ML20206G864

Application for Amends to Licenses NPF-4 & NPF-7,increasing Allowable Leakage from ECCS Components,Establishing Consistent Licensing Basis for Accidents Requiring Dose Analysis & Clarifying Requirements of CREVS

ML20206G864
Person / Time
Site:	North Anna
Issue date:	05/03/1999
From:	Christian D VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20206G866	List: ML20206G864 ML20206G881
References
99-264, NUDOCS 9905100166
Download: ML20206G864 (41)

Text

s s VIRGINIA ELECTRIC ANI> Powicu CONH%NY RialssONI), VIRGINI A 2326I May 3, 1999 U.S. Nuclear Regulatory Commission Serial No.99-264 Attention: Document Control Desk NL&OS/GSS R1 Washington, D.C. 20555 Docket Nos. 50-338 50-339 License Nes. NPF-4 NPF-7  !

i Gentlemen:

VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA POWER STATION UNITS 1 AND 2 PROPOSED TECHNICAL SPECIFICATION CHANGE I CONTROL ROOM EMERGENCY HABITABILITY Pursuant to 10 CFR 50.90, Virginia Electric and Power Company requests amendments, in the form of changes to the Technical Specifications for Facility Operating License Numbers NPF-4 and NPF-7 for North Anna Power Station Units 1 and 2, respectively. The proposed changes support an increase in the allowable leakage from ECCS components, establish a consistent licensing basis for accidents requiring a dose analysis, and clarify the requirements of the control room emergency ventilation system. A discussion of the proposed Technical Specifications changes is provided in Attachment 1.

The proposed Technical Specifications changes have been reviewed and approved by the Station Nuclear Safety and Operating Committee and the Management Safety Review Committee. Based on the small increase in dose consequences due to the i increased allowable ECCS leakage, it has been determined that the proposed Technical Specifications changes do involve an unreviewed safety question as defined f in 10 CFR 50.59 but do not involve a significant hazards consideration as defined in 10 CFR 50.92.

The proposed Technical Specifications changes are provided as a mark-up in Attachment 2 and as a typed version in Attachment 3. The basis for our determination 0

N that the changes do not involve a significant hazards consideration is provided in 0

Attachment 4. p i

9905100166 990503 i PDR ADOCK 05000338 P PDR l

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l If you have any further questions, please contact us.

Very truly yours, l 1

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- i D. A. Christian l Vice President- Nuclear Operations Attachments 1

1. Discussion of Changes )

2. Mark-up of Technical Specifications Changes

3. Proposed Technical Specifications Changes

4. Significant Hazards Consideration Determination 1

1 Commitments made in this letter:

1. There are no commitments in this letter cc: U.S. Nuclear Regulatory Commission Region il Atlanta Federal Center 61 Forsyth Street, SW Suite 23T85 Atlanta, Georgia 30303 l

Mr. M. J. Morgan NRC Senior Resident inspector North Anna Power Station Commissioner Department of Radiological Health Room 104A 1500 East Main Street '

Richmond,VA 23219 Mr. J. E. Reasor Old Dominion Electric Cooperative Innsbrook Corporate Center 4201 Dominion Blvd.

Glen Allen, Virginia 23060 t

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l'-'l COMMONWEALTH OF VIRGINIA- )

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. COUNTY OF HENRICO - )

The foregoing docurneni was acknowledged before me, in and for.the County and Commonwealth aforesaid, today by D. A. Christian, who is Vice President -

Nuclear Operations, of Virginia Electric and Power Company. He has affirmed before me that he is duly authorized to execute and file the foregoing document in behalf of that Company, and that the statements in the document are true to the best of his knowledge and belief.

Acknowledged before me this) day of bb[ .19 .

My Commission Expires: March 31,2000. 4 1

1 05 ( U1A

'_ / Notary Public l

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1 Attachment 1 l Discussion of Change l l

North Anna Power Station j Units 1 and 2 '

Virginia Electric and Power Company

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INTRODUCTIOJ Pursuant to 10 CFR 50.90, Virginia Electric and Power Company (Virginia Power) requests changes to the Technical Specifications for the Control Room Habitability System for North Anna Units 1 and 2. The proposed changes will revise the Technical Specifications to ensure the emergency ventilation system is maintained operable consistent with the assumptions in the radiological dose consequence reanalysis from a Large Break Loss of Coolant Accident ( LOCA) and to clearly identify that the ventilation system is a shared system between the two units.

Changes are also being made to the Technical Requirements Manual by adding a new section, "ECCS Ixakage Limits." This section will establish the Technical Requirements and Action Statements for addressing ECCS leakage.

The dose consequences of a LOCA have been reanalyzed to support a proposed increase in the allowable leakage from emergency core cooling systems (ECCS) components from 900 cc/hr to 4800 cc/hr. This reanalysis was performed using the LOCADOSE code. The offsite and control room dose consequences of the Main Steam Line Break Accident (MSLB), the Steam Generator Tube Rupture Accident (SGTR), and the Locked Rotor Accident (LRA) have also been reanalyzed using the LOCADOSE code. The purpose of these re-analyses is to establish consistent technical and licensing bases for accidents that require a dose analysis. A new section j to the Technical Requirements Manual will also be added to establish guidance for addressing j ECCS allowable leakage. l l

The dose consequences of the LOCA, MSLB, SGTR, and LRA accidents were determined to meet the 10 CFR 100 limits and the limits specified in GDC-19. While the radiation doses meet the required design limits, the projected control room thyroid dose for the LOCA, calculated under the conditions specified by the revised Technical Specifications, has increased above those calculated in the current analysis of record. In addition, some of the exclusion area boundary (EAB) and/or low population zone (LPZ) doses for certain accidents exceed those values l reported in the UFSAR. Based on this fact, and the fact that the some of the key analysis assumptions and methods have changed, it was determined that an unreviewed safety question exists as defined in 10 CFR 50.59. However, since the dose consequences continue to meet the regulatory limits, neither the new dose analyses nor the proposed changes to the Technical Specifications constitute a significant hazards considerction as defined in 10 CFR 50.92.

BACKGROUND ON CONTROL ROOM HABITABILITY SYSTEM Current Licensing and Design Basis On March 1,1989, Virginia Electric and Power Company requested license amendments that incorporated the results of control room habitability reanalyses based on proposed plant modifications and interim operator actions into the facility licenses (3). These reanalyses modified existing analyses to address the impact of the control room door openings required by the emergency operating procedures during accidents. These 1989 analyses also assumed that two fan / filter units provided recirculation of the control room air for the duration of the accident, and that one additional fan / filter unit provided a filtered intake supply after the bottled air supply was exhausted. The NRC approved those licenses amendments on February 28,1990 (15).

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In May of 1990, modifications to the North Anna control room habitability systems were implemented to relieve the operators from having to perform manual actions to properly actuate the systems in the event of certain accidents. The modifications included the following:

l e. The control circuits for the Unit I and Unit 2 bottled air system were modified to provide for the automatic discharge of bottled air upon receipt of a HI-HI radiation signal in either i the new fuel storage area or fuel bridge area of the fuel building. This automatic

! discharge can be defeated from a control switch and is intended to be active only when fuel handling operations are being performed. The previous circuitry would discharge bottled air only upon a safety injection or manual actuation signal. r

. The normal control room ventilation control circuits were modified to isolate the control room whenever the bottled air discharge was initiated. This was accomplished by l automatically closing the air operated intake and exhaust dampers.

• The control circuits for the control room and switchgear room emergency ventilation fans l

were modified to automatically start the fans whenever the bottled air discharge was initiated. The fans are procedurally controlled to startup in the recirculation mode.

l On October 17, 1995 Virginia Power requested changes to the North Anna Technical l Specifications to allow the containment personnel airlock to be open during refueling (16). To l support this change request, the Fuel Handling Accident was reanalyzed assuming a full l unfiltered release from the containment or fuel building. This reanalysis of the Fuel Handling l

Accident, which did not take credit for any recirculation of the control room air during the )

accident, evaluated both the offsite and the control room doses. This analysis of record for the '

Fuel Handling Accident was performed with the LOCADOSE code, and was approved by the NRC in Reference (2).

l l During normal operation, the control room air for breathing and pressurization is supplied by air

! handler 1-HV-AC-4, and the control room exhaust is continuously withdrawn and discharged to the atmosphere. An air conditioning system is also provided for recirculation cooling of the control room. In addition to the normal inlet air to the control room, air for breathing and 1 pressurization could be supplied by the bottled air system and emergency ventilation system. The emergency ventilation system which consists of four fans and filters are designed to start the fans in the recirculation configuration whenever a bottled air system is initiated.

