ML20235T532
| ML20235T532 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 03/01/1989 |
| From: | Cartwright W VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 89-022, 89-22, NUDOCS 8903080336 | |
| Download: ML20235T532 (42) | |
Text
, - _ _ _ -
s VIRGINIA 13LECTHIC AND PowEn COMPANY Hicirnoxn.VINGINIA 20261 W.H.CARTWRMHT
" ')"E"'
March 1, 1989 m
United States Nuclear Regulatory Commission Serial No.89-022 Attention: Docuinent Control Desk NAPS /JHL Washington, D. C. 20555 Docket Nos.
50-338 50-339 License Nos. NPF-4 NPF-7 Gentlemen:
VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA POWER STATION UNITS 1 AND 2 CONTROL ROOM HABITABILITY ENGINEERING EVALUATION AND PROPOSED LICENSE AMENDMENT In December 1986, the NRC conducted a Control Room Habitability Survey at North Anna Power Station.
The survey was documented in your May 4, 1987 letter and in NRC Inspection Report 87-19, dated September 10, 1987.
The results - of the survey identified various control room habitability issues that needed to be resolved. The response to the survey was documented in our letters dated June 8, 1987 and September 11, 1987.
One of the survey issues dealt with control room access during an emergency.
The survey identified that the procedure for limiting access to the control room may not be workable.
Our response to this issue indicated that an engineering evaluation would be performed to determine the best method for limiting access to the control room following an eccident.
In September 1988, a followup NRC inspection was performed to resolve the issues opened in the Control Room Habitability Survey.
This followup inspection is documented in Inspection Report 88-28, dated November 23, 1988.
During this inspection, we committed to provide the results of the engineering evaluation and any proposed corrective actions by February 28, 1989.
This letter is to document the results of the engineering evaluation and describe the corrective actions that were taken and ensure compliance with General Design Criteria (GDC) 19 of Appendix A to 10 CFR 50.
The results of the engineering evaluation to allow multiple control room entries in the event of an accident is provided in Attachment 1.
The engineering evaluation reviewed various design basis accidents with relation to radiation exposure to the control room operator.
The engineering evaluation determined that the Main Steam Line Break and the Steam Generator Tube Rupture are the two most limiting accidents for radiation exposure to the i
1 89030GO336 890301 PDR ADOCK 05000338
_______________P___
PDC 2
control room operator.
As a result of this evaluation, the operation of the control room emergency ventilation system has been revised such that the system is normally aligned in the recirculation mode.
This system alignment will allow control room access and egress during an accident and ensure the limits of GDC 19 are not exceeded. Operating procedures have been revised to ensure the control room ventilation system is operated in the recirculation mode.
Emergency operating Procedures have also been revised to ensure the initiation of the emergency ventilation system with filtered recirculation within 10 minutes of the initiation of a Safety Injection Signal.
Emergency Plan Implementing Procedures are being reviewed to determine the need to require radiation surveys of the control room when the plant has entered into the Emergency Plan due to a design basis accident. Also, Operating Procedures will be revised to provide guidance to control room operators as a result of the presence of high radiation in the control room.
In addition, the engineering evaluation determined that certain controls needed to be implemented, to ensure radiation exposures to the control room operator do not exceed the limits of GDC 19, in the event of a Fuel Handling Accident.
As a result, Operating Procedures have been revised to have the control room operator manually isolate the normal supply and exhaust for the control room during fuel movement in the fuel building, The control room ventilation is then provided by an operating emergency fan and filter aligned to the Turbine Building.
Abnormal Operating Procedures will be revised, by February 28, 1989, to require the manual initiation of the bottled air system on a high high radiation signal in the fuel building.
The attached engineering evaluation determined that radiation exposures to control room operators do not exceed the limits of GDC 19, when multiple control room entries are allowed during an accident.
As indicated in the engineering evaluation, radiation exposure to control room operators would be increased above the assumptions currently delineated in Chapter 15 of the North Anna Updated Final Safety Analysis Report.
As a result of the increased radiation exposure to control room operators, it has been determined, by the Station Nuclear Safety and Operating Committee, that multiple control room entries during an accident involves an unreviewed safety question as defined in 10 CFR 50.59.
Therefore, pursuant to 10 CFR 50.59, the Virginia Electric and Power Company requests a license amendment, per the requirements of 10 CFR 50.90, for North Anna Units 1 and 2.
The amendment would add a license condition stating that the limiting doses to control room operators shall be revised in accordance with this submittal.
The proposed changes to the North Anna 1 and 2 facility operating license is provided in Attachment 2.
This request has been reviewed and approved by the Station Nuclear Safety and Operating Committee and the Safety Evaluation and l
Control Staff.
It has been determined that this request does involve an unreviewed safety question as defined in 10 CFR 50.E Dut does not involve a significant hazards consideration as defined in 10 U R 50.92.
Therefore, NRC review and approval is required.
Inspection Report 87-19 identified a concern (Inspector Followup Item 87-19-07) related to maintaining the control room envelope at a positive pressure with relation to adjacent areas.
This concern was also discussed in
s Inspection Report 88-28 dated November 23, 1988.
To resolve this concern, measures have been taken to verify that the control room envelope is maintained at an adequate positive pressure.
Operator Logs have been revised to require verification, every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, that the control room envelope is at an adequate positive pressure with relation to the Turbine Building.
If the control room envelope is not at an adequate positive pressure with relation to the Turbine Building, a ventilation balance is performed across the control room envelope.
