ML20134J988

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Safety Evaluation Supporting Amends 190 & 173 to Licenses DPR-70 & DPR-75,respectively
ML20134J988
Person / Time
Site: Salem  
Issue date: 02/06/1997
From:
NRC (Affiliation Not Assigned)
To:
Shared Package
ML20134J983 List:
References
NUDOCS 9702130054
Download: ML20134J988 (18)
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Text

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h UNITED STATES g

j NUCLEAR REGULATORY COMMISSION 2

WASHINGTON, D.C. 20655-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS.190 AND 173 TO FACILITY OPERATING LICENSE NOS. DPR-70 AND DPR-75 f.UBLIC SERVICE ELECTRIC & GAS COMPANY PHILADELPHIA FLECTRIC COMPANY DELMARVA POWER AND LIGHT COMPANY ATLANTIC CITY ELECTRIC COMPANY SALEM NUCLEAR GENERATING STATION. UNIT NOS. 1 AND 2 DOCKET NOS. 50-272 AND 50-311

1.0 INTRODUCTION

By letter dated June 10, 1996, as supplemented June 24, July 1, August 13, September 20 and October 17, 1996, the Public Service Electric & Gas company (the licensee) submitted a request for changes to the Salem Nuclear Generating Station, Unit Nos. I and 2, Technical Specifications (TSs). The requested i

changes would revise TS 3/4.3.3.1, " Radiation Monitoring Instrumentation," and 3/4.7.6, " Control Room Emergency Air Conditioning System," to reflect a control room design in which the common Unit I and Unit 2 control room envelope is supplied by 2 one hundred percent capable Control Room Emergency Air Conditioning System trains. The June 24, July 1, August 13, September 20 and October 17, 1996, letters provided clarifying information that did not change the initial proposed no significant hazards consideration determination nor expand the scope of the initial submittal as described in the original Federal Reaister notice.

2.0 EVALUATION The licensee performed offsite and Control Room (CR) dose analyses for the following design basis accidents (DBAs); 1) loss of coolant accident (LOCA),

2) fuel handling accident (FHA), 3) locked rotor accident (LRA), and 4) the steam generator tube rupture accident (SGTR). The proposed amendment only changed the design of the control room envelope which would not affect calculations of DBA offsite dose analyses. However, the licensee submitted revised DBA offsite dose analyses along with the CR operator dose analyses.

The licensee's calculated doses met the applicable dose acceptance criteria of 10 CFR Part 100, Standard Review Plan (SRP) (NUREG-0800), and General Design Criterion (GDC) 19.

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. 2.1 Loss of Coolant Accident l

The licensee evaluated the CR and offsite doses resulting from a LOCA using the methodologies in Section 15.6.5 of the SRP (NUREG-0800) and Regulatory Guide (RG) 1.4, " Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss-Of-Coolant Accident for Pressurized Water Reactor."

The methodology used to evaluate the CR atmospheric dispersion factors (X/Q values) was based on the ARCON95 computer code, and the dose conversion 1

factors in ICRP 30 were used. Core radionuclide inventory was based on a i

power level of 3600 Mt which is 105 percent of the license rated power level of 3411 M t.

The revised LOCA calculations used a CR makeup flow of 2200 cfm.

l The updated CR dose analyses were based on the modified Control Room Envelope (CRE) arrangement, the upgraded Control Room Emergency Air Conditioning System (CREACS) design and performance parameters, selective air intake logic, and a

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charcoal filter efficiency for radioiodine of 95%.

The staff analyzed offsite and CR operator doses using the licensee's assumptions described above and the methodology of the SRP.

For CR operator doses, the staff used a revised version of the ARCON 95 computer code to assess the meteorological factors. The HABIT code was used to calculate the CR operator doses.

The doses are within the acceptance criteria of 10 CFR Part 100, SRP (NUREG-0800) and GDC-19. The assumptions used by the staff are presented in Table 1, and the resulting calculated doses are listed in Table 2.

2.2 Fuel Handling Accident The licensee calculated the radiation doses at the Salem station following a postulated fuel handling accident in the Fuel Handling Building. The CR calculation assumes a simultaneous loss of offsite power (LOOP) following the CR isolation signal generated by the operation of the CR intake radiation monitors of the CR ventilation system.