In the event of an emergency, the normal outside air supply and exhaust are automatically isolated on a safety injection (SI) signal either from Train A or Train B from either Unit 1 or Unit 2, a Hi-Hi radiation condition in the fuel building during fuel handling operations, or manually in the control room. Following a design basis accident (DBA) the control room envelope is isolated and pressurized for the first hour by the bottled air system, and the emergency ventilation system is aligned to recirculate the air within the control room envelope. After the first hour, one emergency ventilation fan will be aligned to ensure control room pressurization. The emergency ventilation fan will supply air from the Turbine Building through HEPA and charcoal filters for the duration of the accident.

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Discussion of Reanalysis The Locked Rotor accident, although not originally included among those accidents for which control room doses were addressed, has been reanalyzed to ensure that all potentially limiting (Condition IV) accidents have been evaluated for their impact on control room habitability. The only Condition IV accident discussed in Chapter 15 of the North Anna UFSAR that is not specifically addressed in the offsite and control room habitability calculations is the RCCA ejection accident. For this accident, it has been determined that less that 10% of the fuel will fail (Section 15.4.6.2.3.5 of the North Anna UFSAR). The consequences of the limited amount of fission product release for the rod ejection accident are bounded by the analysis for the LOCA.

Large Break Loss of Coolant Accident (LOCA) have been reanalyzed to support a proposed increase in the amount of allowable leakage from ECCS components from 900 cc/hr to 4800 l cc/hr. This analysis was performed with the LOCADOSE code in a manner consistent with i applicable sections of the Standard Review Plan (1). In addition, the dose consequences resulting from the following accidents were evaluated using the LOCADOSE code in order to establish a consistent technical and licensing basis for the North Anna dose analyses:

. Major Secondary Steam Pipe Rupture (i.e., Main Steam Line Break, MSLB) e Steam Generator Tube Rupmre (SGTR) j e Locked Reactor Coolant Pump Rotor Accident (i.e., LRA)

To minimize doses to control room personnel during accidents at the North Anna Power Station, the control room envelope is pressurized (upon receipt of a control room isolation signal) to minimize in-leakage of airborne radioactive material. This pressurization is initially provided by a bottled air system, which has sufficient capacity to supply the control room for 60 minutes.

When the bottled air supply is depleted, breathing and pressurization air is provided by an emergency filtered air system which draws air from the turbine building through HEPA and charcoal filters. The emergency ventilation system maintains the control room envelope at a positive pressure relative to the outside atmosphere.

The analysis methods, assumptions, and results for the accidents discussed above are described below in the Technical and Safety Evaluation section. Compared with our 1989 submittal (3)

(4), the North Anna control room habitability calculations (reanalysis) assume no recirculation of the control room air during the accidents. As a result, changes to the control room ventilation system Technical Specifications are being proposed to make the Technical Specifications consistent with the analyses and to clearly identify that the ventilation system is a shared system between the two units.

PROPOSED TECHNICAL SPECIFICATIONS CHANGES The proposed Technical Specifications changes for North Anna Units 1 and 2, included in l Attachments 2 and 3 respectively, formalize the operability and surveillance requirements of the

} control room emergency habitability systems consistent with the revised control room dose l calculations: Technical Specification 3.7.7.1 for each unit is being revised to indicate that the emergency ventilation is shared with the other unit.

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1. Technical Specification 3.7.7.1 for each unit is being revised to indicate that the bottled air system is shared with the other unit.

- 2. A statement is being added to the . APPLICABILITY discussion for Technical Specification 3.7.7.1 to indicate that Specification 3.7.7.1 is applicable during movement of irradiated fuel assemblies or during CORE ALTERATIONS.

3. An ACTION statement is being added to Technical Specification 3.7.7.1 that indicates that, with either the emergency ventilation system or bottled air pressurization system inoperable, suspend movement ofirradiated fuel assemblies or CORE ALTERATIONS.

4. A statement is being added to BASES Section 3/4.7.7 to define an operable control room l emergency ventilation system as follows: An OPERABLE control room ventilation consists i of at least two operable emergency filtration trains out of the four emergency filtration trains serving the combined Unit I and Unit 2 control room. An operable emergency filtration train shall include an operable fan, an operable charcoal filter an operable HEPA filter, an operable flow path, and normal and emergency electrical power available. Note that limiting-condition-of-operation (LCO) 3.0.5 may only be used if the normal and emergency electrical power sources are required by LCO 3.8.1.1 and 3.8.1.2.

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! SAFETY SIGNIFICANCE

SUMMARY

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The dose consequences of the LOCA, MSLB, SGTR, and LRA accidents were determined to meet the 10 CFR 100 limits and the limits specified in GDC-19. While the radiation doses meet j the required design limits, the projected control room thyroid dose for the LOCA, calculated under the conditions specified by the revised Technical Specifications, has increased above those calculated in the current analysis of record. In addition, some of the EAB and/or LPZ doses for certain of the accidents exceed those values reported in the UFSAR. Based on this fact, and the j fact that the some of the key analysis assumptions and methods have changed, it was determined that an unreviewed safety question exists as defined in 10 CFR 50.59. However, since the dose consequences continue to meet the regulatory limits, neither the new dose analyses nor the proposed changes to the Technical Specifications constitute a significant hazards consideration as defined in 10 CFR 50.92.

l TECHNICAL AND SAFETY EVALUATION Virginia Power has re-analyzed the offsite and control room dose consequences for the following postulated accidents at North Anna:

1. Large Break Loss of Coolant Accident (LOCA)

< 2. Major Secondary Steam Pipe Rupture (i.e., Main Steam Line Break, MSLB) 1

3. Steam Generator Tube Rupture (SGTR) 4._ Locked Reactor Coolant Pump Rotor Accident (i.e., LRA)

The LOCA was re-analyzed to increase the amount of allowable leakage from Emergency Core Cooling System (ECCS) components from 900 cc/hr to 4800 cc/hr. This new analysis was performed using the LOCADOSE code. The remaining accidents were also analyzed with the LOCADOSE code (5) (6) (7) using methodologies consistent with applicable sections of the Standard Review Plan (1) in order to establish a consistent technical and licensing bases for ,

accidents which require dose analyses.

1.0 Models, Input Data, and Analyses 1.1 Loss of Coolant Accident (LOCA)

The methodology used to evaluate the control room and offsite doses resulting from a LOCA is consistent with Section 15.6.5 of NUREG-0800 (1) and Regulatory Guide 1.4 (8).

1.1.1 LOCA Analysis Approach The following approach was used to calculate the offsite and control room dose consequences resulting from a LOCA:

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1. Doses from all postulated release paths to the environment were calculated and compared  !

with 10 CFR 100 and GDC 19 exposure guidelines. Radiological consequences of both containment leakage and post-LOCA leakage from ECCS components outside containment 6

were considered.

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! 2. A power level of 2958 MWt was assumed for all North Anna dose analyses. This represents l_ a'slightly conservative value of 2900 MWt for the core power level for North Anna plus 2%

for instrument uncertainty.' The core inventory of radionuclides for North Anna was determined from the Ci/MWt values in the LOCADOSE computer code.

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3. At the time of the accident,25% of all equilibrium iodine fission products and 100% of the noble gas fission products were assumed available for release from the containment within a very short time (effectively instantaneously) after the accident. The iodine was assumed to be composed of 91% elemental iodine,4% organic iodides, and 5% particulate iodine.

4. The containment leak rate used in the LOCA analysis was 0.1% per day of the containment volume for first hour, when the containment pressure was above atmospheric, and 0%

thereafter. North Anna has a sub-atmospheric containment system that has been shown by analysis to return to sub-atmospheric within one hour of the start of a LOCA. When the containment pressure is sub-atmospheric, any leakage would be into the containment.

5. North Anna does not have a vent / purge system that had to be considered as a LOCA release pathway.

6. The following containment spray factors were used in this analysis:

Elemental lodine (1) = 10/ hour Organic Iodine (A) = 0/ hour Particulate lodine (A) = 0/ hour Sprayed Volume = 70%

Mixing Rate = 2 unsprayed volumes / hour

7. A radionuclide filter efficiency of 90% for leakage of iodine from the ECCS components in i

the safeguards area was used in the North Anna LOCA analysis. All leakage from the r safeguards area is assumed to be elemental, so the 90% filter efficiency applies to all iodine i releases from the ECCS system. For the purpose of this analysis, any iodine in the ECCS leakage that became airbome during the first hour of the accident was assumed to be unfiltered. This assumption was made to account for the operator action time required to line up the exhaust from the charging pump cubicles to the filter banks.