If an adequate positive pressure cannot be maintained following the ventilation balance, a maintenance work request is submitted to correct the deficiency.
North Anna is currently in compliance with GDC 19.
- However, further engineering review is being performed in order to relieve the control room operators from having to perform the manual ventilation system lineups.
The engineering review is expected to be completed by June 1989, and any required modifications will be subsequently implemented.
The attached analysis, with the procedural controls described above, assures compliance with GDC 19 and justifies continued operation.
If you have any questions, please contact us immediately.
Verytrly70ups/
A
/
W. R. Cartwri ht Attachments.
cc:
U. S. 9Jclear Regulatory Commission 101 Marietta Street, N. W.
Suite 2900 Atlanta, Georgia 30323 Mr. J. L. Caldwell NRC Senior Resident Inspector l
North Anna Power Station l
1
. s
'4 L
- COMMONWEALTH OF VIRGINIA )
)
COUf1TY OF HENRICO-
)
The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by W. R. Cartwright who is Vice. President -
Nuclear, of Virginia Electric and Power Cotapany.
He is duly authorized.to execute and file the foregoing' document in. behalf of that Company, and..the statements in the' document are true to the best of his knowledge and belief.
Acknowledged before me this day of
///r/
,19 8C/
My Commission expires:
3ho 85,19 94 J
/bdi YhuR Notary Public~
(SEAL)
i.-
i 6
ATTACHNENT I-i SAFETY EVALUATION FOR REVISED CONTROL ROOM HABITABILITY CALCULATIONS 1
I i
1 4
i I
l i
NORTH ANNA POWER STATION UNITS 1 AND 2 l
1 -
L I
E_.___________________.__________.._.____
i Safety Evaluation North Anna Control Room Habitability 1.0 -INTRODUCTION l
During accidents at the North Anna Power Station, the control room envelope is pressurized (upon receipt of a control room:ssolation signal) to 0.05 inch water gauge to minimize in-leakage of air and airborne radioactive material.
The 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 minimum pressure of 0.04 inch water gauge relative to the outside atmosphere.
Previous radiological dose evaluations for the control room during accidents assumed zero in-leakage of unfiltered air while the control room envelope is pressurized. Inherent in this assumption is a condition where no doors to the control room are opened. However, the emergency operating procedures at North Anna require personnel to enter and exit the control room envelope during the accident.
It was therefore necessary to modify the North Anna control room habitaisility calculations to address the impact of the required door openings on the control room doses.
This safety evaluation addresses the impact of both this input modification on the calculated doses for the control room, and some ventilation system 1
m g.
I modifications which are recommended as a result of these evaluations.
The recommended modifications, which are discussed in more detail in Section 3.0, include:
use of the control room ventilation recirculation in conjunction with the bottled air system; automatic. initiation of.
recirculation on safety injection (SI); and automatic isolation of the control. room upon receipt of a high-high radiation alarm from the fuel building.
Per Section 6.4.1.3.1 of the North Anna UFSAR, 31-day cumulative doses to control room personnel for accident conditions were originally determined not to exceed 2.5 rem whole body and 10 rem to the thyroid.
The basis for this evaluation was that the LOCA was the limiting accident-for control room dose calculations. Therefore, no specific analyses were originally performed for other types of accidents.
It has since been recognized that other types of accidents are potentially more limiting then the LOCA for control room dose calculations.
Consequently, the current reevaluation of North Anna control room habitability analyzed several potentially limiting types of accidents.
Because a complete set of such calculations was not originally performed, the selection of the accidents for the current North Anna control room habitability etaluation was based on Part 15 of the Standard Review Plan (NUREG-0800).
The only Condition IV accident discussed in SRP 15 (and Chapter 15 of the North Anna UFSAR) which was not specifically addressed in the control room habitability calculations was the RCCA ejection accident, for which it has been determined that less than 10% of the fuel will fail (Section 15.4.6.2.3.5 of the North Anna UFSAR).
The 2
O consequences of the limited amount of fission product release for the rod ejection accident are bounded by the analysis for the LOCA.
It should be noted that the changes evaluated in this report affect only the control room ventilation system: there are no changes to any assumed accident conditions or calculated release rates which would affect the Chapter 15 offsite dose calculations.
3
L y
2.0 SYSTEM DESCRIPTION l
' The North Anna control room habitability system for radiological protection includes a compressed breathing air system and an emergency i
filtered air system.
The calculations described in Section 4.0 were.
performed to verify that each of these systems will ensure operator doses do not exceed the limits of GDC-19.
The compressed breathing air system is provided to maintain a positive interior control room pressure to assure outward leakage when the outside air is contaminated.
The compresed breathing air system has been designed to provide one hour of positive pressure.
The system is automatically initiated by the Safety Injection System (SIS), and may also be manually activated. The emergency' filtered air system, taking suction from the turbine building through HEPA and charcoal filters, is provided to ensure continued outward leakage and to continue the supply of breathing and pressurization air indefinitely upon depletion of the bottled air supply.
Each reactor unit is provided with an emergency filtered air system comprised of two fan / filter units, Each fan / filter unit is designed to provide 1000 cfm filtered air makeup at 99% efficiency.
The nominal filter efficiency of 99 percent is consistent with Regulatory Guide 1.52.
Each fan / filter unit is capable of two modes of operation: pressurization and recirculation.
The difference between the two modes is the suction.
path:
the pressurization mode takes suction from outside the pressure 4
envelope, while the recirculation. mode takes suction from.within the pressure envelope.