In performing this analysis, the licensee used the assumptions and methodology prescribed by RG 1.25,

" Assumptions Used for Evaluating the Potential Radiological Consequences of a fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors." The analysis assumes a decay time of 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> and 23 feet of water coverage.

The staff has completed its evaluation of the potential radiological consequences of an FHA at Salem on the basis of the conditions of the proposed TS changes.

In addition to reviewing the licensee's submittal, the staff performed an independent analysis to determine conformance with the criteria of 10 CFR Part 100 and GDC-19 of Appendix A to 10 CFR Part 50. The staff l

analysis utilized the assumptions contained in RG 1.25 and the review i

procedures specified in SRP Sections 15.7.4 and 6.4.

The staff assumed an instantaneous puff release of noble gases and radioiodine from the gap and plenum of the broken fuel rods. These gas bubbles will pass through at least 1

t l

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  • l 23 feet of water covering the fuel prior to reaching the containment atmosphere. All airborne activity reaching the containment atmosphere is assumed to exhaust to the environment within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. As stipulated in the plant TSs, the gap activity is assumed to have decayed for a period of 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br />.

l The staff computed the offsite doses for Salem using the assumptions described above and NRC's ACTICODE computer code. CR operator doses were determined using the methodology in SRP Section 6.4.

The computed offsite doses and CR operator doses are within the acceptance criteria given in SRP Section 15.7.4 l

and GDC-19. The assumptions used in calculating the doses are listed in Table i

3, and the resulting calculated dose values are in Table 4.

2.3 Locked Rotor Accident (LRA)

The licensee evaluated the CR and offsite doses resulting from an LRA as specified in Sections 15.3.3-15.3.4 of the SRP (NUREG-0800).

In accordance l

with the SRP, the reactor is initially assumed to be operating at 102 percent i

of the licensed power level. The licensee assumes a simultaneous LOOP l

following the high radiation alarm signal generated by the CR intake monitors, l

with the subsequent delay in switching from the normal operation mode to the l

emergency operation mode of the CR ventilation system. The calculation utilizes the automatic selection capabilities of the radiation monitor to select the less contaminated CR intake.

The LRA analysis assumes a pre-accident iodine spike and is based on 5 percent i

failed fuel. This 5 percent of the fuel in the core (briefly) enters departure from nuclear boiling (DNB) during the accident and is therefore i

assumed to fail. These fuel failures are assumed to occur instantaneously at the start of the accident.

The staff calculated the offsite and CR doses for the LRA by adjusting the I

license's calculations with staff-evaluated meteorological factors. The l

computed doses met the acceptance criteria of SRP (NUREG-0800) and GDC-19.

Table 5 presents the calculational assumptions used, and Table 6 presents the doses calculated by the staff.

2.4 Steam Generator Tube Rupture The licensee calculated offsite and CR doses for a steam generator tube rupture using the methodology of SRP 15.6.3.

Two assessments were performed for the most limiting scenario, which is an SGTR with a loss of offsite power l

and fully stuck open atmospheric dump valve. The assessments included an l

accident-initiated iodine spike and a pre-existing iodine spike. The staff l

calculated the offsite and CR doses for the SGTR by adjusting the licensee's calculations with staff-evaluated meteorological factors. The computed doses met the acceptance criteria of 10 CFR Part 100, SRP (NUREG-0800) and GDC-19.

The parameters used by the staff are presented in Table 7, and the doses l

computed by the staff are presented in Table 8.

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  • TABLE 1 INPUT PARAMETERS FOR SALEM UNITS 1 AND 2, EVALUATION 0F A LOSS-0F-COOLANT ACCIDENT Power level, (Nwt) 3600 Fraction of core inventory available for leakage, (%)

Iodines 25 Noble Gases 100 Initial iodine composition in containment, (%)

Elemental 91 Organic 4

Particulate 5

3 Primary Containment volumes, (ft )

Sprayed 1.56 x 106 Unsprayed 1.04 x 106 Primary containment leak rate, (%)