8. Based on a review of the Basis for North Anna Technical Specification 3/4.7.7 and Regulatory Guide 1.52, a control room ventilation system filter efficiency of 95% was determined to be appropriate for the removal of elemental iodine. A filter efficiency of 99%

was assumed for the removal of particulate iodine based on the use of HEPA filters.

9. The appropriate x/Q values used in these calculations are discussed below in Section 1.1.2.

The distances to the exclusion area boundary and to the LPZ outer boundary were taken from Sections 2.1.2 and 2.1.3 of the North Anna UFSAR.

10. The LOCADOSE models used in the LOCA analysis are discussed below in Section 1.1.3.

. I1. The total allowable leakage from ECCS components was increased from 900 cc/hr to 4800 Page 6 of 37

cc/hr in this analysis. The leakage modeled in the LOCADOSE mn was twice the allowable i

leakage of 4800 cc/hr. A passiva failure of the charging pump seal was not assumed in this analysis since the charging pump cubicle is capable of being filtered by an ESF filtration system.

12.The radiological consequences resulting from the postulated leakage from the ECCS components were based on the assumption that 50% of the core iodine inventory was mixed with the sump water.

13.Except for a short period (10 minutes) at the beginning of the recirculation phase of the accident, the water in the ECCS system at North Anna is taken from the containment sump at temperatures less than 212 F. When the water temperature in the sump exceeds 212 F, the flashing fraction is less than 10%. Therefore, for the purposes of this LOCA analysis,10% of the iodine in the ESF system leakage was assumed to become airborne.

1.1.2 De*ermination of x/Q Values 1.1.2.1 Determination of Control Room x/Q Values The control room x/Q values used in the LOCA analysis for the ECCS leakage were calculated using the ARCON96 Code and are presented in Table 1.1-1. The control room r/Q values for the containment leakage were calculated using the methodology outlined in Murphy and Campe (12) and are listed in Table 1.1-1. It should be noted that the control room occupancy factors were not included in any of the x/Q values listed in Table 1.1-1 because these occupancy factors, which are listed in Table 1.1-2, were explicitly inputted into the LOCADOSE computer code.

1.1.2.2 EAB and LPZ x/Q Values and Occupancy Factors The EAB x/Q value used in this LOCA analysis was 3.lE-4 sec/m3. This value has been used at I North Anna since the PSAR LOCA analysis was submitted.

The LPZ x/Q values used for this LOCA analysis were taken from the original analysis (11). The x/Q values were taken from a graph of x/Q as a function of distance found in Regulatory Guide 1.4 (8). The distance used to determine x/Q was 6 miles. A review of the original calculation revealed that the graphs had been read conservatively. Use of the original LPZ x/Q values remains conservative.

1.1.3 LOCADOSEModelforNorth Anna LOCA Analysis LOCADOSE was used to model the release of radionuclides for a LOCA at North Anna.

LOCADOSE first calculates radionuclide concentrations and releases to the environment. These radionuclide releases and concentrations are then used along with breathing rates, and occupancy factors to calculate the resulting doses. The LOCADOSE computer code system modeled a LOCA at North Anna with five volumes: 1) the environment, 2) the containment sump,3) the portion of the containment covered by the containment chemical spray system,4) the portion of i the containment not covered by the chemical spray system, and 5) the control room. The volumes used in the computer model were:

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Unsprayed containment volume .= 5.520 E+5 cubic ft Sprayed containment volume = 1.288 E+6 cubic ft Sump volume = 1.500 E+6 liters Control room volume = ~ 2.30 E+5 cubic ft The transfer of radionuclides was modeled by specifying flow rates between the various modeled volumes. The mixing between the sprayed and unsprayed containment volumes was modeled

. based on 2 unsprayed volumes per hour. The containment leakage was modeled as 0.1 volume percent per day for the first hour. After the first hour, no containment leakage was assumed because the containment pressure returns to sub-atmospheric.The 0.1 volume percent per day

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was applied equally to the sprayed and unsprayed volumes. Thus, 0.1 volume percent of each of these volumes was modeled as leaking to the environment during the first hour of the LOCA.

The ECCS leakage (ESF leakage) is modeled at twice the maximum leakage rate allowed:

Time Period Maximum Leak Rate Allowed Twice Allowable Leakage (cc/hr) (liters / min) 0 min. - 30 days 4800 1.60E-1 The control room ventilation system was modeled in a manner consistent with North Anna control room ventilation system design and operation. An unfiltered inleakage of 10 CFM was modeled from time 0 to 30 days. The control room was assumed to be on bottled air for the first hour of the accident. After the first hour, a filtered intake of 1000 CFM along with the 10 CFM of unfiltered in-leakage was modeled. Although the control room is designed with filtered tcirculation flow, the recirculation was not modeled.

The core radionuclide inventory was modeled as initially being distributed as follows:

Containment Unsprayed Volume: 30% of core noble gas isotopes, and 30% of 25% of core iodines with a distribution of 91% Elemental,5%

Particulate,4% Organic.

Containment Sprayed Volume: 70% of core noble gas isotopes, and 70% of 25% of core iodines with a distribution of 91% Elemental,5%

Particulate,4% Organic.

Sump Volume: C0% of core inventory of iodines,10% airborne, with 100% of airborne iodine being ele, ental.

This modeling is consistent with the requirements of SRP 15.6.5, Appendices A and B and assumes instantaneous release of the radionuclides from the reactor core.

The LOCADOSE code system calculates radio aclide releases to the environment, and radionuclide concentrations versus time in each volume. The time-dependent radionuclide concentrations are then used along with the occr ncy factors and breathing rates to calculate doses. The occupancy factors (fraction of time an individual occupies the control room) assumed Page 8'of 37 o

TL 1

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for the control room are shown in Table 1.1-2. These occupancy factors are based on the guidance from the Murphy and Campe paper (12). The breathing rate used for the control room 3

dose calculations was 3.47 E-4 m /sec, which is consistent with the Murphy and Campe paper.

The LOCADOSE code ICRP-30 dose conversion factors were used to determine the doses from inhalation and immersion.

i 1.1.4 Results of Dose Calculations for LOCA j The LOCA models described above were used to analyze doses in the control room and offsite and the resulting control room and offsite doses are given in Table 1.1-3. The calculated doses are less than the 10 CFR 100 limits for the EAB and LPZ, and are less than the GDC-19 and SRP 6.4 limits for the control room.  ;

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Table 1.1-1 North Anna Containment and ECCS x/Q Time Period Control Room x/Q EAB x/Q LPZ x/Q 3 3 3 (hours) (sec/m ) (sec/m ) (sec/m )

CONTAINMENT 0- 2 4.35 x 10-3 3.10 x 10-4 1.10 x 10-5 2- 8 4.35 x 10-3 -

1.10 x 10-5 8 - 24 2.74 x 10-3 -

7.30 x 10-6 24 - 96 1.82 x 10-3 -

3.00 x 10-6 96 - 720 8.09 x 10-4 -

8.20 x 10-7 ECCS LEAKAGE O- 2 7.43 x 10-3 3.10 x 10-4 1.10 x 10-5 2- 8 4.73 x 10-3 -

1.10 x 10-5 8 - 24 2.04 x 10-3 -

7.30 x 10-6 24 - 96 1.28 x 10-3 -

3.00 x 10-6 96 - 720 9.84 x 10-4 -

8.20 x 10-7 Table 1.1-2 North Anna Control Room Occupancy Factors Time (hours) Occupancy Factor 0- 8 hr 1.0 8- 24 hr 1.0 24 - 96 hr 0.6 96 - 720 hr 0.4 i

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l Table 1.1-3 LOCA Control Room and Off Site Doses Control Room GDC-19 EAB 2-hour LPZ 30-Day 10 CFR 100 30-Day Dose Dose Limit

• Dose Dose Limit (Rem) (Rem) (Rem). (Rem) (Rem)

Thyroid 27.8 30 72.5 3.1 300 l l

Skin <l .0 30 3.5 <l .0 -- l Whole Body'* 1.1 5 1.6 <l .0 25

• Control room skin dose limit is not specified in GDC-19; limit shown is taken from the SRP Section 6.4.