Although both modes of operation employ the use of the same filters, the impact on. dose calculations is different,for each mode of operation.
Therefore different efficiencies were sometimes assumed for the same filters, depending on the mode of operation of the control room emergency ventilation system.
Throughout this evaluation, nominal-filter efficiencies or lower were used. As explained in more detail below, use of a lower filter efficiency represents a
conservatism.in the calculations. The assumed efficiencies were generally sr -ai.ed to ensure consistency with other (previous) dose calculations (e.g., for the LOCA calculations, where conservatively low filter efficiencies. were assumed for consistency with the previous UFSAR analysis), or were based on the guidelines of Regulatory Guide 1.52.
For all calculations, the assumed filter efficiencies were conservative.
During the pressurization mode of operation, envelope pressurization is provided via the turbine building suction dampers.
Because the turbine building radionuclides concentration is greater than the control room envelope concentration, a small amount of contamination enters the pressure envelope in the makeup air when the ventilation system operates in the the pressurization mode.
The unfiltered fraction of makeup air (for example 5 percent of the makeup air for an assumed 95% filter efficiency) contributes to the control room radionuclides inventory.
Thus, in the pressurization mode the use of a lower filter efficiency than 5
p 1'
l L
the nominal value-conservatively increases the calculated airborne l
l activity of the pressure envelope.
l In the recirculation mode of operation, the control room pressure envelope atmosphere is recirculated through the filters via manual dampers on the suction.ductwork.
The pressure envelope radionuclides concentration is greater than the concentration at the discharge of the filter units. The clean up rate is determined by the ratio of filtered air flow to control room volume.
In this. case, ' the filters act as a sink for airborne activity from the pressure envelope, with the filter efficiency providing the fraction of 2000 cfm available as fresh air to the envelope (for example,. 99% of 2000. cfm, or 1980 cfm).
The Technical Specifications.
require that the filters be 99% efficient at a 1000 cfm fluw rate, which means that in the recirculation mode 990 cfm of clean air must enter the control room.
During the first hour of an accident involving the use of the habitability
- system, a-minimum of two of the fuur available fan / filter units are required to operate in the recirculation mode coincident with bottled air system operation.
Following bottled air system depletion, an additional fan / filter unit is placed in operation in the pressurization mode to provide breathing and pressurization air.
The recirculation operation of two fan / filter units is required to mitigate the effects of ingress and egress of the control room during the accident scenario. NUREG-0800 assigns a continuous in-leakage value of 10 cfm to account for this access.
6 l
l l.
l 3.0 RECOMMENDED SYSTEM MODIFICATIONS Based on the values in NUREG-0800, a 10 cfm continuous in-leakage was assumed to account for ingress and egress from the pressure envelope during accidents.
To achieve an acceptable level of. filtration of contaminated' air associated with ingress 'and egress from the pressure envelope, operation of two fan / filter units of the emergency ventilation system-in the recirculation mode is recommended for the duration of each accident scenario.
.The additional amount of recirculation (and associated filtering) capacity reduces the amount of airborne contamination - in the control room, to ensure that control room doses resulting from ingress and egress will not exceed the limits of. GDC-19.
The North Anna operating procedures have been revised to incorporate this change.
The dose calculations addressed in Section 4 of this report also assumed that the emergency ventilation recirculation operation is initiated concurrently with the actuation of the bottled air system on SI.
Previously, recirculation was not used in conjunction with the bottled air system. Therefore concurrent initiation of the emergency ventilation system and bottled air system is not currently an automatic action, and modification of the emergency fan start logic will be required.
In the interim, activation of the emergency ventilation recirculation system (two fan / filter units) will be manually initated upon SI. The North Anna Emergency Operating Procedures have already been revised to require this operator action.
Dose calculations were also performed for the two most 7
limiting accidents -(main steam line break and steam generator tube rupture) to. evaluate the adequacy of manual activation of the recirculation system until.the logic modifications are made.
The calculations' allowed a ten minute time delay for operator action in lieu of an automatic start signal.
The results of the calculations show"that the accumulated doses remain within the limits of GDC-19.
The fuel handling accident (FHA) dose calculations assumed that the control room normal supply is automatically isolated when the source term is released from the spent fuel pool.
This would require modification of the normal ventilation-isolation logic to automatically initiate isolation upon receipt of a high-high radiation alarm for the fuel building.
Such isolation is necessary.primarily because the radioactive sources of interest in the analysis of a fuel handling accident are noble gases.
Without the proposed modifications, a significant source of beta-activity can accumulate in the control room during a fuel handling accident due to the proximity of the normal ventilation suction to the discharge path of the noble gases to the environment.
In addition, the operation of the emergency ventilation system in recirculation will have minimal effect on filtration of the contaminated environment, because filtration is an ineffective means of removing noble gases, which are the primary sources in this type of accident. The recommended change to the normal ventilation isolation logic would minimize control room doses during a fuel handling accident by preventing the noble gas sources from entering the control room.
8 1
____.1______________________
J
As an interim measure, the North Anna operating procedures have been revised to require that the emergency ventilation system be operating, with the normal ventilation system isolated, whenever fuel is being handled in the fuel building.
This interim alignment has been reviewed and determined to be acceptable to maintain doses within GDC-19 limits.
Additional administrative controls are being considered for radiation monitoring of the control room to provide assurance that the actual in-leakage of contaminated air associated with control room access does not result in operator doses exceeding the limits of GDC-19.
