the maximum design basis doses specified in 10CFR50, Appendix A, Design Criterion 19.

The response to FSAR Question 9.60 provided an analysis of the toxic chemical shipments transported by waterway past the Salem site, including frequency of shipments and quantities of toxic materials shipped.

Regulatory Guide 1.78, Paragraph C.2, states that hazardous chemicals such as those indicated in Table C-1 of the Guide, must be included in the analysis if they are frequently shipped within a five mile radius of the plant.

The Guide also defines frequent shipments as being 50 or more trips per year for barge traffic and specifies, in Paragraph C.l, that chemicals stored or situated at distances greater than 5 miles from the facility need not be considered.

Following is the analysis of Control Room habitability dur-ing a postulated hazardous chemical release occuring either on the site or within a five mile radius of the plant.

As indicated in Section 2.5.7 of the FSAR, the Salem site is located in a rural area with no major manufacturing or chemical plants located within five miles of the site.

The only major transportation route within five miles of the plant is the Delaware River, with the intra-coastal waterway passing 1 mile west of the site.

The Salem Nuclear Generating Station uses a sodium hypo-chlorite biocide system,-thus eliminating an onsite chlorine hazard.

The Control Room is equipped with smoke and com-bustible detectors located in the air conditioning unit ducts.

These detectors provide alarms in the Control Room in the event of smoke or combustible hazards present.

The Control Room is equipped with radiation detectors which pro-vide annunciation, automatically isolate the Control Room, and switch the ventilation system to the recirculation mode.

There are a. sufficient number of structures on the site be-tween any release point and the Control Room air intake, such that dispersion of any effluent is provided.

The calculated X/Q is sufficiently small to reduce any concentrations of these hazardous chemical releases to low values.

The Control Room air intakes for both Units 1 and 2

J,i I

• • • are located inboard of their respective containments, i.e.,

in between the containments on the north side of the Unit 1 containment and the south side of the Unit 2 containment.

Any potentially hazardous chemical is stored on the opposite side of the containment, thus prohibiting any direct undi-luted intake of hazardous chemicals.

It also should be recognized that these hazardous chemicals are stored at ground level (elevation 100 1 ), whereas the Control Room air intake is located at elevation 131'.

In the calculations, only a 15 meter elevation was assumed.

This location pro-vides, as a minimum, the full cross section of the contain-ment and adjacent buildings to act as a building wake factor.

Table 1 presents the chemicals stored on site or shipped by the site on the Delaware River which are identified in Regu-latory Guide 1.78, Table C-1.

Table 2 provides information on the Control Room ventilation system, as required by Regu-latory Guide 1.78, Paragraph C.7.

As can be seen from Table 1, the only hazardous chemicals stored on site are sulfuric acid and nitrogen.

The sulfuric acid is stored in a 3000 gallon tank, and calculations show that (assuming the failure of the tank) the rate of vaporization would be only 1.2 X lo-6 mg/second due to the low volatility of sulfuric acid.

Using a X/Q=7.35Xlo-4 second/cubic meter, for a distance of 600 feet, the sulfuric acid concentration at the Control Room air intake is calculated to be 8.81 X lo-10 mg/m3, which is much lower than the toxicity limit given in Table C-1 of Regulatory Guide 1.78.

Also, two tanks with a capacity of 4,000 gallons each, containing sulfuric acid, are located in the Turbine Building, at elevation 88'.

They are located on a diked area, and, assuming the failure of one tank, the rate of vaporization would be 1.2 X lo-6 mg/sec.

The Turbine Building ventilation system will dis-perse and evacuate the sulfuric acid vapor.

Assuming (conservatively) that there would be no dispersion from the Turbine Building exhaust to the Control Room air intake, the concentration at the Control Room air intake would be 1.2 X lo-6 mg/m3, much lower than the toxicity limit given in Table C-1 of Regulatory Guide 1.78.

Nitrogen is stored in a 500 gallon tank, 1600 feet away from the Control Room air intake, and in 36 bottles, each con-taining 300 cubic feet of nitrogen, at a pressure of 2300 psi located at elevation 122' in the Auxiliary Building.

Due to the small quantity stored, there would be no discern-ible increase in ~he natural concentration of nitrogen in

~. ()..