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• The whole body dose as calculated by LOCADOSE code is dose from the external exposure I only. For the control room, the whole body dose contribution due to the shine from the containment and from the cloud of radionuclides surrounding the control room is added to the whole body dose calculated by the LOCADOSE code. The LOCADOSE calculation i represents the doses from the iodine and noble gases that enter into the control room. The i contribution from the direct shine was previously calculated in the FS AR to be 0.748 Rem. ;

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i 1.2 MAIN STEAM LINE BREAK (MSLB)

A Main Steam Line Break (MSLB) involves the double-ended failure of one of the steam lines l carrying steam from a steam generator to the turbine generator. To maximize the radiological impact of the accident, it was assumed that the break in the steam line occurred in the turbine building. Two cases were considered. Off-site doses were determined based on a case with minimal retention of radionuclides in the turbine building. This case assumed the turbine

building fans were operating. The control room dose analysis assumed that the turbine building ventilation fans fail to operate. This allowed the radionuclides concentration in the turbine building to reach a higher level than the case in which the fans operate. Since the emergency control room air inlets are in the turbine building, this assumption maximized the control room dose.

Because the MSLB releases were assumed to occur in the turbine building, the normal x/Q methodology used for the control room did not apply. x/Q is used to determine the concentration 3

of a radioisotope X per m from the release rate Q in Ci/sec. The control room x/Q values are normally determined with the ARCON96 Code (9) based on the source-receptor geometry, release characteristics and site meteorology. However, for the MSLB, the releases were assumed to occur in the same building as the control room emergency inlets, so the ARCON96 x/Q I methodology did not apply. Therefore, the direct pathway from the steam line break to the turbine building was modeled along with the intake of control room air from the turbine building.

1.2.1 MSLB Analysis Assumptions As indicated above, analysis of doses in the control room from a MSLB could not be performed solely with the normal x/Q methodology. There is no control room x/Q defined for a situation where the releases are into the same building where the inlet to the control room is located.

Therefore, it was necessary to use a different approach to model the transport of radioactive steam releases from the broken steam line to the control room. (The normal x/Q methodology is ,

l applicable to the modeling of releases through unaffected steam generators.)

The control room has normally been modeled in the LOCADOSE computer code as a special volume " connected" only to the environment with the inlet concentrations based on releases to the environment and the x/Q for the control room. However for the MSLB case, the control room radioisotope concentrations were calculated with LOCADOSE by defining one of the user-specified volumes as the control room, and appropriately modeling the air flows, including the inlet air from the turbine building, to this control room volume.

As a starting point for the MSLB analysis, the concentrations of each radioisotope in the primary liquid, secondary liquid, and secondary steam were determined as discussed in Section 1.2.2.

Radionuclides were released with the steam from these sources through the break.

The flow rates used in this analysis considered the volume expansion that occurs when pressurized liquid or steam is discharged from the steam generator to the turbine building. The i l

flow rate from the steam generator to the turbine building is based on the density of steam or liquid inside the steam generator, while the flow rate from the turbine building to the Page 12 of 37 L

1 environment is based on the expansion of steam'to atmospheric pressure inside the turbine building. This MSLB model is summarized below.

l 1.2.2 -Initial Radioisotope Concentrations l

! For the MSLB, the radioactive material releases were determined by the initial radionuclide

- concentrations present in_ primary liquid, secondary liquid, and secondary steam, plus any l releases from failed fuel rods. The amount of activity in the primary and secondary coolant at the initiation of the MSLB was assumed to be the maximum levels allowed by the plant Technical Specifications. For North Anna, Technical Specification 3.4.8 requires the primary coolant specific activity to be less than or equal to 1.0 micro-Ci/ gram dose equivalent I-131. Technical '

Specification 3.7.1.4 requires the specific activity of the secondary coolant system to be less than or equal to 0.10 micro-Ci/ gram dose equivalent I-131. The determination of these radionuclide l inventories and concentrations corresponding to these limits is described below.

Consistent with NUREG-0800, Section 15.1.5, Appendix A, iodine spiking above the iodine 3 value allowed for normal operations was considered. Both a pre-accident iodine spike and a i concurrent accident iodine spike were considered. The maximum iodine concentration allowed in )

North Anna Technical Specifications for an iodine spike is 60 micro-Ci/g dose equivalent I-131.

A concurrent iodine spike is defined as an accident initiated increase in the release rate ofiodine from failed fuel rods to a value 500 times the release rate corresponding to the Technical Specifications limit for normal operations. A concurrent iodine spike is more likely than a pre-accident spike since the pressure change caused by an accident can increase iodine releases from failed fuel rods. A pre-accident iodine spike is unlikely, since some independent e,ent would have had to occur shortly before the accident to cause the spike.

The primary liquid, secondary liquid, and secondary steam radionuclide inventories, and the concurrent accident iodine spike appearance rates are given in Table 1.2-1 for the concurrent accident case. The primary liquid, secondary liquid, and secondary steam radionuclide l

inventories are given in Table 1.2-2 for the pre-accident case. The secondary side activity '

levels are initially the same (at the Technical Specification activity limit) for both cases. Only the primary liquid activities differ: the concurrent iodine spike case assumed that the primary coolant activity was initially at the steady state activity limit of 1.0 micro-Ci/g dose equivalent I-131, with iodine added at the appearance rates shown. The pre-accident spike case assumed that the  !

primary liquid activity was initially at the short term Technical Specifications limit of 60 micro-Ci/g dose ^ equivalent I-131. These inventories and appearance rates were input to the LOCADOSE code system to calculate doses from an MSLB. The volume of the primary liquid, secondary liquid, and secondary steam used in this MSLB dose analysis are listed in Table 1.2-3.

1.2.3 Determination of x/Q Values Control room atmospheric dispersion factor (x/Q) values for the release point from the unaffected generators were calculated using the ARCON96 code. The ARCON96 code used meteorological data that covers the years 1989 through 1993 inclusive. These x/Q values are listed in Table 1.2-4.

I Page 13 of 37

The atmospheric dispersion factor (x/Q) values for the EAB and LPZ for both the broken steam line and for the unaffected steam generators are the same as the LOCA analysis values, which are discussed in Section 1.1.2 and listed in Table 1.1-L 1.2.4 Main Steam Line Break LOCADOSE Models The LOCADOSE computer code system was used to model the MSLB. Two LOCADOSE models were created, one for the pre-accident iodine spike and the other for the concurrent accident spike. The only differences between the two models were the initial radioisotope inventories, and the modeling of iodine release from the fuel rods for the concurrent accident case.

The flow rates from the primary coolant to the steam generators prior to the start of the accident were based on the maximum leak rates allowed by Technical-Specifications. The maximum j leakage from one generator (500 GPD) was chosen to be into the generator affected by the steam {

line break. i l

The affected steam generator was modeled as discharging through the turbine building, while the other two generators were modeled as discharging directly to the environment. The flow rates from the affected steam generator liquid to the turbine building and from the turbine building to the environment are summarized in Table 1.2-5.

All of the iodine being released to the environment was conservatively assumed to be airborne. In practice, some of the steam generator discharge to the environment would be as water, which would retain some of the iodine in the liquid phase.

The volume released in thirty minutes (Table 1.2-5) was several times the initial mass of the l steam in the steam generator. Therefore, the volume released from the steam generator to the turbine building was increased above the calculated values to ensure that substantially all of the radionuclides initially present in the affected steam generator were released.

Because the affected steam generator was essentially emptied of liquid during the MSLB, no partitioning of iodine between the liquid and steam was assumed for discharges from the affected generator. A partition factor of 0.01 was assumed for the unaffected generators.

The flow rate from the turbine building to the environment considered the expansion as the steam pressure was reduced to atmospheric in the turbine building. In addition, because the turbine building is not a sealed building, air flow through the building was considered. The building has a forced ventilation system capable of about i volume exchange every six minutes.

However, this forced ventilation system would not function after a loss of off site power. One volume exchange per hour was assumed to be a reasonable air flow rate for the turbine building without forced ventilation. For conservatism, the control room doses were calculated assuming

.only a 0.2 volume / hour air flow rate. A forced ventilation rate of 10 volumes / hour was also evaluated to provide a bounding case for off-site dose calculations.

The model for the control room ventilation system for the MSLB is consistent with that used for the LOCA analysis. The control room was assumed to be on bottled air for the first hour of the Page 14 of 37

accident, while a filtered intake of 1000 CFM was modeled after the first hour, An unfiltered inleakage of 10 CFM was again assumed for the full duration of the accident (0 to 30 days).

Control room occupancy factors were also incorporated into the dose calculations to reflect that personnel would not be exposed to the released activity 100% of the time over the entire 30-day period. The factors that were used were taken from Murphy and Campe (12), and are given in Table 1,1-2. The breathing rate used for the control room dose calculations was 3.47 x 10-4 3

m /sec, which again is consistent with the Murphy and Campe paper. The LOCADOSE code ICRP-30 dose conversion factors were used to determine the doses from inhalation and immersion.