0-24 hours after accident 0.1 After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 0.05 Containment spray iodine removal efficiencies, (hr )

d Elemental (main sprayed region) 20 Organic 0

Particulate (main sprayed region) 6.7 Decontamination factor Elemental iodine 6.51 Particulate iodine 50 ESF leak rate, (cc/hr) 3840 Volume sump water (gallon) 295,000

)

3 Atmospheric dispersion factors, (sec/m )

Exclusion area boundary (0-2 hrs) 2.4 x 10

Low population zone (0-8 hrs) 2.2 x 10'3 Control room Unit 1 intake (0-2 hrs) 1.98 x 10'3 Unit 2 intake (0-2 hrs) 9.52 x 10

. Control room parameters 3

Volume (ft )

81,420 Makeup flow (cfm) 2,200 Makeup and recirculation flow (cfm) j One fan operational 5,000 Two fan operational 12,200 Makeup and recirculation filter efficiency (%)

elemental, organic iodines 95 particulate iodine 99 Unfiltered inleakage (cfm) 60 l

i TABLE 2 CALCULATED DOSES FOR SALEM UNIT 1 AND 2 LOSS-OF-COOLANT ACCIDENT LOCATION Thyroid Dose Whole Body Exclusion Area Boundary 42.0 rem

(EAB)

Low Population Zone 15.0 rem

(LPZ)

Control Room Operators 27.0 rem ***

1.7 rem ****

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. 4 TABLE 3 4

I INMIT PARAMETERS FOR SALEM UNIT 1 Ale 2 EVALUATION 0F A FUEL HAE LING ACCIDENT Power Level (MWt) 3600 Number of Fuel Rods Damaged 204 Total number of Fuel Rods 39,372 Power Peaking Factor 1.7 Fission Product Release Duration 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Release Fraction

Noble gas 10%

Krypton gas 30%

Iodine Forms 5

Elemental 75%

Organic 25%

Receptor Point Variables Exclusion Area Boundarv 3

3' Atmospheric Relative Concentration, X/Q (sec/m )

0-2 hours 2.4 x 10

4 Control Room l

Atmospheric Relative Concentration, X/Q (sec/m )

3 Unit 1 intake 1.98 x10'3 Unit 2 intake 9.52 x10

3 Control Room volume cubic feet 8.1 x 10

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Filter Recirculation F1nw (cfm) 5,000 Unfiltered inleakage (cfm) 60 4

Pressurization Air Filtration (cfm) 2200 Iodine Protection Factor (IPF) 41

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TABLE 4 CALCULATED DOSES FOR SALEM UNIT 1 AND 2 FUEL HANDLING ACCIDENT LOCATION Thyroid Dose Whole Body Dose Exclusion Area Boundary 47.6 rem *

> 0.1 rem **

Control Room Doses 5.6 rem ***

> 0.1 rem ****

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        • GDC-19 Acceptance Criteria Whole Body = 5 rem i

1 TABLE 5 INPUT PARAMETERS FOR SALEM UNITS 1 AND 2 EVALUATION OF A LOCKED ROTOR ACCIDENT Steam Releases (1bs) i 0-2 hr 654,600 l

2-8 hr 540,300 j

8-32 hr 2,161,200 Primary to Secondary Leak Rate (gpm) 1 Primary Coolant Mass (1bs) 493,000 Steam Generator Liquid Mass (1bs) 106,860 Iodine Fraction after the accident (%)

5 Control Room Parameters 3

Atmospheric Dispersian X/Q (s/m )

Unit 1 intake (0-2 hrs) 1.94 x 10'3 Unit 2 intake (0-2 hrs) 3.45 x 10'3

s TABLE 6 CALCULATED DOSES FOR SALEN UNIT 1 AND 2 LOCKED ROTOR ACCIDENT I

LOCATIDN Thyroid Whole Body Exclusion Area Boundary 0.29 rem

  • 0.05**

Low Population Zone 0.51 rem

  • 0.04**

Control Room Doses 21 rem ***

1.7 rem ****

        • GDC-19 Acceptance Criteria Whole Body - 5 rem l

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TABLE 7 INPUT PARAMETERS FOR SALEM UNITS 1 AND 2 EVALUATION OF A STEAM GENERATOR TUBE RUPTURE ACCIDENT i

