ML19340B824
| ML19340B824 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee |
| Issue date: | 11/06/1980 |
| From: | Groce R Maine Yankee |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| WMY-80-147, NUDOCS 8011120361 | |
| Download: ML19340B824 (11) | |
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ENGINEERING OFFICE Ws ITBoRO, MASSACHUSETTS 01581 g_%
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W.s:~J MfY-80-147 B.3.2.1 November 6, 1980 M
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=k United States Nuclear Regulatory Commission
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Washington, DC 20555
-J Ud A*tention: Office of Nuclear Reactor Regulation C:
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References:
(a) License No. DPR-36 (Docket No. 50-309)
Us (b) USNRC Letter to All Operating Nuclear Power Plants,"
dated October 30, 1979 (c) MYAPC Letter to USNRC dated March 5, 1980, WMY 80-39
Dear Sir:
Enclosed herewith is the Maine Yankee shielding ctesign review including calculated dose rates, completed earlier this year part of the TMI accident requirements detailed in your letter, Reference f This report supersedes the preliminary shielding report submitted in Refe-
.a (c).
We trust that this information is acceptable to you; however, should you have any questions, please contact us.
Very truly yours, MAINE YANKEE ATCMIC POWER CCMPANY M
Robert H. Groce Senior Engineer - Licensing hMG/ncj Enclosure 41 i
S sommo361
4 MAINE YANKEE SHIELDING DESIGN REVIEW I.
INTRODUCTION The United States Nuclear Regulatory Comm.ission's rules and regulations require measures be taken to concrol radiation exposure to personnel associated with normal and abnormal plant operation. Subsequent to an accident which involves significant core damage such as experienced at TMI-2, certain systems and components important to post-accident operations may become more ra Jactive than was previously analyzed. As a result, environnental and huma,.,onsiderations utilized in the design of inis equipment may now be inadequate to cope with these increased radiatica levels; consequently, some equipment may become degraded and uneble to perform a required function and/or be inaccessible to plant personnel who must manually perform the require t operations.
In light of the above, NUREG-0578, Section 2.1.6.b recommends that licensees perform a dcsign review of plant shielding for spaces / systems which may be used in post-accident operations. This design review will identify the locations of vital areas which may be effected by high radiation fields as a result of an accident and/or post-accident plant operation.
Recommendations for design changes and corrective actions to allow for sufficient access to these areas are based on this review using the clarificatico of Section 2.1.6.b to NUREG-0578 (issued by H. Denton 10/30/79) as guidance. This report describes the work that has been done and summarizes the results o# this desig review.
II METHODOLOGY This preliminary review was done in two stages. The first stage consisted of a careful review of piping and macMnery layout drawings followeo by a survey of the plant areas containing radioactive systems. All piping systems that would contain radioactive recirculating fluids following an accident were identified and marked on the appropriate drawings. These drawings were then used as a guide in conducting the walking survey in the plant. The nlant survey resulted in a qualitative assessment as to what areat would or would not be accessible. Special attention was paid to areas such as the sample sink, and H2 monitoring station, which were obviously vital areas, and in close proximity to recirculating safety systems.
The second stcge consisted of reviewing all of the plants' Emergency Procedures to detennine if that procedure required an operator to enter any area containing recirculating fluid systens. A listing of each required operation was completed and evaluated. The evaluation determined whether additional shielding or remote operation would b. required, or whether additional analytical calcula-tions would be necessary.
i III. OPERATIONS REQUIRED POST-ACCIDENT A.
Discussion The shielding analysis was performed on systems that are required to mitigate the consequences of an accident. Analyses have not been performed on systems
that would be automatically isolated during in accident. However, the leak reduction program included systems that might be used for cleanup.
The original plant design had numerous trip valves which require local manual re-opening. These valves cannot be re-opened if core damage had occurred. Transients which resulted in safety injection actuation or containment isolation isolate all potential leakage paths outside of the containment except the systems required to mitigate an incident.
As part of the review for Iten 2.1.4, Maine Yankee is evaluating the desf r-ability of changing this local re-opening requirement.
In the event that modifications are made which alters this feature, a decision will be made as to whether a shielding review is appropriate. This review would be based on the source terms recommended by the NRC adjusted for length of time since the accident.
The following syste.ns were identified in the response to Item 2.1.6.a and have undergone a leak inspection program:
Waste Gas Systen Letdown System Purification System Charging System Emergency Core Cooling System Residual Heat Remova~i System Primary Sampling System Post Accident Purge System Contaiment Air Particulate and Gas Monitoring System Seal Water Supply and Ceturn The Emergency Core Cooling, Residual heat Removal and Post Accident Purge Systems are required to mitigate the consequences of an uccident. These systems were included in the shielding analysis as was the Primary Sampling System. The reasons for not including other systems are discussed below:
Letdown System Tne Letdown System is a nonessential system which isolates on a SIAS. Per hRC directive, this systen is included in the shielding analysis. The system layout is such that no additional shielding or equipment modifications are expected. Manual local re-opening of the containment isolation valve is not required.