1.2.5 Results of Dose Analysis for MSLB The control room, EAB, and LPZ doses calculated for the MSLB are shown in Table 1.2-6. As indicated in this table, skin and whole body doses resulting from a MSLB were calculated to be less than 0.1 Rem and, thus, are substantially less than the regulatory limits. The limiting accident scenario for the calculation of the thyroid doses was determined to be a concurrent iodine spike case for the control room and EAB. The limiting thyroid dose for the LPZ was determined to be for a pre-accident iodine spike case.

The calculated control room thyroid dose from a MSLB was determined to be below the GDC 19 limit. The doses calculated at the EAB and the LPZ were determined to meet the 10 CFR 100 limits.

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Table 1.2-1 Primary Coolant and Secondary Side Radionuclide Inventories:

Technical Specilication Limits Plus Concurrent Iodine Spike

' Concurrent Primary Coolant Secondary Side' Activities Spike Activity Liquid (3 SG) Steam (3 SG). Appearance Isotope (Ci) (Ci) (Ci) Rate (Ci/hr)

Kr-85m 1.285 E02 3.201 E-03 Kr-85 3.119 E02 7.771 E-03 Kr-87 7.433 E01 1.852 E-03 Kr-88 2.246 E02 5.597 E-03 Xe-133m 1.907 E02 4.751 E-03 1 Xe-133 1.721 E04 4.288 E-01 I Xe-135m - 1.155 E01 2.879 E-04 Xe-135 3.733 E02 9.302 E-03 Xe-138 4.105 E01 1.023 E-03 I

I-131 1.495 E02 1.052 E O1 7.754 E-03 1.155 E04 I-132 5.446 E01 1.276 E 00 9.410 E-04 1.231 E04 I-133 2.416 E02 1.415 E 01 1.043 E-04 2.226 E04 I-134 3.369 E01 3.777 E-01 2.784 E-04 1.583 E04 I-135 1.301 E02 5.450 E 00 - 4.018 E-03 1.659 E04 I

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O a Table 1.2-2 Primary Coolant and Secondary Side Radionuclide Inventories:

Technical Specification Limits Plus Pre-Accident Iodine Spike Primary Coolant Secondary Side Activities Activity Liquid (3 SG) Steam (3 SG)

Isotope (Ci) (Ci) (Ci) ,

Kr-85m 1.285 E02 3.201 E-03 Kr-85 3.I19 E02 7.771 E-03 Kr-87 7.433 E01 1.852 E-03 Kr-88 2.246 E02 5.597 E-03 Xe-133m 1.907 E02 4.751 E-03 Xe-133 1.721 E04 4.288 E-01 Xe-135m 1.155 E01 2.879 E-04 Xe-135 3.733 E02 9.302 E-03 Xe-138 4.105 E01 1.023 E-03 I-131 8.969 E03 1.052 E 01 7.754 E-03 1-132 3.267 E03 1.276 E 00 9.410 E-04 1-133 1.450 E04 1.415 E 01 1.043 E-02 I-134 2.022 E03 3.777 E-01 2.784 E-04 I-135 7.805 E03 5.450 E 00 4.018 E-03 l

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Table 1.2-3

- Volumes Used in Analysis of Main Steam Line Break

- Description Volume (ft3) Notes Environment --

Primary Coolant 9786 Secondary Liquid 2054 One steam generator Secondary SMam 3838 One steam generator 5

Control Room 2.3 x 10 6

Turbine Building 6.0 x 10 Table 1.2-4 x/Q for Releases from the Unaffected Steam Generators 3

Location Time x/Q (Sec/m )

l Control Room 0 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 7.68 x 10-3 )

2 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 5.21 x 10-3 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1.83 x 10~3 24 to 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> 1.34 x 10'3 96 to 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> 1.15 x 10'3 i

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, Flow From Affected SG to the Turbine Building and Froni the Turbine I

Building to the Environment Time Volume Released (CFM) Volume Released (CFM)

Seconds SG to Turbine Building Turbine Building to Environment 4 7 0- 10 1.01 x 10 1.14 x 10 3 6 10 - 180 5.72 x 10 1.36 x 10 3 3

'180 - 1800 1.57 x10 1.83 x 10 j l >l800 0.0 0.0 Table 1.2-6 i

MSLB Control Room and Off Site Doses .

l Control Room GDC-19 EAB 2-hour LPZ 30-Day 10 CFR 100 l 30-Day Dose

• Limit Dose Type Dose (Rem) Limit (REM) Dose (Rem) Dose (Rem) (Rem)

Thyroid 5.0 30 0.8 <0.1 300 Skin <0.1 30 <0.1 <0.1 --

Whole Body ** <0.1 5 <0.1 <0.1 25 Control room skin dose limit is not specified in GDC-19; limit shown is taken from the SRP Section 6.4.

The whole body dose as calculated by LOCADOSE code is dose from the external exposure l

only.

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O e 1.3 STEAM GENERATOR TUBE RUPTURE (SGTR)

A steam generator tube rupture (SGTR) is a break in a tube carrying primary coolant through the l steam generator. This postulated break allows primary liquid to leak to the secondary side of the steam generator with an assumed release to the environment through the steam generator Power Operated Relief Valves (PORVs) or the steam generator safety valves. Steam is assumed to be discharged from the affected generator to the environment for 30 minutes until the generator is isolated. As required by the NRC Standard Review Plan, the SGTR analysis was performed assuming both a pre-accident iodine spike and a concurrent accident iodine spike. In addition, both loss-of-offsite power (LOOP) and no loss-of-offsite power conditions were considered.

In a SGTR, the release point for all three steam generators is the same as for the unaffected steam generators in the MSLB. However, since there is a short time delay between the tube rupture and isolation of the control room inlet during a SGTR under non-LOOP conditions, a control room x/Q for the normal inlet also had to be calculated.

1.3.1 SGTR Analysis Assumptions The Westinghouse Owner's Group has evaluated the impact of tube bundle uncovery on releases from a Steam Generator Tube Rupture. As indicated in this evaluation (Reference 13), STGR releases consist of four components:

1) Releases from secondary liquid boiling including allowance for a partition factor of 0.01 for

_ iodine between secondary liquid and steam.

2) Releases froa the fraction of primary liquid break flow that flashes to steam. A partition factor of 1 is assumed for this flashing fraction.

3) Releases from primary liquid bypassing the secondary side.

4) Releases caused by secondary moisture carryover.

Releases from a SGTR are dominated by the first two terms above for a case with a stuck open PORV. A stuck open PORV also produces a larger radionuclide release than a cycling PORV or a PORV that fails closed and causes the steam generator safety valves to open to relieve secondary side pressure.

1 Uncovery of the tube bundle in a SGTR does not significantly increase radionuclide releases for the stuck open PORV case. If the tube bundle is uncovered in a SGTR and the PORV is stuck open, the third release term described above increases, but it is still only a small part of the total

release.

The LOCADOSE computer models for the SGTR analysis shown below are based en the methodology developed by the Westinghouse Owner's Group (13). These models include only the first two terms discussed above. This does not significantly affect the model results because these two terms dominate the releases for the stuck open PORV case, which is the limiting case for radionuclide releases.

Page 20 of 37

I 1.3.2 Initial Radioisotope Concentrations Initial radionuclide concentrations of the primary and secondary systems for the SGTR accident ,

were assumed to be the same as those for the MSLB. The transient analyses of both the SGTR j and the MSLB accidents indicate that no additional fuel rod failures occurred as a result of these !

transients. Thus, radioactive material releases are determined by the radionuclide concentrations i initially present in primary liquid, secondary liquid, and secondary steam, plus any releases I from fuel rods that 'have failed before the transient. These radionuclide inventories and concentrations are discussed in Section 1.2.2 ( Tables 1.2-1 and 1.2-2).

1.3.3 Determination of x/Q Values During a SGTR with loss of off site power, the steam generators release steam through the secondary system PORVs. tie antrol room x/Q values for releases frem the steam generator l PORVs to the control room v. . ; normal air inlet and the emergency air inlets are shown in Table 1.3-1 for the unaffected generators. These values were calculated with the ARCON96 code. Since the affected steam generator x/Q values are smaller than the values for the unaffected generators because of the higher discharge velocity, only the more conservative unaffected steam generator x/Q values were used.

1.3.4 Steam Generator Tube Rupture LOCADOSE Models The LOCADOSE computer code system (5)(6)(7) was used to model the SGTR. Models were developed for both a pre-accident iodine spike case and a concurrent accident iodine spike case.