9 L
f 4.0 REVIEW 0F CALCULATIONS AND ASSUMPTIONS L
The Condition IV accidents defined in Chapter 15 of the North Anna UFSAR are those which result in the most significant impact on control room habitability.
Specifically, the following accident conditions were addressed in this evaluation:
i (1)
Loss of Coolant (LOCA)
(2) Main Steam Line Breaks (MSLB)
(3) Fuel Handling Accident (FHA)
(4) Steam Generator Tube Rupture (SGTR)
(5)
Locked Rotor Accident (LRA)
Calculations.were performed to determine the'30 day dose to control room inhabitants.
These calculations were based on models and methodology which are consistent with NUREG-0800.
The input assumptions and results for each of these five accidents are summarized below. In general, the analyses considered the pressurization of the control room envelope by bottled air and coincident operation of the emergency ventilation system filters in the recirculation mode.
Normal ventilation was assumed prior to isolation, and the time of isolation actuation was accident dependent.
After depletion of the bottled air system (60 minutes after the control room was isolated), both pressurization and recirculation were provided by the emergency ventilation system for the remainder of each accident scenario.
Consistent with Part 6.4 of the Standard Review Plan (NUREG-0800), an 10
)
l l
unfiltered in-leakage at 10 cfm was assumed for the duration of the accident to account for control room access.
1.
Loss of Coolant Accident (LOCA)
The LOCA' evaluation was based on methodology defined in Regulatory Guide 1.4 and NUREG-0800.
Analysis of the consequences of this i
accident assumed that releases start at the moment the Safety Injection (SI) signal is initiated. The control room is automatically isolated, and the bottled air system is actuated, on an SI (that is, at T = 0.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />).
Transport time to the control room ventilation system intakes was conservatively treated as negligible (transport time = 0.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />).
The evaluation considered releases from the containment to the atmosphere for a duration of one hour, by which time the containment pressure was assumed to have been reduced to subatmospheric. The analysis further assumed the bottled air system provides positive pressure for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
At the end of this time, a filtered air intake of 1000 cfm was assumed, with the following filter efficiencies:
l a.
90% removal of elemental iodines b.
0% removal of methyl iodines c.
99% removal of particulate These values are conservative compared to the ratvl efficiencies of 99%, 85%, and 99.5%, respectively. A 0% removal efficiency was also 11
i I
L assumed for noble gases.
As noted previously, a 10 cfm unfiltered in-leakage was assumed for the duration of the accident (starting at
.T = 0.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />) to account for the effects of multiple entries of the.
control room envelope during the accident. The emergency ventilation system was assumed to provide a recirculation rate of 2000 cfm (i.e.,
no equipment modifications), and to be activated on bottled air actuation, at T = 0.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />.
A nominal core power level of 2900 MWt' was used.
A 2% instrument error was added, per Regulatory Guide 1.49, which resulted in a core power of 2958 MWt being used for the calculations. The core inventory used for this evaluation was taken from Table 11.1-1 of the North Anna UFSAR.
The doses received due to ECCS leakage, to leakage of particulate from the containment, and to iodine leakage from containment were recalculated.
The doses due to direct shine from the cloud and the containment (calculated previously, and unchanged by the new assumptions of the current evaluation) were then added to determine the total control room doses.
The cumulative doses received in the i
control room at the end of 30 days were therefore determined to be:
12
~,
~ Allowable limits
- Thyroid:
11.3 rem 30' rem Whole Body Gamma:
0.386 rem 30 rem The North Anna UFSAR currently reports that the' dose to control room personnel for this accident does not exceed 2.5 rem whole body or 10 rem to the thyroid. These newly calculated control room doses exceed the doses previously calculated for the LOCA, most obviously for.the
. thyroid, but are still clearly within the allowable limits.
2.
Main Steam Line Break (MSLB)
The analyses of the main steam line break accident were based on guidelines provided in part 15.1.5 of the Standard Review Plan, NUREG-0800.
The break was assumed to be a 1.4 square foot double ended rupture, occurring outside containment, in the. turbine building. An accident duration (in terms of releases) of 8 days was assumed,.as was a coincident loss of offsite power.
The primary coolant and secondary side concentrations used for the analysis were based on the plant Technical Specifications.
All iodines in the affected steam generator were conservatively assumed to be released to the environment (PF = 1.0).
For the unaffected steam generators, only 1% of the iodines which may be present to the
13 t-_-___
secondary side (due ' to primary-to-secondary system leakage) were assumed to-be released to the atmosphere (PF = 0.01).
The iodine partition factor for the unaffected steam generators is consistent
' with the value used previously for analysis of the MSLB for North Anna, and is given in part.15.1.5 of the Standard Review Plan.
The
~
mass and energy releases used for the analysis were developed for a 1.4 square foot double ended rupture occurring in the containment.
These data were determined to be conservative for use in analysis of a break in the turbine building, which' would be downstream of the isolation valves.
The radionuclides concentrations in the released steam were also conservatively basod on the properties of.the steam inside the pipe:
no credit was taken for the effects of the steam expansion upon release to the turbine building.
As for the LOCA analysis, the MSLB analysis assumed that the reactor trip and initiation of releases both occur on generation of the' SI actuation signal, at T = 0.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />.
The transport time of the releases to the control room ventilation intakes was assumed to be negligible (T = 0.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />).
The SI signal automatically initiates the isolation of the control room and activation of the bottled air system.