Charging Systen Those portions of the Charging System at Maine Yankee which are part of the ECCS and puli water out of the contaiment are included in the shielding review.
Seal Water Supoly and Return The reactor coolant pump seal water supply system isolates on a SIAS. Seal water supply is not required for pump operation as long as component cooling water to the pumps is available. Main coolant pump operability is not required to mitigate the consequences of an accident.
The seal water return portion of the system is required for pump operation; however, the seal water return does not create a dose problem.
There is no source of leakage into the seal water system which is not isolated by the system's containment -isolation valves.
Containment Air Particulate and Gas Monitoring System This system is isolated -by a Safety Injection Actuation Signal and/or a Containment. Isolation Signal. This is a nonessential system which requires local, manual re-opening of the. containment isolation valves. These valves will be modified so that remote opening of the valves is possible.
Purification System and Waste Gas System The purification and waste gas systems were designed to process effluents
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from the' reactor coolant system assuming 1% failed fuel. An accident with the postulated core damage would require the construction of new facilities as is being done at TMI.
B.
System Operations - Problems and Solutions i
1.
Long Term Core Cooling Realignment
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A plant inspection of the four manually operated valves located in the PAB-Containment Tunnel has detennined that the installation of reach rods, remote operators, or shadow shielding is not possible.
Therefore, bypass piping will be installed. This piping will contain isolation valves so located as to allow installation of reach rods.
The reach rods will extend to a safe area of the 21' level of the PAB.
Operating procedures will be revised to reflect the new system design.
2.
Emergency Boration This procedure calls for manual operation of four valves in the lower level (high radiation area) of the Primary Auxiliary Building. Al ternate means exist for. meeting all single failure requirements without this system. The procedure will be changed to use alternate (controllable) valve lineups.
3.
Plant Cooldown i
Many procedures in this category are not required under accidcnt conditions. However, nine manual valves must be operated for Residual Heat Removal. Seven of these are located in the Safeguards Valve Operating Room. Two valves in the Safeguards Equipment Room will be in a very high radiation area and will have reach rods installed to allow operation from the Safeguards Valve Operating Room.
. h. Control Room Backup t-a.
Motor Control Center The Motor Control Center is located in a separate well-shielded
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cubicle just outside the containnent.
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Emergency Shutdown Panel The Emergency Shutdown Panel and the Inlet Valve to Post Accident Hydrogen Purge are located in a well-shielded cubicle just outside the containment.
5.
Primary Auxiliary Building a.
Primary Auxiliary Control Board The Primary Auxiliary Control Board area will be exposed to app'roxi-mately 0.7 r/hr from recirculated safety injection fluids at T=1 hour after a DBA.
b.
Emergency Safety Shutdown Panel The Emergency Safety Shutdown Panel area is in a separately shielded area in the PAB with a short ingress / egress route.
C.
Sampling / Monitoring 1.
Hydrogen Monitoring Station Dose rate at the Hydrogen Monitoring Station will be about 50 r/hr at 3 days from te open doorways of the Charging Pump Cubicles when handling recirculated safety injection fluids. These cubicles have staggered shoulders for additional shielding, which will be provided.
2.
Liouid Monitoring Station Present layout of the liquid sampling station would result in very high doses to an operator following a DBA. A new sampling station is being designed to meet the requirements of NUREG-0578, Item 2.1.8.b.
IV.
RADIOLOGICAL ANALYSES A.
Source Terms Source terms are generally in accordance with NUREG-0578, as follows:
Liquids - 100% noble gases, 50% halogens and 1% solids dissolved in the reactor coolant and containtent recirculated water.
Gases - 100% noble gases and 50% halogens (either mixed in the con-tainnent atmosphere or plated on the contaimnent wall).
B.
Primary Auxiliary Building Those spaces requiring access af ter a TMI or LOCA type accident have gamma dose rates as indicated.
Sources - The systems involved are:
Ch. & Vol. Cont.
- 10" pipes used for Safety Injection Cool. P. Seal Lkg. - 3" pipes Letdown
- 3" pipes
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. Discussion - The lower level, basically an extension of the Containment-PAB pipe tunnel,will be uninhabitable.
(Doses will be in excess of 500 r/hr.) The two upper levels (21' & 36'), after enclosure of the three charging pump cubicles with 24" overlapping concrete walls, will be habitable outside of cubicles for short term access.