The two models are identical except for the initial radioisotope inventories, and the inclusion of modeling of iodine release from the fuel rods for four hours for the concurrent accident case, j l The primary system, steam generator, and control room volumes for the SGTR are the same as I for the MSLB (Table 1.3-3). The liquid properties are also the same. As for the MSLB analysis, the release of the radionuclides contained in the steam from all three steam generators was modeled as essentially a puff release occurring when the PORVs open.

The primary coolant leakage to the unaffected steam generators was based on the maximum leakage allowed by Technical Specifications. The maximum leakage allowed from all three generators in North Anna Technical Specification 3.4.6.2-C is 1 GPM.

For conservatism, all of this leakage was assumed to occur into the two unaffected steam generators. This assumption is conservative because the unaffected generators release steam to the environment for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> compared to 30 minutes for the affected generator.

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The break flow rates through the ruptured tube to the affected steam generator were based on a

, thermal hydraulic analysis of a complete double ended tube rupture. To be consistent with the regulatory guidance in SRP Section 15.6.3, the liquid and steam break flows are modeled separately. The break flow rates and release rates to the environment are summarized in Tables 1.3-2 and 1.3-3 for both no LOOP and LOOP cases respectively.

f Page 21 of 37 t

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The liquid break flow from the primary system is modeled as mixing with the secondary liquid in the affected steam generator. The flow from the secondary liquid to the secondary steam is then j modeled assuming a partition factor of 0.01 for iodine. This technique for modeling a SGTR with uncovery of the tube bundle was developed in a generic study by the Westinghouse Owners Group as discussed in Reference 13.

The fraction of the break flow that flashes to stemn is modeled as being transferred to the 3

. affected steam generator steam space. Once in the steam generator steam space, the radionuclides j in this part of the break are almost immediately released to the environment.

The primary and secondary system releases are replaced with safety injection and auxiliary I feedwater flows. Therefore, the volume of the primary and secondary liquids remains relatively constant during this transient.'

The flow from the affected generator through the condenser was represented for the time interval between the tube rupture and the opening of the PORV. During this period, there is some build-up of radionuclide inventory in the affected generator liquid and steam volumes. A very small volume and a large return flow to the steam generator liquid space was used for the condenser.

This conservatively ignores the dilution and retention of radionuclides in the condenser. The flow through the condenser for the unaffected generators is not modeled because there is no rapid build-up of radionuclides in these generators. The radionuclide inventory in these generators is modeled based on the initial inventory and the primary to secondary leakage.

The model for the control room ventilation system for the SGTR is similar to that used for the LOCA and MSLB analyses, with some differences incorporated to more accurately model the timing of the sequence of events of the SGTR. The start of the accident was assumed to be the tube mpture itself. For the case where no loss of offsite power was assumed, the PORV on the faulted steam generator was determined to open 107 seconds after the break, and the SI signal was generated at 196 seconds. Until the SI signal is generated the control room was supplied air via the normal ventilation system with a 2800 CFM intake air flow rate. Subsequent to the SI signal, the control room is isolated, and the bottled air system dumps. For the case where loss of offsite power was considered, the PORV on the faulted steam generator was determined to open 103 seconds after the break and the SI signal was generated at 232 seconds. For this case, the control room was isolated automatically on the loss of offsite power (LOOP) and the bottled air i system activated. In either case once the bottled air system was activated the control room was  ;

assumed to be on bottled air for a period of 1.0 hour0 days <br />0 hours <br />0 weeks <br />0 months <br />. After 1.0 hour0 days <br />0 hours <br />0 weeks <br />0 months <br /> the control room was provided with a filtered air supply of 1000 CFM for the duration of the accident. An unfiltered i inleakage of 10 CFM was assumed to occur during the entire period that the control room was isolated. The' control room intake filter efficiency was assumed to be 95% and 0% for elemental and organic iodine, respectively. Control room occupancy factors were also incorporated into the dose calculations to reflect that personnel would not be exposed to the released activity 100% of th: time over the entire 30-day period. The factors that were used were taken from Murphy and Campe (12), and are given in Table 1.1-2. The breathing rate used for the control room dose 3

calculations was 3.47 x 10-4 m /sec, which again is consistent with the Murphy and Campe paper.

Page 22 of 37

1.3.5 Results of Dose Calculations for SGTR Both pre-accident and concurrent accident iodine spike cases were analyzed for the steam generator tube rupture. The skin and whole body dose results for both cases were below 0.1 Rem. These low doses are well below the regulatory criteria.

l The limiting case for the control room thyroid dose was determined to be for the non-LOOP pre-accident iodine spike case. A comparison of the doses calculated for the limiting SGTR accident scenario to the regulatory limits is shown in Table 1.3-4. All calculated control room doses for the North Anna steam generator tube rupture remain below the SRP section 6.4 and GDC-19 criteria.

The EAB and LPZ doses shown in Table 1.3-4 are less than the 10 CFR 100 limits and meet the SRP 15.6.3 review criteria of less than a small fraction (10%) of 10 CFR 100.

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i Table 1.3-1 x/Q for Releases from the Unaffected Steam Generators Location Time x/Q (Sec/m3)  ;

Control Room NormalIntake O to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 1.40x10-3 Control Room Emergency Intake 0 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 7.68x10-3 Control Room Emergency Intake 2 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 5.21x10-3 Control Room Emergency Intake 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1.83x10-3 Control Room Emergency Intake 24 to 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> 1.34x10-3 Control Room Emergency Intake 96 to 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> 1.15x 10-3 EAB 0 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 3.10x10-4 LPZ 0 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 1.10x10-5 LPZ 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 7.30x10-6 i LPZ 24 to 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> 3.00x10-6 j LPZ 96 to 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> 8.20x10-7 1

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Table 1.3-2 l

Steam Generator Tube Rupture Break Flow Rates and Releases l For No Loop Case Affected Steam Generator RCS to SG RCS to SG SG Liquid Steam Release l Time Liquid Flow Steam Flow to Steam to Environment i

! (Sec)- (lb) (CFM) (Ib) (CFM) (CFM) (CFM) 0- 107 6399.9 80.21 1466.2 18.38 1494.0 0

, 107 - 196 5644.3 85.05 211.4 3.19 168.6 4270 l

196 - 1800 98418.I 82.28 2937.1 2.46 153.0 3875 l

Time SG Liquid to SG Steam to Steam Flow Condenser Flow I (Sec) (CFM) (CFM) 0 - 107 ----

37800 Unaffected Steam Generators Integral Flows and Release Rates j Time Period Steam Mass Flow Rate (Ib) (CFM)

O sec - 107 sec 0 0 i 107 sec - 1800 sec 256533 191.35 30 min - 2 hr 280000 65.48 2 hr -

8 hr 861000 50.34 Page 25 of 37

Table 1.3-3 Steam Generator Tube Rupture Break Flow Rates and Releases For Loop Case Affected Steam Generator RCS to SG RCS to SG SG Liquid Steam Release Time Liquid Flow Steam Flow to Steam to Environment (Sec) (Ib) (CFM) (Ib) (CFM) (CFM) (CFM) 0- 103 6986.3 90.96 894.5 11.64 1494.0 0 103 - 232 7367.4 76.59 333.8 3.47 164.2 4158 232 - 1800 92112.7 78.78 5679.3 4.86 131.3 3327 Time SG Liquid to SG Steam to Steam Flow Condenser Flow (Sec) (CFM) (CFM) 0 - 103 ----

37800 Unaffected Steam Generators Integral Flows and Release Rates Time Period Steam Mass Flow Rate (Ib) (CFM) 0 sec - 103 sec 0 0.

103 sec - 1800 sec 153049 113.90 30 min - 2 hr 230000 53.79 2 hr -

8hr 662000 38.70 Page 26 of 37

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Table 1.3-4 SGTR Worst Case Control Room and Off Site Doses Control Room GDC-19 EAB 2-hour LPZ 30-Day 10 CFR 100 30-day Dose Dose Limits' Dose _ Dose Dose Limit Dose Type (Rem) (Rem) (Rem) (Rem) (Rem) i 2 3 Thyroid 12.31 30 19.5 0.72 300 Skin <0.1 30 <0.1 <0.1 - l 4

Whole Body <0.1 5 <0.1 <0.1 25 l

1. Control room skin dose limit is not specified in GDC-19; limit shown is taken from SRP Section 6.4.

2. No LOOP Pre-Accident Iodine Spike case.  !

3. LC P Pre-Accident Iodine Spike case.

5. The wi. body dose as calculated by LOCADOSE code is dose from the external exposure only.

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1.4 LOCKED ROTOR ACCIDENT (LRA)

The Locked Rotor Accident (LRA) evaluates the consequences of the sudden seizure of the rotor of one of the reactor coolant pumps. Similar results would be expected for a shear failure of a

! shaft in the reactor coolant pump. In these types ci accidents, flow through the affected loop l drops rapidly while the core is still at power, and some degree of reverse flow would be expected through the affected loop. The low flow ir, the affected loop leads to a reactor and turbine trip, but the partial loss of flow while the core is at power results in a degradation in heat transfer which could in turn result in fuel damage.

l Although there is no increase in the leakage of primary coolant to the secondary side in the LRA, activity (from the failed fuel) may be transported to the secondary side via any preexisting leaks in the steam generators. If there is a loss of off site power, activity is released to the atmosphere through the steam generator safety valves and/or the Power Operated Relief Valves (PORVs)

! until the plant cools down and the reactor is secured in a safe condition.

