ML19257D427
| ML19257D427 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 01/31/1980 |
| From: | Crouse R TOLEDO EDISON CO. |
| To: | Reid R Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0578, RTR-NUREG-578 TAC-12453, NUDOCS 8002040376 | |
| Download: ML19257D427 (10) | |
Text
I Docket No. 50-346 License No. NPF-3 Serial No. 585 January 31, 1980 TOLEDO u EDISON Director of Nuclear Reactor Regulation Attention:
Mr. Robert N. Reid, Chief Di.aO P C AOUaE Operating Reactors Branch No. 4 7"
Division of Operating Reactors auw United States Nuclear Regulatory Commission Washington, D. C.
20535
Dear Mr. Reid:
Toledo Edison ec=mitted in our letter of November 21,1979 (Serial No. 559) to conduct a preliminary analysis of current shielding ef fectiveness of the sampling, letdown and accident mitigation systems for the Davis-Besse Nuclear Power Station, Unit 1 (DB-1). This would be done with the assumed source terms referenced in your letter dated October 30, 1979 (Log No. 454) concerning recommendation 2.1.6.b of NUREG 0578 "TMI-2 Lessons Learned Task Force Status Report and Short Term Recommendations".
The analysis was evaluated on the basis of operator accessability requirements.
This letter transmits by attachment a description of the analysis and the results of the shielding evaluation.For the control room and the interim Technical Support Center, the results of this evaluation indicate that the direct dose rate f rom the contaminated systems in the study is less than 15 millirem per hour, allowing continuous occupancy of these areas.
For areas requiring infrequent operator access, the study was done incorporating two planned modifications not yet completed. These include:
- 1) Installation of motor operators with control and position indication in the control room for valves DH63 and DH64. This modification provides for control room series alignment of the decay heat and high pressure injection pumps if required. This modification is scheduled to be completed during the upcoming spring refueling outage as required by the Davis-Besse Facility Operating License Condition 2.C(3)(g).
- 2) The addition of a remote reactor coolant sampling syst2m which allows utilization of a shielded sampling f acility physically separated from the present sample station. This will minimize temporary shielding and extensive personnel radiation exposure pre-planning prior to sampling at the normal location in case of extremely high activity in the reactor coolant system.
The results identified acceptable operator accessability to infrequently and con-tinuously occupied areas.
Therefore, with the exceptions of the two items identified above, no additional shielding modifications are proposed for the Davis-Besse Nuclear Power Station, Unit 1.
Very truly youta,
) n H /.' v #. & j/,>
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-r ds i865 239 RPC:TJM: cts THE TOLEDO EDISON COMPANY EDISON PLAZA 300 MAO! SON AVENUE TOLECO OH!O 43652 h
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Docket No. 50-346 License No. NPF-3 Serial No. 585 January 31, 1980 DESIGN REVIEW OF PLANT SHIELDING FOR SPACES OUTSIDE CONTAINMENT MIICH MAY BE USED IN POST ACCIDEFT OPERATIONS FOR DAVIS-BESSE NUCLEAR POWER STATI'JN, UNIT 1
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Table of Contents I.
Introduction II.
Scope of Design Review A.
Systems Engineering Methodology B.
Shielding Design Review Methodology D.
Personnel Exposure Limits and Methodology III. Results of Review e
t 4
. I.
INTRODUCTION This report describes the design review of plant shielding of spaces for post accident operations for Davis-Besse Unit 1.
Systems required to process primary reactor coolant outside the containment during post accident conditions were selected for evaluation.
Large radiation sources were postulated to be present in the selected systems.
Areas which are vital for post accident occupancy were evaluated to determine if access and performance of required operator activities might be unduly impaired due to the presence of the postulated radiation source in these systems.
As a byproduct of this review, a number of radiation zone maps and associated curves have been produced which will alert operational personnel to potential radiation levels in the plant at various times following the accident.
II.
SCOPE OF DESIGN REVIEW A.
Systems Engineering Methodology 1.