• the air in the event of a failure of the 500 gallon tank.

A failure of one nitrogen bottle would lead to the dispersion of its contents in the Auxiliary Building, dilution, and its evacuation by the Auxiliary Building ventilation system, 70' above the Control Room air intake at a speed of 7 m/sec and in a direction opposite the Control Room air intake.

There-fore, the nitrogen release would pose no significant hazard to Control Room personnel.

Fire fighting agents used at the Salem Nuclear Generating Station are Halon, stored in tanks located inside the Auxiliary Building at Elevation 144' and C02 stored at elevation 84' in the Auxiliary Building.

Since they_are stored inside the building, they will have no effect on the Control Room air intake.

Table 1 also _shows that the only mobile source of hazardous chemicals is the sulfuric acid shipped on the Delaware River.

As discussed in the response to Question 9.60, a total of 90 vessel trips per year were made in transporting a total of 73,000 short tons, which resulted in an average of 812 short tons per vessel per trip.

Assuming that in a collision accident, the vessel cargo is released, a quantity of 812 short tons is assumed to spill in the water.

The sulfuric acid would be diluted by the river water.

Since an insignificant amount of vaporization would occur, the Control Room habitability would not be affected by this event.

Conclusions Our analysis of the Control Room habitabiiity requirements demonstrates that the Control Room personnel are adequately protected against the effects of accidental release of toxic and radioactive gases, and shows that the plant can be safely operated or shut down under design basis accident conditions.

Due to the use of sodium hypochlorite, there is no chlorine hazard.

The postulated sulfuric acid spill results in a miniscule concentration reaching the Control Room air intake and the postulated nitrogen release will dissipate in the environment well before reaching the Control Room.

Therefore, no changes in the plant design are necessary.

GD 11 1/3-B

TABLE 1 HAZARDOUS CHEMICALS STORED ON-SITE NAME OF CHEMICAL TYPE OF SOURCE HUMAN DETECTION THRESHOLD (mg/m3)

MAXIMUM ALLOWABLE TWO-MINUTE CONCEN-TRATION (mg/m3)

MAXIMUM QUANTITY OF CHEMICAL IN-VOLVED MAXIMUM CONTINUOUS RELEASE RATE VAPOR PRESSURE (TON)

FRACTION OF CHEMICAL FLASHED AND RATE OF BOIL OFF WHEN SPILLING OCCURS DISTANCE OF SOURCE FROM CONTROL ROOM FIVE-PERCENTILE METEOROLOGICAL DILUTION FACTOR SULFURIC ACID ONSITE MOBILE 1

2

1) 3000 GAL.

ONSITE

2) 812 SHORT-TONS N/A N/A N//A 600 feet (onsite) 1 mile (mobile) 7.3sx10-4

• sec/m3 NITROGEN ON SITE ASPHYXIANT ASPHYXIANT

1) 500 GAL.

2) 1 Bottle =

300 cu. ft N/A 2300 psi for bottle

1) 25 % for tank

2) All for bottle 1600 Feet for tank 7.3sx10-4

• sec/m3

• The dilution factor is calculated as described in the response to FSAR Question 14.16.

The cross-sectional area of the containment was reduced to account for only the first 31 feet of elevation.

The wind speed was assumed to be 1.5 m/sec.

As indicated in the response to FSAR Question 14.16, the stability has little effect on X/Q, GD 11 04-B

TABLE 2 CONTROL ROOM VENTILATION SYSTEM PARAMETERS

l.

VOLUME OF CONTROL ROOM

2.

NORMAL FLOW RATES:

unfiltered inleakage

- filtered makeup air filtered recirculated air

3.

EMERGENCY FLOW RATES:

- unfiltered inleakage filtered makeup air filtered recirculated air

4.

TIME REQUIRED TO ISOLATE THE CONTROL ROOM NOTE:

CAA CS EACS 5000 m3 500 m3 100 cfm 100 cfm 0-33,840 0-33,840 (when needed)(when needed) 6850 cfm 6850 cfm 100 cfm 100 cfm 300 cfm 300 cfm (when needed) (when needed) 6550-6850 6550-6850 l minute manually (conser-vative) 5 seconds automatically EACS = Emergency Air Conditioning System CAACS = Control Area Air Conditioning System GD 11 05-B