1.4.1 LRA Analysis Assumptions Consistent with Sections 15.3.3-15.3.4 of the Standard Review Plan (1) the reactor was initially l assumed to be operating at 102 % of the stretch power level for this analysis. A turbine trip and l coincident loss-of-offsite power (LOOP) was incorporated into the analysis. With the assumed loss-of-offsite power, releases were through the steam generator PORVs and safety valves.

Both primary and secondary side coolant activities were set at the maximum levels permitted by the plant Technical Specifications. The primary coolant activity level also assumed a pre-accident iodine spike to the maximum level allowed by the North Anna Technical Specifications as well as the contribution due to fuel failures that resulted from the LRA. Pre-existing primary to secondary leakage was modeled at the maximum levels permitted by the Technical Specifications.

The possibility of uncovery of the upper portion of the steam generator tube bundle (based on collapsed liquid levels) during a LRA was not considered in previous LRA dose calculations.

For the current evaluation, the approach taken was that developed by the Westinghouse Owners Group. The probability of coincidental occurrence of a LRA, a pre-existing steam generator tube leak above the collapsed liquid level, and condenser unavailability due to a loss-of-offsite power 4

was determined to be sufficiently small (<5x10 ) that it was not necessary to evaluate this combination of conditions. Therefore, any leaks in the steam generator tube bundle were l assumed to remain covered throughout the accident.

When the tubes are covered, the secondary side water provides a sembbing action, trapping some of the activity from iodine in the primary fluid in the secondary liquid. The noble gases are unaffected by this process. Whenever the steam generator tubes were covered, this analysis used an iodine partitioning factor of 0.01 to account for this effect. The value of this partitioning factor was consistent with that used in previous LRA analyses, and is also given in Section 15.6.3 of the SRP for the Steam Generator Tube Rupture (SGTR), which has a release mechanism similar to that seen in the LRA.

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For a LRA where the PORVs cycle open and closed as they are designed to do, conditions that would generate an SI signal are not created. Without an SI signal, the North Anna control room is not automatically isolated. However, consistent with Sections 15.3.3 - 15.3.4 of the Standard Review Plan, loss-of-offsite power was assumed to be concurrent with the locked rotor accident.

That resulted in immediate isolation of the control room. If no loss-of-offsite power was assumed, then the condenser would be available for dumping of steam and no release would occur. Therefore, for releases to occur the loss-of-offsite power assumption had to be made. This scenario resulted in more conservative results for the control room dose calculations.

The assumed sequence of events used in this current dose evaluation for a LRA at North Anna were as follows: The accident was initiated when one reactor coolant pump rotor locks. Power to the other two reactor coolant pumps was assumed to be lost shortly thereafter, after the reactor trips. Assuming the steam condensers are unavailable due to loss of off site power, the PORVs on the two unaffected steam generators open within seconds of the accident, and the PORV on the steam generator in the affected loop also opens within one minute. Most of the releases in the first minute were through the unaffected steam generators; after the third PORV opened, the steam release through all three steam generators was assumed to be essentially identical. Releases were conservatively modeled as starting immediately.

These releases were assumed to occur for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, by which time the Reactor Coolant System (RCS) temperature has been decreased to 350 F. At this point the Residual Heat Removal (RH) system was activated, and releases to the atmosphere through the steam generator PORVs ceased.

1.4.2 Initial Radioisotope Concentrations The amount of activity released is dependent on both the amount of activity in the coolant at the time of the accident and the activity released due to fuel failures. As for the MSLB and SGTR accident analyses, the amount of activity in the primary and secondary coolant at the initiation of the locked rotor accident was assumed to be the maximum levels allowed by the plant Technical Specifications.

The LRA analysis assumed a pre-accident iodine spike, so the primary coolant iodine levels at the start of the accident were conservatively set at the short term limit of 60 micro-Ci/ gram, rather than at the 1.0 micro-Ci/ gram dose equivalent I-131 limit for normal operation. An additional source of activity in the primary coolant was due to the releases from additional fuel failures. The Locked Rotor Accident analysis was based on the calculation that 13% of the fuel enters DNB during the accident and fails. These fuel failures were assumed to occur instantaneously at the start of the accident. The total amount of activity in the primary coolant at the start of the LRA was then the Technical Specification activity level plus the activity due to failed fuel. Tables 1.4-1 and Table 1,4-2 give the primary and the secondary side radionuclide inventories for the LRA.

1.4.3 Locked Rotor Accident LOCADOSE Model The LOCADOSE computer code system (5)(6)(7) was used to calculate the doses for the LRA.

The primary and secondary system volumes used in this analysis were the same as those used in l

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l the MSLB, and are given in Table 1.2-3. The leakage from the primary coolant to the secondary l system through the steam generators was set at the maximum leakage allowed by the North Anna Technical Specifications, or i gpm through all three steam generators. As discussed in Section 1.4.1, the LRA was modeled assuming that any leaks in the steam generator tube bundle remain covered throughout the accident. The primary coolant was therefore modeled as leaking to the steam generator liquid volume. The flow from the secondary liquid to the secondary steam assumed a partition factor of 0.01 for iodine. <

In the LRA, most of the releases in the first minute were through steam releases from the two unaffected steam generators. After the third PORV opens (within one minute), the steam release through all three steam generators was assumed to be essentially identical. To simplify the modeling of this accident, the releases were treated as being identical through all three steam generators for the entire release period. The releases were also modeled as starting immediately, rather than a few seconds after initiation of the accident (when the PORVs open), and continue for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. The steam releases for the LRA are given in Tablel.4-3.

Consistent with SRP Sections 15.3.3 - 15.3.4, loss-of-offsite power was assumed to be concurrent with the locked rotor accident. This results in immediate isolation of the control room. The air into the control room was supplied by banks of bottled air for the first I hour. At the end of the I hour period, the control room was supplied via an emergency ventilation system, which provides the control room with filtered air drawn from the turbine building. There are 4 fan / filter trains, rated at 1000 CFM capacity each, which can provide . intake to the control room !

pressure envelope. Only one of these fans was assumed to be supplying the control room after the l bottled air supply is depleted (i.e., 1000 CFM filtered intake). The iodine removal efficiency for the emergency ventilation air supply filters was assumed to be 95%, in accordance with Regulatory Guide 1.52 (Reference 14). This emergency ventilation supply is assumed to be used for the remainder of the 30-day period for which control room doses are calculated.

A 10 CFM unfiltered inleakage into the control room was assumed to occur when the control room was isolated, and to continue until the end of the 30-day dose calculation period. This unfiltered inleakage was intended to simulate the effects of personnel entries to (and exits from) the control room pressure envelope as required by emergency operating procedures. The value of 10 CFM was taken from SRP Section 6.4 (Reference 1).

The atmospheric dispersion factors used for the LRA control room dose calculations were the .

same as those used for the MSLB analysis for the control room because both scenarios involve l the same release and receptor points. These values are listed in Tables 1.2-3. Control room occupancy factors were also incorporated into the dose calculations to reflect that personnel would not be exposed to the released activity 100% of the time over the entire 30-day period. The factors that were used were taken from Murphy and Campe (12), and are given in Table 1.1-2.

3 The breathing rate used for the control room dose calculations was 3.47 x 10-4 m /sec, which again is consistent with the Murphy and Campe paper.

l 1.4.4 Results of Dose Calculations for LRA )

l The dose calculations for a LRA with the model and assumptions described above are summarized in Table 1.4-4. The thyroid doses reported are inhalation doses, and the skin doses Page 30 of 37

I.

are immersion doses. The whole body doses given in Table 1.4-4 include the whole body doses

' due to both inhalation and immersion. The calculated control room doses for a LRA are less than the limits specified by SRP Section 6.4 and GDC-19 and the off-site doses are below the limits

'specified in 10 CFR 100.