(It is assumed that rad waste functions will be bypassed.)
Specific Dose Rates:
1.
Hydrocen Control and New Sampling Stations Location: 21' level opposite charging pump doorways - on wall shielding valve access area Dose rate with no additional shielding (T=1 hr) = 600-700 r/hr Dose rate with 24" concrete shielding (T=1 hr)) = 2 r/hr (T=24 hr = 0.05 r/hr 2.
Primary Auxiliary Control Board Location: 21' level near Primary Drain Tank Dose rate (T=1 hr) = 0.6 r/hr Dose rate (T=24 hr) = 0.015 r/hr 3.
Emergency Shutdown Panel Location:
11' level near entrance ramp behind the wall shielding the Aerated Drain Tanks Dose rate (T=1 hr) = less than 0.1 r/hr C.
Safeguards Valve Operatino Room Location: 30' level of Containment Spray Pump Area Dose rates:
(T=24 hrs) = 16 r/hr (T=1 wk)
= 3 r/hr Note: This radiation level is caused by the residual neat exchanger 10" outlet piping.
D.
Auxiliary Ste: Generator Feed Pump Room Sources:
Radiation directly through containment concrete Location: 21' level of Auxiliary Steam Generator Feed Pump Room Dose rates:
(T=1 hr) = 0.4 r/hr (T=24 hrs) = <0.1 r/hr E.
Motor Control Centers Sources: Radiation directly through containment concrete Location: 21' & 46' levels of Motor Control Center Room
Dose rates:
(T=1 hr) = 0.4 r/hr (T=24 hrs) = <0.1 r/hr F.
Technical Support Center The Technical Support Center is located in the 21' level of the Service Building and includes the computer room and the Reactor Eng".er's office.
The post LOCA whole body dose in the TSC has been evaluatta and found to be within the limits set forth in Criterion 19 to 10CFR50. Conservative assumptions were made to evaluate the concrete shielding thicknesses in the TSC because of the non-uniform configuration. The following table shows the contributions to whole body dose in the TSC.
TABLE I MY TECH SUPPORT CENTER 30 DAY INTEGRATED WHOLE BODY DOSES FOLLOWING A DBA LOCA SOURCE SHIELDING ASSUMED DOSE - R A.
Overhead Cloud 4" concrete - through roof and 1.4 R north wall of room over TSC B.
Containment 8" concrete - Two 6" composite 0.041 R steel floors (4" equivalent concrete each) 8.1 Through 4'6" concrete site walls B.2 Through 2'6" Same as B.1 1.5 R concrete dome B.3 Skyshine Same as B.1 0.044 R through 2'6" concrete dome C.
Indoor Cloud 0.35 R TOTAL WHOLE BODY DOSE 3.3 R The total whole body gamma dose in the TSC with occupancy as per SRP 6.4 is 3.3 Rem which is within the limits of Criterion 19 to 10CFR50.
G.
Control Room i
The control room whole body gamma dose was computed as part of the stretch power analysis. This calculation determined a 30-day whole body gamma dose of 1.0 Rem. A review of this calculation has shown that the assumptions used were conservative.
In calculating the contribution from the overhead cloud a shielding factor of 100 was conservatively assumed {gr 2' of concrete shieldirl (roof of C.R. - value of 100 based on Co
).
This factor was checktJ using a model to determine the actual energy spectrum as a function of time from the overhead cloud. The shielding factor computed from this analysis was 800. With more realistic assumptions
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- the actual dose would be considerab13 less than the 1.0 Rem and in any case well below the 5 Rem in 30 day limit from Criterion 19 to 10CFR50.
H.
Health Physics Area The Health Physics area is located in the 21' level of the Service Building and includes the Health Physics offices, the chemistry laboratory and the counting room. Table II shows the max. predicted dose rates as well as integrated dose for the Health Physics area. For conservatism the shielding assumed was 4" concrete which would correspond to the areas which only have one 6" composite layer of concrete above them.
In evaluating the 2
component from the containment the direct dose through the 4'6" walls and the skyshine from the dome were neglected as these components have been shown to be negligible compared to the direct dose fr-- the spherical dome.
TABLE II MAX. HEALTH PHYSICS AREA DOSE RATES 100% occupancy for cumulative dose 4" shield for overhead cloud and containment sphere (factor of 4.3 for overhead cloud, factor of 8 for containnent sphere)
TIME AFTER DOSE RATE DOSE CUM.
LOCA SOURCE R/HR.
REM 1 hr.
Indoor Cloud 0.13 0.026 Overhead Cloud 0.28 0.52 Containment Schere 0.73_
l.1 Total 1.1 1.6 8 hrs.