Selection of Systems for Shielding Review The criteria applied in selection of plant systems used in the shielding review results in several classifications of systems selected for various reasons as discussed below.
Category A (Recirculation Systems)
The first group of systems are those systems required by p ant design to mitigate a design basis loss of coolant accident and which might contain highly radioactive sources in excess of the current design basis.
The emergency core cooling systems were designed to mitigate the conse-quences of a loss of coolant accident and prevent extensive core damage as required by 10CFR50 Appendix K.
Nevertheless, for the purposes of this study these systems were postulated to contain significant additional sources of radioactivity above and beyond the original plant design basis for the purposes of the required shielding review.
These plant systems are designed to be used following a loss of coolant accident. Operators are trained to respond to a loss of coolant accident by using the emergency core cooling systems. Procedures are written to specify the proper use of these systems by plant operators following an accident.
In summary, the emergency core cooling systems are designed and expected to operate following a loss of coolant accident to prevent significant core damage.
If a significant radioactive source is to be considered above and beyond the normal plant design basis as has been recommended by NRC, then a first priority safety concern is to ensure that operation of these systems containing a significant source will not adversely impact operator functions required outside the containment.
Therefore, the following systems have been selected to ensure that this first priority safety concern is adequately addressed by the existing plant shielding design:
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. o Those portions of the containment spray systems used to recirculate water from the containment vessel emergency sump back into the containment vessel.
o Those portions of the decay heat removal system used to recirculate water from the containment vessel emergency sump back into the containment vessel.
o Those portions of the high pressure injection system used to recirculate water from the low pressure injection system back into the containment vessel.
Category B (Extensions of Containment Atmosphere)
In addition to systems listed above, there are other systems or portions of systems which would contain radioactivity by virtue of their connection to the containment following an accident. Proper operation of the emergency core cooling system would prevent extensive core damage and mean that these systems would not be expected to contain the significant radioactive sources required by this special analysis.
Nevertheless, such sources have been postulated in the following systems for this study:
o Those portions of the containment ventilation systems external to the containment up to the first closed isolation valve,
which could contain the atmosphere from the containment.
o Those portions of the sampling systen used to obtain a containment atmosphere sample.
Category C (Liquid Samples)
Lessons Learned Task 2.1.8 requires that certain post accident liquid samples be obtained from the reactor coolant system.
Those portions of the sampling system which must be used to meet the intent of Task 2.1.8 were selected for this shielding review.
Category D (Letdown)
There is no reason to operate additional systems while the reactor coolant system is contaminated to the significant levels required as part of the shielding design review. Since there is no systems engineering logic for mechanistically analyzing either the letdown and makeup system or the waste gas system, an arbitrary, nonmechanistic assumption has been selected to be responsive to the NUREG 0578 recommendation Therefore,the following portion of the letdown system has been selected for analysis:
o That portion of the letdown system from the reactor coolant system past the failed fuel detector up to the inlet valves to the purification demineralizers.
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. 2.
Quantification of Potential Radioactive Source Release Fractions The following release fractions were used as a basis for determining the concentrations for the shielding review:
o Source A:
Containment atmosphere: 100% noble gases, 25% halogens o Source B: Reactor coolant: 100% noble gases, 50% halogens, 1% solids o Source C:
Containment sump liquid: 50% halogens,1% solids The above release fractions were applied to the total curies available for the particular chemical species (i.e. noble gas, halogens, or solid) for an equilibrium fission product inventory for a light water reactor.
The Fegulatory Guide 1.7, solids release fraction,1%, was used in this review. No noble gases were included in the containment sump liquid (Source C) consistent with Regulatory Guide 1.7.
Cursory analyses have indicated that the halogens dominate all shielding requirements and that contributions to the total dose rates from noble gases are negligible for the purposes of a shielding. design review.
3.
Source Term Models Section 2 above outlines the assumptions used for release fractions for the shielding design review.
These release fractions are, however,' only the first step in modeling the source terms for the activity concentrations in the systems under review.
The important modeling parameters, decay time and dilution volume, obviously also affect any shielding analysis. The following sections outline the rationale for the selection of values for these key parameters.
a.