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Table 1.4-1 Pre - Accident Spike TS Weighted 1-Equivalent Primary Coolant and Secondary Side Nuclide Inventory:

Isotopes ' Primary Secondary Secondary

- Coolant Liquid Steam Activity (Ci) Activity (Ci-3Sg) Activity (Ci-3Sg)

Kr-85m 1.285E+02 3.201E-03 Kr-85 3.119E+02 7.771E-03 Kr-87 7.433E+01 1.852E-03 Kr-88 2.246E+02 - 5.597E-03 Xe-133m 1.907E+02 4.751E-03 Xe-133 1.721E+04 4.288E-01 Xe-135m 1.155E+01 2.879E-04

)

Xe-135 3.733E+02 9.302E-03 Xe-138 4.105E+01 1.023E-03 I-131 8.969E+03 1.052E+01 7.754E-03 I-132 3.267E+03 1.276E+00 9.410E-04 I-133 1.450E+04 1.415E+01 1.043E-02 I-134 2.022E+03 3.777E-01 2.784E-04 1-135 7.805E+03 5.450E+00 4.018E-03 Table 1.4-2 Pre - Accident Spike And 13% Failed Fuel Primary Coolant Nuclide Inventory:

Isotopes Primary Coolant Activity (Ci)

Kr-85m 4.989E+05 Kr-85 4.763E+04 I Kr-87 8.980E+05 ,

Kr-88 1.231E+06 i

Xe-133m 5.341E+04 Xe-133 2.179E+06 i Xe-135m 5.987E+05 Xe-135 2.063E+06 Xe-138 1.836E+06 I-131 9.734E+05 I-132 1.467E+06 I-133 2.176E+06 I-134 2.530E+06 I-135 1.970E+06 l

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1 Table 1.4-3

)

Volumes Released from Steam Generators During a Locked Rotor Accident ,

Time Averaged Flow Rate (CFM)*

Time (sec) Time (Hr) Liauid Steam 0 0.00 242 6141 900 0.25 101 2563 1800 0.50 46 1169 3600 1.00 39 998 7200 2.00 35 896 l

1

• Note: The flow rate shown at a given time is the average flow rate from that time until the next time shown: for example, the liquid flow rate from 0 to 900 see is 242 CFM.

Table 1.4-4 ,

l LRA Control Room and Off Site Doses Control Room GDC-19' EAB 2-hour LPZ 30-Day 10 CFR 100 30-day Dose Dose Limit Dose Dose Dose Limit Type (Rem) (Rem) (Rem) (Rem) (Rem)

Thyroid 6.8 30 0.9 0.3 300 Skin 8.3 - 30 0.2 0.1 --

2 Whole Body 0.21 5 0.1 0.01 25

1. Control room skin dose limit is not specified in GDC-19; limit shown is taken from the SRP Section 6.4.

2. The whole body dose as calculated by LOCADOSE code is dose from the extemal exposure only.

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2.0

SUMMARY

OF DOSE ANALYSES Virginia Electric and Power' Company has reanalyzed the offsite and control room dose consequences for the following postulated accidents at North Anna:

1. Large Break Loss of Coolant Accident (LOCA)

2. Major Secondary Steam Pipe Rupture (i.e., Main Steam Line Break, MSLB)

3. Steam Generator Tube Rupture (SGTR)

4. Locked Reactor Coolant Pump Rotor Accident (i.e., LRA)

Control room and offsite (i.e., EAB and LPZ) dose results for the above analyses are summarized in Table 2.0-1. All EAB and LPZ doses were determined to be less than the appropriate 10 CFR 100 limits. All control room operator doses were determined to meet the regulatory limits specified by GDC-19.

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Table 2.0-1 Summary of Control Room and Off Site Doses CONTROL ROOM Accident Thyroid Whole Body Skin Tvoe Dose (Rem) Dose (Rem) Dose (Rem)

LOCA 27.8 1.1 < l .0 MSLB 5.0 <0.1 <0.1 SGTR 12.3 <0.1 <0.1 LRA 6.8 0.2 8.3 l

EXCLUSION AREA BOUNDARY Accident Thyroid Whole Body' Skin i Tvoe Dose (Rem) Dose (Rem) Dose (Rem)

LOCA 72.5 1.6 3.5 MSLB 0.8 <0.1 <0.1 SGTR 19.5 <0.1 <0.1 LRA 0.9 0.1 0.2 LOW POPULATION ZONE Accident Thyroid Whole Body' Skin Tvoe Dose (Rem) Dose (Rem) Dose (Rem)

LOCA 3.1 < l .0 < l .0 MSLB <0.1 <0.1 <0.1 SGTR 0.7 <0.1 <0.1 LRA 0.3 <0.1 0.1

1. The whole body dose as calculated by LOCADOSE code is dose from the extemal exposure only.

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3.0 REFERENCES

l (1)- U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, " Standard

j. Review Plan," NUREG-0800, Revision 2, July 1981.

(2) Ixtter from L. B. Engle (NRC) to J. P. O'Hanlon (Virginia Power), " North Anna Units 1 And 2 -Issuance Of Amendments Re: Revisions To The Facility Operating Licenses And Technical Specifications To Allow The Containment Personnel Airlock Doors To Remain Open During Refueling Operations (Tac Nos. M94076 And M94078),"

Amendment Nos.198 and 179, Serial No.96-123, February 27,1996.

~(3) Letter from W. R. Cartwright (Virginia Power) to U.S. Nuclear Regulatory Commission,"

Virginia Electric and Power Company, North Anna Power Station Units I and 2, Control Room Habitability, Engineering Evaluation And Proposed License Amendment," Serial  !

Number 89-022,' March 1,1989.

(4) Letter from W. L. Stewart (Virginia Power) to U.S. Nuclear Regulatory Commission,

" Virginia Electric and Power Company, North Anna Power Station Units 1 and 2, Control Room Habitability - Supplemental Information," Serial Number 89-022A, December 22, 1989.

(5) "LOCADOSE NE319, A computer Code System for Multi-Region Radioactive Transport and Dose Calculation," Theoretical Manual, Revision 3, July 1990, Bechtel Power Corporation, San Francisco, CA.

(6) "LOCADOSE NE319, A computer Code System for Multi-Region Radioactive Transport and Dose Calculation," User's Manual, Revision 3, July 1990, Bechtel Power Corporation, San Francisco, Ca.

(7) "LOCADOSE NE319, A computer Code System for Multi-Region Radioactive Transport and Dose Calculation," Validation Manual, Revision 3, July 1990, Bechtel Power Corporation, San Francisco, Ca.

(8) Regulatory Guide 1.4, " Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Pressurized Water Reactors," Rev. O, Nov.1970.

(9) Computer Code ARCON96, " Atmospheric Relative Concentrations in Building Wakes,"

NUREG/CR-6331, PNNL-10521, Rev.1, May 1997.

(10) U. S. Nuclear Regulatory Commission Regulatory Guide 1.145, "AtmospNric Dispersion Models For Potential Accident Consequence Assessments At Nuclear Power Plants,"

Revision 1, November 1982, Re-issued February 1983.

(11) North Anna FSAR Chapter 15.4, January 3,1973.

4 Page 36 of 37

. e (12) K. G. Murphy and K. M. Campe, " Nuclear Power Plant Control Room Ventilation System Design for Meeting General Design Criterion 19," 13* AEC Air Cleaning Conference, August 1974.

-(13) WCAP-13132 "The Effect of Steam Generator Tube Uncovery of Radio iodine Release,"

January 1992.

I (14) U.S. Nuclear Regulatory Commission, Office of Standards Development, " Design, Testing, and Maintenance Criteria for Post Accident Engineered-Safety-Feature ,

Atmosphere Cleanup System Air Filtration and Adsorption Units of Light-Water-Cooled i Nuclear Power Plants," Regulatory Guide 1.52, Revision 2, March 1978.

(15) letter from L. B. Engle (USNRC) to W. L. Stewart (VEPCO), " North Anna 1 and 2 -  !

Issuance of Amendments Re: Limiting Dose to Control Room Operators (TAC Nos.

73788 and 73789)," Amendment Nos.126 and 110, Serial No.90-116, February 28, 1990.

(16) Letter from James P. O'Hanlon of. Virginia Power to U.S. Nuclear Regulatory l Commission, " Virginia Electric And Power Company North Anna Power Station Units 1 and 2 Proposed Technical Specifications Changes To Allow The Containment Personnel Airlock Doors To Remain Open During Refueling Operations," Serial No.95-506, October 17,1995.

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