Indoor Cloud 0.027 0.12 Overhead Cloud 0.048 1.1 Containment Sphere 0.14 3.5 Total 0.22 4.7 24 Hrs.
Indoor Cloud 0.0078 0.16 Overhead Cloud 0.0073 1.3 Containment Sphere 0.01 4 4.3 Total 0.029 F.T 96 hrs.
Indoor Cloud 0.00086 0.19 Overhead Cloud 0.00106 1.4 Containment Schere 0.00021 4.5 Total E0021 El-
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720 hrs.
Indoor Cloud 0.000017 0.21 Overhead Cloud 0.000020 1.5 4
I Containment Sphere 0.0000049 4.5 Total Negligible 6.2 I
8 I.
. Diesel Rooms Table III shows the dose rates in the Diesel Rooms as a function of time after the LOCA. No shielding was assumed for the overhead cloud component or the dose from the containment dome. As in the case of the health physics area the direct dose through the 4'6" containment side wall as well as the skyshine from the containment dome were both neglected in this analysis.
Cumulative doses are not presented in Table III as long term occupancy is not anticipated for the Diesel Room area.
TABLE III '
WHOLE BODY GAMMA DOSE RATES IN THE DIESEL ROOM FOLLOWING A LOCA AT MY TIME DOSE RATE R/HR.
I hr Overhead Cloud 1.2 Indoor Cloud 0.13 Containment Sphere
- 1.9 Total 3.2 8 hrs Overhead Cloud 0.21 Indoor Cloud 0.031 Containnent Sphere 0.38 Total 0.62 24 hrs Overhead Cloud 0.031 Indoor Cloud 0.0078 Containnent Sphere 0.044 Total 0.083 4
96 hrs Overhead Cloud 0.0046 Indoor Cloud 0.00086 Containment Sphere 0.00083 Total 0.0063 l
720 hrt Overhead Cloud 0.000088 Indoor Cloud 0.000017 Containment Schere 0.000027 i
Total 0.00013
- For conservatism used 75m to containment center J.
Primary Emergency Coordination Center l
The Primary ECC is located in the MY Visitor's Center approximately 200 meters from the containment centerline. The primary ECC was evaluated for whole 2
body gamma dose as a function of time following a LOCA at MY. Table IV shows dose rates and cumulative doses for this area. Note that the cload dose was calculated assuming infinite mixing with the outside plume. A'so note that cumulative doses are based on 100% occupancy.
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TABLE IV DOSE RATES AND INTEGRATED DOSES AT THE PRIMARY ECC FOLLOWING A LOCA AT MY l
TIME SOURCE DOSE RATE R/HR DOSE R 1
1 Hr Cloud dose
- 0.54 1.0 Containment spheret 0.36 0.57 Total 0.90 llli-8 Hrs Cloud dose 0.049 1.8 Containment schere 0.071 1.8 Total 0.12 3!T 4
t 24 Hrs Cloud dtse 0.0098 2.0 Containment schere 0.0077 2.2 Total 0.018 4.2 96 Hrs Cloud dose 0.00092 2.1 Containment sphere 0.00012 2.3 Total 0.0010 4.4 720 Hrs Cloud dose 0.000025 2.2 Containment sphere Neoligible 2.3 Total 0.000025 4.5
- Assumed infinite mixing with outdoor cloud t Used 200 meters from center of containment K.
Secondary Emergency Coordination Center The Secondary ECC is located in a house approximately 600 meters from the containment centerline. The Secondary ECC was evaluated for whole body gamma dose as a function of time following a LOCA at MY.
Table V shows dose rates and cumluative doses for this area.
Note that the cloud dose was calculated assuming infinite mixing with the outside plume. Also note that cumulative doses are based on 100% occupancy.
TABLE V WHOLE BODY GAMMA DOSES AT SECONDARY EMERGENCY C0 ORDINATION CENTER (600m FROM CONTAINMENT) FOLLOWING A LOCA AT MY TIME SOURCE DOSE RATE mR/HR DOSE mR 1 Hr Cloud 64.
120.
Containment sphere 6.8 11.
Total 71.
131.
8 Hrs Cloud 5.6 210.
Containment sphere 1.2 33.
Total 6.8 243.
24 Hrs Cloud 0.26 220.
Containment sphere 0.096 39.
Total 0.36 259.
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TABLE V'(Cont'd)
TIME SOURCE DOSE RATE mR/HR DOSE mR 96 hrs Cloud'-
Negligible 220.
- Containment sphere Negligible 40.
260.
Total 720 hrs Cloud Negligible 220.
Containment sphere Negligible 40.
2 260.
Total 1
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