Decay Time For the first stage of the shielding design review process, no decay time credit was used with the above releases.
The primary reason for this was to develop a set of accident radiation zone maps normalized to no decay that could be used as a tool by the plant staff along with a set of decay curves to quantitatively assess the in-plant dose status quickly following any abnormal occurrence.
For identifying problem areas, however, the following decay times were used in assessing anticipated potential personnel radiation exposure due to those operator actions that may be required post LOCA.
For analyses of personnel exposures in vital areas outside the control room, radioactive decay of one hour was considered.
This is consistent with the NRC recommendation to obtain a reactor coolant sample one hour af ter an accident, which is the only short term evolution outside the control room.
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. Additional decay time was also allowed for the review of all those ECCS systems previously outlined in Section II that are used to recirculate water from the containment sump back into the containment. That decay time was 40 minutes which is consistent with the minimum time for initiation of recirculation as per FSAR Section 15.4.6.4.
b.
Dilution Volume The volume used for dilution is important, affecting the calculations of dose rate in a linear fr.shion.
The following dilution volumes were used with the release fractions and decay times listed above to arrive at the final source termr. for the shielding reviews:
o Source A:
Containment free volume. The volume occupied by the ECCS water was neglected.
o Source B: Reactor coolant system volume based on reactor coolant density at the operating temperature and pressure.
o Source C: The volume of water present at the time of recirculation (Reactor coolant system + borated water storage tank +
core flood tanks).
c.
Sources Used in Piping and Equipment for Each System Under Review In defining the limits of the connected piping subject to ' contamination listed below, normally shut valvec were assumed to remain shut.
o Cont 11nment s' pray system - At the initiation of recircul'ation, Source C was used.
High pressure injection system - At the initiation of recirculation, o
Source C was used. This assumes that the HPI and LPI are cross connected.
o Decay beat removal system - Source C was used for the sump recirculation mode.
o Sampling systems - The sources used in the shielding design review for sampling systems were as follows:
Containment air sample -
Source A Reactor coolant sample -
Source B o Letdown system - The liquid source was Source B.
B.
The Shielding Design Review Methodology 1.
Analytical Shielding Techniques The previous sections outlined the rationale and assumptions for the selection of the systems that would undergo a shielding design review as well as the formulation of the sources for those systems.
The next step in the review process was to use those sources along with standard point kernel shielding analytical techniques to estimate dose rates from those selected systems.
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. For compartments containing the systems under review, estimates were made for a general area dose rate rather than te superimpose the maximum dose rate at contact with the surfaces of all individual components of that system in the compartment.
For corridors outside compartments, reviews were done to check the dose rate transmitted into the corridor through the walls of ad-jacent compartments.
Checks were also made for any piping or equipment that could directly contribute to corridor dose rates, i.e. piping that may be run directly in the corridor or equipment / piping in a compartment that could shine directly into corridors with no attenuation through compartment walls. The source. term from containment vessel leakage which is contained within the negative pressure boundary was considered in the dose rates within and adjacent to the negative pressure boundary.
2.
Accident Radiation Zone Maps One of the two principal products of this review process is the series of accident radiation zone maps for Davis-Besse Unit 1.
These zone maps represent the correlation of the dose rates as estimated above with the required operator actions and resultant necessary accessibility to vital areas.
By using these zone maps along with the decay curves, potential problem areas were identified as noted in Section III.
The zone boundaries were formulated based on the following ationale:
Zone Designation Rationale D, Zone Dqse Rate Limits (Rem /hr)
A-I The first zone is consistent 0 / D A 0.015 with the personnel radiation exposure guidelines of Section II.C.2 for vital areas requiring continuous occupancy.
A-II The second zone is consistent 0.015JCDj$.0.1 with the personnel radiation exposure guideline of Section II.C.2 for vital areas requiring infrequent access or corridors to these areas. Such zones involve no time and motion evaluations.
A-III The third zone is consistent with 0.100 di D j$. 5 the personnel radiation exposure guidelines of Section II.C.2.