The bottled air pressurizes the control room for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, after which time a filtered air intake of I'00 cfm was assumed.
A 0
control room filter efficiency of 99% removal of elemental iodines was assumed, in accordance with the guidelines in Regulatory Guide 1.52. Once again, a 10 cfm unfiltered in-leakage was assumed for the duration of the accident, starting at T = 0.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />, to allow for the effects of multiple entries to the control room envelope during the 14
l accident. The control room recirculation rate was assumed to be 2000 cfm3 for the duration of the event, which is based. on the use of t
existing equipment.
Doses' were calculated for two cases: (1) a pre-accident iodine' spike to 60 times the normal concentration, and (2).a concurrent (accident initiated) iodine spike. The calculated doses included releases from both the affected and unaffected steam generators.
No failed fuel was assumed for the MSLB accident; that is, no (additional) fuel failures were assumed.to occur due to this accident.
The cumulative doses received by personnel in the control room at the
-end of 30 days were calculated to be:
Pre-accident Concurrent Allowable Iodine Spike Iodine Spike Limits Thyroid:
24.0 rem 22.3 rem 30 rem Whole Body Gamma:
1.41E-3 rem 1.24E-3 rem 5 rem Beta (Skin):
2.78E-2 rem 2.41E-2 rem 30 rem The North Anna UFSAR currently reports that the dose to control room personnel for this accident does not exceed 2.5 rem whole body or 10 rem to the thyroid.
The new doses calculated for the control room during a MSLB accident exceed the previously calculated doses, particularly for the thyroid, but are still within the allowable limits based on 10CFR50 Appendix A, GDC-19.
15
1 9
IT 3.
Fuel Handling Accident (FHA)'
The evaluation of the fuel handling accident assessed the cumulative doses received by control room personnel for 30 days following a fuel handling accident in. the spent fuel pit.
Analysis of the. fuel handling accident was based on guidance provided in part 15.7.4 of.
the Standard Review Plan (NUREG-0800) and in Regulatory Guide 1.25 (Safety Guide 25).
For evaluation of this accident, the fuel was assumed to have operated at a core power level of 2958 MWt (nominal core power of 2900 MWt plus a 2% instrument error), and the radial peaking factor for the damaged fuel assembly'was assumed to have been 1.65.
(This is higher than the 1,55 currently allowed by the Technical Specifications, and will result in a conservatively high. inventory in the fuel.') 'The fuel assembly with the greatest amount of activity (e.g.,
the highest burnup assembly) was' assumed to be damaged during the FHA.
The FHA was conservatively assumed to occur. 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after. reactor shutdown,* which is consistent with previous analyses of the fuel handling accident in the North Anna UFSAR. Per Regulatory Guide 1.25, the gap activity in the damaged fuel assembly is assumed to consist of 10% of the iodines, 10% of the noble gases (except Kr-85) and 30%
of the Kr-85 in the assembly. One hundred percent of the gap activity was assumed to be released to the spent fuel pool during the FHA
- Technical Specification 3/4.9.3 precludes removal of the reactor head until 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> after shutdown, so this analysis - which assumed 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> less decay time - is conservative.
16
=
l
]
l (i.e.,'all fuel rods in the assembly are damaged). The minimum depth of-the water between the-top of the damaged fuel rods and the fuel pool surface was 23 feet. All noble gases released from the damaged fuel assembly were released to the atmosphere, but only 1% of the i
iodines released from the assembly were released from the pool (PF =
0.01). The iodines which are released to the atmosphere were assumed to consist of 75% elemental iodines and 25% methyl iodines.
The atmospheric releases are assumed to continue for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, with a negligible transport time from the spent fuel pool to the control room ventilation intakes. The activity is assumed to be released through the iodine filters, since fuel building ventilation is filtered during any fuel handling in the fuel building.
When fuel is being moved, the control room is currently ventilated via the emergency ventilation system and the fuel building exhaust is filtered. This analysis, however, took into account a modification being implemented where the control room is automatically isolated at' the start of a fuel handling accident due to a high radiation alarm from the fuel building radiation monitors.
The bottled air system was then assumed to be manually actuated by an operator 10 minutes after the FHA occurs, in addition to control room air recirculation (2000 cfm). The bottled air system provides envelope pressurization for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, after which time a filtered air intake of 1000 cfm is initiated. A 10 cfm unfiltered in-leakage was assumed to exist during the control room isolation period (starting at T = 0.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />) and when the emergency ventilation system was in use, to account for multiple entries of the control room envelope during the accident.
17
p t-l
. The filter efficiencies assumed for this evaluation were as follows:
l a.
Control room ventilation filters:
- 1) 99% removal of elemental iodines-
'2) 99% removal of methyl iodines b.
Fuel building exhaust filters:
- 1) 90% removal of elemental iodines
- 2) 70% removal of methyl iodines The fuel. building exhaust filter efficiencies are consistent with Section 6.4.1.1 of the North Anna UFSAR, 'and are more conservative than ~the nominal-efficiencies-required by the Technical Specifications (99% removal of both elemental and methyl iodines).
The integrated doses in the control room for 30 days following a fuel handling accident were calculated to bc Allowable limits Thyroid:
1.67 rem 30 rem Whole Body Gamma:
5.73E-3 rem 5 rem Beta'(Skin):
4.52E-1 rem 30 rem The North Anna UFSAR currently reports that the dose to control room personnel for a FHA does not exceed 2.5 rem whole body or 10 rem to 18
--____-_L_L---___-_---__-__.-.
the thyroid.