Power level (MWt) 3600 Primary coolant concentration of dose equivalent '3'I Pre-existino Spike Value (uti) 1*'I = 31.9 1321 = 63.0

'33

'3'1 = 71.6 I - 83.6

'33 1 = 64.5 j

l Volume of primary coolant Primary coolant mass (1bs) 493,000 j

TS limits for dose equivglent (DE) '3'I in the primary and secondary coolant Primary coolant DE concentration (pci/g) 1.0 Secondary coolant DE'3{I concentration (pCi/g) 0.1 Post-accident steam generator liquid mass (1bs/per SG) 106,860 Steam releases (1bs)

Faulted SG:

0-55 minutes 56,460 Intact SG:

0- 2 hrs 465,130 Release gte for IgCi/g of dose equivalent

'3' I (gCi/g)

I - 0.07

'32 1 = 0.12

'33

'3'I = 0.16 I - 0.19

'33I = 0.14 3

Atmospheric dispersion factor (s/m )

Exclusion Area Boundary (0-2 hrs) 2.4 x 10

Low population zone (0-8 hrs) 2.2 x 10'5 Control Room 1.8 x 10'3 Control room pa[)ameters Volume (ft 81,420 Makeup flow (cfm) 2,200 Recirculation flow (cfm) 12,200 Makeup and recirculation filter efficiency (%)

95 Unfiltered inleakage (cfm) 60 Occupancy factor I

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i TABLE 8 CALCULATED DOSES FOR SALEM UNITS 1 AND 2 STEAM GENERATOR TUBE RUPTURE ACCIDENT Thyroid Thyroid Whole Body LOCATION Pre-accident Accident Spike Initiated Spike Exclusion 2.40 rem

  • 0.53***

0.03 rem **

Area Boundary Low 0.66 rem

0.01 rem ****

Population Zone Control Room 3.5 rem *****

2.57 rem *****

>.0lrem******

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            • GDC-19 Acceptance Criteria Whole Body-5 rem 2.5 The licensee proposed to rewrite the Limiting Londition for Operation (LCO) in TS 3.7.6.1 as follows:

The common control room emergency air condition system (CREACS)* shall be operable with:

a. Two independent air conditioning filtration trains (one from each unit) each consisting of:

1.

Two fans and associated outlet dampers, 2.

One cooling coil, i

3.

One charcoal absorber and HEPA filter array, 4.

One return air isolation damper.

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b. All other automatic dampers required for operation in the pressurization or recirculation modes.
c. The Control Room Envelope intact.

4 The proposed LC0 reflects the CREACS design changes. The licensee added an asterisk (*) on the CREACS to clarify that the CREACS is a shared system with both plant units.

Each unit's CREACS fans are replaced with two redundant 100% capacity fans, such that any one of the four fans is capable of providing 100% of the required CRE pressurization air to 1/8" water gauge (w.g.)

positive pressure within the CRE.

There are dampers installed on the makeup

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air supply and return ducts with failure positions consistent with isolation of the CRE. The CRE is considered intact when the CREACS is capable of maintaining a pressurized CRE with all boundary doors closed.

The proposed LCC will protect the CR operators against any hazardous conditions by either pressurizing or recirculating the CRE with the redundant systems. The design changes are based on the licensee's revised design bases j

which will restore the CR operation to satisfy GDC 19. The proposed LCO has corrected the plant design deficiency and improved operation safety and is, therefore, acceptable.

2.6 The licensee proposed to revise the Applicability Statement by adding "during movement of irradiated fuel assemblies and during core alterations". The current LC0 is applicable to Modes 1 thru 6.

The change is required to maintain the control room integrity during fuel handling operation and is consistent with the Westinghouse Standard Technical Specifications (STS).

2.7 The licensee proposed to rewrite the Action Statement as follows:

Modes 1, 2, 3, and 4

a. With one filtration train inoperable, align CREACS for single filtration train operation within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, and restore the inoperable filtration train to OPERABLE status within 30 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
b. With CREACS aligned for single filtration train operation and with one of the two remaining fans or associated outlet dampers inoperable, restore 1

the inoperable fan or damper to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

c. With the control room envelope inoperable, restore the control room envelope to OPERABLE status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
d. With one or both series isolation damper (s) on a normal CAACS outside air intake or exhaust duct inoperable, close the affected duct within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least one isolation damper secured in the closed position or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. (Refer to ACTION 24 [27 for Unit 2] of Table 3.3-6.)