Zones in this range required that a tine and motion study be done to ensure that integrated exposure was not greater than 5 Rem as given in General Design Criteria 19, of 10 CFR 50, Appendix A.
The subsequent zones were selected by grouping them by powers of 10 so that rapid assassment of additional shielding measures could be used via " tenth value layers" of common shielding materials.
e
- Zone Designation D, Zone Dose Rate Limits (Rem /hr)
A-IV 5si D 6, 50 3,
A-V 50 /= D f 500 A-VI 500 d D ;f; 5000 A-VII 5000d? D g6 50,000 A-VIII 50,000d1Ds@,500,000 Note: These zone designations should not be confused with those used for the normal plant operation zone maps found in Chapter 12 of the Final Safety Analysis Report (FS AR).
C.
Personnel Exposure Limits and Methodology 1.
Access Those operator actions required post LOCA were reviewed to ensure that first priority safety actions can be achieved in the postulated radi' tion fields.
a This review ensures that access is available and required operator actions can be achieved except as noted in Section III.
In addition, sined other urspecified operator actions may be desired, the evolutions involved in a ncrmal shutdown sequence were reviewed as well as the general provisions for occupancy or access to the following key areas:
o Control room o System control panela o Emergency power supplies o Instrument areas Sampling and monitor areas per Task 2.1.8.a of NUREG 0578 o
Onsite technical support center per Task 2.2.2.b of NUREG 0578 o
Onsite operational support center per Task 2.2.2.c of NUREG 0578 o
2.
Personnel Radiation Expcsure Guidelines The general basis for persor nel radiation exposure guidelines was 10 CFR 50, Appendix A, CDC 19.
The following additional radiation limit guidelines were used to evaluate occupancy and accessibility of key plant a reas.
General area dose rates were used rather than maximum surface dose rates.
Contributions from all sources were considered.
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. Key areas requiring continuous occupancy.
o Key areas such as the control room and the onsite technical support center were verified to ensure the direct dose rate was less than 15 mr/hr at all times.
o Key areas requiring infrequent access.
For these areas the dose rate was verified to be less than 5 R/hr when required for access.
For dose rates greater than 100 mr/hr, a man-rem calculation including time and motion analysis was performed to insure that the integrated exposure for an operator action aid not exceed 5 rem as required by GDC 19.
For dose rates less than 100 mr/hr, a man-rem calculation was not necessary.
III. Results of Review A review was made of the required operator tasks outside the control room following a loss of coolant accident to see if the doses received by plant personnel would exceed allowable limits.
The basis for this review was emergency procedure EP 1202.06,
" Loss of Reactor Coolant and Reactor Coolant Pressure".
All required operator actions were found to be accomplished from within the control room with the exception of aligning power to motor operated valves and verifying boron dilution flow rate.
These actions can be delayed until seven days after the start of emergency operations.
These actions are:
N 1.
Observe the auxiliary spray flow on flow indicator FI 4999 located in room 227 on Elevation 565.
- 2. Align emergency power from motor control centers MCCE11B in room 304 on Elevation 585 and MCCF11A in room 427 on Elevation 603.
Based on the seven day decay time the dose received in performing the above actions and traveling to and from the areas was well below the allowable limits.
Depending on the reactor coolant system leak rate, the HPI pumps may have to be operated in the " piggy back" mode of operation with the decay heat pumps.
In order to accomplish this evolution manual valves DH63 and DH64 must be opened. These valves are located in Rooms 115 and 105, respectively.
During the first refueling outage, as required by Operating License condition 2.C. (3)(g),
DH63 and DH64 will be modified to motor operated valves with control and position indication in the control room.
The post accident samples considered were the reactor coolant system, containment vessel atmosphere, and the station vent.
Time and motion analyses have been performed to Lnsure that the integrated exposure to an operator while obtaining post accident samples will not exceed 5 rem as required by GDC 19.
The zone maps and the c'ecay curves can be used to determine expected dose rates in various areas of the plant at any time af ter an accident.
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