These new doses continue to meet those limits, and 1
clearly fall within the allowable limits prescribed in SRP 6.4. (i.e.,
l based on 10CFR50 Appendix A, GDC-19).
4.
Steam Generator Tube Rupture (SGTR)
The SGTR analysis followed the methodology presented in part 15.6.3 l
of'the Standard Review Plan, NUREG-0800.
The break was assumed to i
be a double ended rupture, occurring near the top of the 1 steam l
generator tube bundle, so that the break may be exposed if the liquid level in the steam generator drops below ' the top of the steam generator tube bundle. The defective steam generator was assumed to be isolated within 30 minutes of the accident, which is consistent with previous analyses of the SGTR, as discussed in Section 15.4.3 of the North Anna UFSAR. The main condenser was also assumed not to be available for steam dump due to coincident loss of offsite power.
The primary and secondary side concentrations used for this analysis were based on the plant Technical Specifications.
The releases to the atmosphere are assumed to start when the secondary side pressure exceeds the PORV setpoint at T=5 minutes. Once again, transport time from the site of the releases to the control room ventilation intakes was considered to be negligible (transport time = 0.0 minutes). The steam generator tube bundle was assumed to remain covered until the reactor trip, at 5 minutes after the accident. The evaluation then assumed that the tubes were uncovered for 10 minutes, which is consistent with the uncovery period in the most recently approved SGTR 19
l 1
1 1
i-y
' analysis.for North Anna.
For consistency with the approved SGTR analysis, the iodine partition factors were assumed to be 0.01 (i.e...
1% of the iodine was released from the steam generators) for the unaffected. steam generators as well as for the affected steam generator whenever the tube bundle was covered. During the 10-minute period of tube uncovery, a conservative iodine partition factor. of 1.0 was used for releases from the affected steam generator.
For the SGIR, the control room is isolated and the bottled air system is actuated automatically on an SI signal.
Based on the actual sequence of events for the North Anna SGTR in 1987 (Reference 1), this was assumed to occur 6.5 minutes after the tube rupture actually-occurred.
As significant releases to the atmosphere do not start until 5 minutes after the tube ruptures, the total length of time after releases start during which only unfiltered air enters the l~
control room is only 1.5 minutes. The filtered recirculation of.the control room air is currently manually activated. A 10 minute delay (from initiation of the SI) was assumed for this operator action.
The bottled air supplies the control room for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, at which time a filtered air intake of 1000 cfm was assumed, in addition to the 2000 cfm recirculation.
The control room filters were assumed to be 99%
effective for the removal of iodines, in accordance with the guidelines in Reguiatory Guide 1.52.
A 10 cfm unfiltered in-leakage was assumed during the control room isolation period and when the emergency ventilation system was in use, to account for multiple entries of the control room envelope during the accident.
20 E
1 L
?
Doses were calculated for.two cases: -(1)apre-accidentuiodine' spike to 60Ltimes the normal concentration, and (2) a concurrent (accident-initiated) iodine spike, which was assumed.to last for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
No additional ' fuel failures were assumed to occur due to the SGTR accident.
The cumulative doses received by personnel in the control room at the end of 30' days were. calculated to be:
Pre-accident Concurrent Allowable Iodine Spike Iodine Spike Limits Thyroid:
19.0 rem 1.73 rem-30 rem Whole Body Gamma:
1.16E-2 rem 1.07E-2 rem 5 rem Beta (Skin):
5.69E-1 rem 5.61E-1 rem 30 rem An additional case was also considered which differed only in the time at which the. control room recirculation fan was started.
For this
- additional case, it was assumed that recirculation was automatically initiated upon activation of the bottled air system on SI.
This reflects the control room ventilation system configuration which should be in place at North Anna by the end of 1989.
For this case, the cumulative control room doses at the end of 30 days were determined to be:
21
.A 3
4
^
-Pre-accident Concurrent' Allowable Iodine Spike
. Iodine Spike
_ Limits r
Thyroid:
16.5:
rem l'.55 rem 30 rem-Whole Body Gamme:
1.15E-2 rem 1.07E-2 rem 5 rem Beta (Skin):
5.68E-1 rem 5.61E-1 rem 30 rem The North. Anna UFSAR currently reports only'that the dose to control room personnel for the'SGTR does not exceed 2.5 rem whole body or 10 rem to the thyroid.
The new doses calculated for the control room o
during a SGTR accident exceed the previously calculated doses, particularly for the thyroid, but clearly fall within the allowable
' limits based on 10CFR50 ppendix A, GDC-19.
5.
Locked Rotor Accident (LRA).
The analysis of the locked rotor accident was based on-guidelines provided in part 15.3.3 of the Standard Review Plan, NUREG-0800. For this evaluation, 13% of the fuel was assumed to fail as a result of this accident condition.
This is consistent with the basis for the UFSAR Chapter 15 analysis of the LRA for the North Anna uprating, as-discussed in Section 15.4.4.2.7 of the UFSAR.
Also, a relief valve was assumed to be stuck open on one steam generator, and offsite power was lost at the beginning of the accident.
Two cases were analyzed:
the first assumed that the tubes in the affected steam generator were uncovered for a period of time during the accident, and the second assumed that the tubes in the affected steam generator remained covered for the entire duration of the analysis.