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e e. With one or both isolation damper (s) on an outside emergency air conditioning air intake duct inoperable, close the affected duct within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least one isolation damper securod in the closed

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position and restore the damper (s) to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

f. With any isolation damper between the normal CAACS and the CREACS inoperable, secure the damper in the closed position within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

With one filtration train inoperable, the Action Statement requires the inoperable filtration train to be restored to operable status within 30 days, which is not specified in the current TS. The licensee stated that with a single filtration train in operation two fans will be available to start in the pressurization mode or recirculation mode.

Should a fan fail to start on an initiation signal in the train with two fans operable, the second fan in standby will automatically start.

Since any one of the four fans is sized to provide the required air flow f >r maintaining the CRE pressurization at 1/8" w.g. during an emergency, the required redundancy for the system has been preserved with single train operation.

The staff's review cor.cludes that the 30-day outage time is accept 41e because the redundant train is not required to be in operation for the Action Statement and the time required for restoring the inoperable filtration train is consistent with the TS requirements for other similar redundant systems.

With one of the two remaining fans or associated outlet dampers inoperable during the single train operation, C.e Action Statement requires restoration of the inoperable fan or damper t:, operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

Each CREACS train has two fans, te outlet dampers, and one air return isolation damper (which will fail in the open position upon loss of control power). The air return isolation damper will be administrative 1y controlled in the open position when operating in single train alignment. The CREACS is designed with the capability of maintaining the CRE pressurization or recirculation with one fan in operation. The staff finds that the 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for restoring

'3 the inoperable fan is acceptable because it is less than the current TS requirement of 7 days and is consistent with TSs for similar redundant systems.

Completion times for other Action Statements are consistent with the current TSs and are considered reasonable for its field operation practices and are, 4

therefore, acceptable.

MODES 5 and 6 or during movement of irradiated fuel assemblies and during core alterations.

a. With one filtration train inoperable, align CREACS for single filtration train operation within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, or suspend core alterations and movement of irradiated fuel assemblies.

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> b. With CREACS aligned for single filtration train operation with one of the two remaining fans or associated outlet damper inoperable, restore the fan or damper to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, or suspend core alterations and movement of irradiated fuel assemblies.

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c. With two filtration trains inoperable, immediately suspend core l

alterations and movement of irradiated fuel assemblies.

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d. With the Control Room Envelope inoperable, immediately suspend core alterations and movement of irradiated fuel assemblies.

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e. With one or both series isolation damper (s) on a normal CAACS outside air intake or exhaust duct inoperable, immediately suspend core alterations and movement of irradiated fuel assemblies until the affected duct is closed by use of at least one isolation damper secured in the closed position.

(Refer to ACTION 24 {27 for Unit 2) of Table 3.3-6.)

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f. With one or both series isolation damper (s) on an outside emergency air conditioning air intake duct inoperable, immediately suspend core alterations and movement of irradiated fuel assemblies until the affected duct is closed by use of at least one isolation damper _ secured in the closed position. To resume core alterations or movement of irradiated fuel assemblies, at least one emergency air intake duct must be operable in each unit.

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g. With any isolation damper between the CAACS and the CREACS inoperable, immediately suspend core alterations and movement of irradiated fuel assemblies until the damper is closed and secured in the closed position.

l With one filtration train inoperable during Modes 5 and 6, the Action Statements require the system to be aligned for single filtration train i

operation or that movement of irradiated fuel assemblies or core alterations be suspended if the Action Statements are not satisfied within the specified time frame. The licensee stated that the allowed outage times in the Action i

Statement were specified for inoperable filtration trains and dampers based on 4

safety significance and are consistent with the STS.

With one or both isolation dampers on an outside emergency air conditioning j

air intake duct inoperable, the Action Statements require immediate suspension of core alterations and movement of irradiate ,

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5.0 CONCLUSION

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The Commission has concluded, based on the considerations discussed above, that:

(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common i

defense and security or to the health and safety of the public.

1 Principal Contributors:

J. Minns J. Guo l

Date: February 6, 1997 i

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