22
4 -
L For consistency with the LRA analysis for the North Anna uprating, it was assumed that it takes'15 minutes to isolate the releases from the affected steam generator. Plant cooldown by the secondary system after the accident requires 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
The gap activity from the failed fuel was instantaneously released into.the primary coolant.
Per Regulatory Guide 1.77, the total gap activity is equal to 10% of the total core activity for noble gases and iodine.
Therefore, for this analysis 13% of the total gap activity was released to the primary coolant.
A primary coolant preaccident iodine spike to 60 times the normal concentration was also assumed for the LRA evaluation.
Because offsite power was assumed to be lost at the beginning of the accident, the condenser was not available for steam dump, and all releases were directly to the atmosphere.
For the first case, the steam generator with the stuck open reliefd valve is assumed to go dry in a very short time, with tube uncovery conservatively assumed to occur at the beginning of the accident (T
= 0.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />). It was therefore assumed that the total mass of liquid and steam in the steam generator (i.e., all iodines in the affected steam generator) were released to the environment.
The feedwater which flows to the affected steam generator during the 15 minutes it takes to isolate the stuck open relief valve was &lso assumed to be released. For the unaffected steam generators, only 1% of the iodines which were released from the primary coolant were assumed to be 23
y En released to the atmosphere (iodine partition factor of 0.01). These
' iodine ' release fractions are consistent with those used in' the SGTR and MSLB accident evaluations. The second case assumed that the steam generator tube bundle remains covered throughout the LRA. An iodine partition' factor of 0.01 was therefore used for the affected steam generator for the duration of the LRA.
For thel locked rotor accident, the analysis assumes that the control room is automatically isolated, and the bottled air supply is automatically actuated. at T = 0.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />, on generation of the SI signal.
Initiation of releases was also assumed to occur at T = 0.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />, and the transport time of the releases to the control room ventilation system intakes was assumed to be negligible. The bottled air pressurizes the control room for I hour, after which a filtered air intake of 1000 cfm in addition to 2000 cfm recirculation was assumed for the remainder of the 30 day evaluation time. Control room filter efficiencies of 99% removal of elemental iodines were assumed-for both the pressurization and recirculation modes.
A 10 cfm unfiltered in-leakage was assumed for -the entire duration of the evaluation to incorporate the effects of multiple entries to the control room envelope.
The cumulative doses received by personnel in the control room at the end of 30 days were calculated to be:
24
With Tube Without Allowable Uncovery Tube Uncovery Limits Thyroid:
7.71 rem 4.42 rem 30 rem Whole Body Gamma:
0.260 rem 0.260 rem 5 rem Beta (Skin):
l i
The North Anna UFSAR currently does not report a dose to control room I
personnel for the LRA. The new doses calculated for the control room during a locked rotor accident fall within the allowable limits prescribed in SRP 6.4 (i.e., based on 10CFR50 Appendix A, GDC-19).
The results of the dose calculations are summarized in Table 1.
i i
j 25
e Table I h
30-Day Doses In The Control Room.
Thyroid.
Gamma Beta Accident-
. Dose (Rem)-
Dose (Rem)
. Dose (Rem)-
LOCA 11.3 0.781
-0.386 MSLB with pre-accident 24.0 1.41-3*
2.78-2 I spike MSLB with concurrent 22.3 1.24-3 2.41-2 I spike FHA 1.67 5.73-3 0.452 SGTR with pre-accident
'I spike w/recirc at 6.5 min.
16.5 1.15-2 0.568 w/recirc at 16.5 min.
19.0 1.16-2 0.569 SGTR with concurrent I spike w/recirc at 6.5 min.
1.55 1.07-2 0.561 w/recirc at 16.5 min.
1.73 1.07-2 0.561 LRA with tube 7.71 0.260 4.38 uncovery LRA without tube.
4.42 0.260 4.38 uncovery L
Allowable Limits 30.0 5.0 30.0 l
- 1.41-3 = 1.41x10-3 26
l-
{
4 l
5.0
SUMMARY
AND CONCLUSIONS The doses to control room personnel during accident scenarios at North Anna have been reevaluated to incorporate a 10 cfm unfiltered in-leakage for the duration of the accident.
This unfiltered air was assumed to conservatively incorporate the effects of procedurally required exits and entries to the control room pressure envelope during the accident.
Previous control room dose calculations had considered the control room to be isolated, with no doors opening, during the first hour after an i
accident occurred.
An acceptable means of allowing required entries to the control room pressure envelope during the first hour of an accident has been identified. This involves several modifications of the North Anna control room habitability system, which have been recommended to ensure that doses to control room personnel do not exceed the limits of GDC-19.
These modifications include:
1.
Operation of two fan / filter units of the emergency ventilation system in the recirculation mode for the duration of the accident.
2.
Initiation of the emergency ventilation recirculation concurrent with actuation of the bottled air system (i.e., on SI).
3.
Initiation of control room isolation on receipt of a fuel building high-high radiation alarm.
The dose calculations assumed that these modifications were made.
For those cases where the system modifications have not yet been implemented, it was also verified that intermediate measures identified in Section 3 of this evaluation are adequate to ensure that the requirements of GDC-19 will be satisfied.
All accident scenarios that can have a significant impact on control room operator doses have been evaluated and determined to meet the limits specified by GDC-19.
l l
I i
l l
l l
l 28
1 6.0 10 CFR 50.59 EVALUATION This safety evaluation has described the effects of incorporating a 10 cfm unfiltered in-leakage into the control room dose calculations to address the effects of required access to the control room during control room isolation.
System modifications have been identified which will mitigate the effects of this contaminated air entering the control room.
Even with the emergency ventilation system modifications, the change in the input assumptions of the dose calculations causes the predicted control room dose consequences to increase with respect to those reported in Section 6.1.4.3.1 of the UFSAR.
On this basis, it is therefore concluded that an unreviewed safety question exists as defined in 10 CFR 50.59, and the results should t'^
submitted to the NRC for approval. The results of this evaluation can be stated as follows:
1.
No increase in the probability of occurrence of an accident will occur due to the change in the input assumptions or to the emergency ventilation system modifications.
The effects of these changes on the control room dose calculations have been evaluated. The changes in the input assumptions of the dose calculations (including appropriate changes for the emergency ventilation system modifications) result in predicted control room doses which increase from the doses currently reported in Section 6.1.4.3.1 of the UFSAR.
Therefore, an unreviewed safety question technically exists as defined in 10 CFR 50.59.
Although the predicted consequences of an accident will increase for control room personnel, the allowable dose 29
f Lf li.mits defined in-GDC-19 of Appendix A of 10 CFR.50 are still met for each accident. There is no impact on the offsite doses.
2.
No~ new accident types or equipment malfunction scenarios are
-introduced ?-as a result of these changes.
Therefore, there is no possibility. of an accident of a different type than any previously evaluated in the UFSAR.
3.
The margin of safety is-not reduced. An evaluation of the accidents which have the greatest impact on control room dose calculations was performed. It ha's been concluded that the acceptance criteria defined by GDC-19 will continue to be met,'both with the recommended changes to the emergency' ventilation system, and for the interim system modifications and procedure changes.
L 30 Y..
)
j 7.0 BASIS FOR NO SIGNIFICANT HA7.ARDS DETERMINATION
-)
It has also been determined that the changes to the control room emergency ventilation system.to permit access to the control room during~ accidents
-do not involve a significant hazards consideration as described in 10 CFR 50.92. The results of this determination can be stated as follows:
1.
The change does not involve a significant change in the probability or consequences of an accident previously evaluated.
There are no system changes which increase the probability of an accident occurring.
The effects on the analysis for each accident has been investigated, and the. doses to control room personnel were found to increase. This increase is not significant because tbs revised doses remain below the limits in GDC-19 of Appendix A of 10 CFR 50, and meet the guidelines of NUREG-0800 (Section 6.4).
2.
No new accident types or equipment malfunction scenarios will be introduced as a result of the recommended emergency ventilation system changes.
Therefore, the possibility of an accident of.a different type that any evaluated previously in the UFSAR is not created.
i l
4 3.
There is no significant reduction in the margin of safety.
The revised dost : calculations for all accidents continue to meet the appropriate GDC-19 limits for the recommended modifications to the emergency ventilation system. The GDC-19 limits will also be met for the interim system modifications and procedure changes.
31 j
I
8.0 REFERENCE 1.
Letter from W. L. Stewart to the NRC, " Virginia Electric and Power Company, North Anna Power Station Unit 1, Steam Generator Tube Rupture Event Report, Revision 1, " Serial No. 87-474A, September 15, 1987.
32 L____-___________.
1 k
G ATTACHMENT 2 PROPOSED FACILITY OPERATING LICENSE CHANGES NORTH ANNA POWER STATION UNITS I AND 2
PROPOSED LICENSE AMENDMENT The proposed changes to Facility Operating License No. NPF-4 to the Virginia Electric and Power Company for North Anna Power Station Unit 1 is shown on the next page.
l l
l l
e
- F.
The design of the reactor coolant pump and steam generator supports may be revised in accordance with the licensee's submittal dated November 6, 1986 (Serial No. 86-477A).
G.
The limiting dose to the control room operators shall be revised in accordance with the licensee's submittal dated March 1, 1989 (Serial No.89-022).
H.
This amended' license is effective as of the date of issuance and shall cxpire at midnight on April 1, 2018.
FOR THE NUCLEAR REGULATORY COMMISSION Originally Signed by R. C. DeYoung for Roger S. Boyd, Director Division of Project Management Office of Nuclear Reactor Regulation Attachments:
1.
Construction Related Itens to be completed prior to Initial Criticality 2.
Appendices A and B Technical Specification page changes 3.
Figure 1 4.
Table 1 Date of Issuance: APR 1 1978
.--,-.------------------r----.
-- ---- -----.,. - - - - - - -. - - - - -, - - - -., - -. - -, - - - - - - - - - - -. - - - - _ - - _ - -, - - -, - -., - - - - - _ - - -, - - -. -., - - - - - - - - -. -. - - ~ _ -., - -. -., - - - -. - -. - - - - - - -, - -. -. _, - _ - -
PROPOSED LICENSE AMEN 0 MENT l-The proposed changes 'to Facility Operating License No. NPF 7 to the Virginia Electric' and Power Company. for North Anna Power Station Unit 2 is shown on the next~page.
i I
i 1
i l
1
)
i I
1 I
l
)
o I.
The limiting dose to the control room operators shall be revised in accordance with the licensee's submittal dated March 1,1989 (Serial No.89-022).
J.
This license is effective as of the date of issuance and shall expire l
August 21, 2020.
FOR THE NUCLEAR REGULATORY COMMISSION Original signed by Harold R. Denton Harold R. Denton, Director Office of Nuclear Reactor Regulation
Attachment:
Appendices A & B Date of Issuance: AUG 21 1980 i
_ -.