ML20236N419
| ML20236N419 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 11/09/1987 |
| From: | GENERAL PUBLIC UTILITIES CORP. |
| To: | |
| Shared Package | |
| ML20236N402 | List: |
| References | |
| NUDOCS 8711160155 | |
| Download: ML20236N419 (6) | |
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1 3.9.12 FUEL' HANDLING BUILDING / AUXILIARY BUILDING AIR CLEANUP SYSTEMS:
i FUEL HANDLING' BUILDING AIR ~ CLEANUP SYSTEM "
' LIMITING CONDITIONS FOR OPERATION
'3.9.12.1 The' Fuel Handling Building Air Cleanup Exhaust System shall be
- 0PERABLE with one'of the-four system air. cleanup exhaust fans OPERABLE.
APPLICABILITY: MODES 1, 2 and 3 l
. ACTION:1 h
'With the Fuel' Handling Building Air Cleanup Exhaust System inoperable, restore the system to OPERABLE-status within.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or. suspend all operations
-involving movement of liquid and solid radioactive wastes'in the Fuel Handling Building (other-than sampling evaluations required by the' Technical l
Specifications or the RECOVERY OPERATIONS PLAN), the release of which could l
exceed 50% of the Appendix B Technical Specification instantaneous release rate for gaseous effluents, until the system is restored to OPERABLE status.
AUXILIARY BUILDING AIR CLEANUP EXHAUST SYSTEM l
- 3'. 9.12. 2 The. Auxiliary Building Air Cleanup Exhaust System shall be OPERABLE with one of the_four system air cleanup exhaust fans OPERABLE.
APPLICABILITY: MODES 1, 2 and 3
. ACTION:
With the Auxiliary Building Air Cleanup Exhaust System inoperable, restore the system to operable status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or suspend all operations involving movement of liquid and solid radioactive wastes in the Auxiliary Building (other than sampling evolutions. required by the Technical' Specifications or
-RECOVERY OPERATIONS PLAN), the release of which could exceed 50% of the Appendix B Technical Specification instantaneous release rate for gaseous effluents,'until tne system is restored to OPERABLE status.
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8'/11160155 871109 hDR ADOCK 05000320 PDR THREE MILE ISLAND - UNIT 2 3.9-5 I
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'3/4.9 RADI0 ACTIVE' MATERIALS STCRAGE i
BASES l
3/4.9.1 SPENT FUEL STORAGE POOL "A" WATER LEVEL MONITORING Spent Fuel Storage Pool "A" Water Level Monitoring instrumentation has been I
provided to assure the capability to monitor water level in the Spent Fuel Storage Pool "A".
3/4.9.2 SPENT FUEL STORAGE POOL "A" WATER LEVEL The water level.in the Spent Fuel Storage Pool "A" has been established to limit the dose rate, due to the storage of Canisters, to acceptable levels.
3/4.9.3 FUEL TRANSFER CANAL (DEEP END) WATER LEVEL MONITORING Fuel Transfer Canal Water Level Monitoring instrumentation has been provided to assure the capability to monitor water level in the deep end of the Fuel Transfer Canal.
3_/4 a.4 FUEL TRANSFER CANAL (DEEP END) WATER LEVEL The water level in the Fuel Transfer Canal (deep end) has been established to limit the dose rate, due to the storage of the plenum assembly and Canisters, to acceptable levels.
3/4.9.12 FUEL HANDLING BUILDING / AUXILIARY BUILDING AIR CLEANUP SYSTEM The requirement-for the Fuel Handling Building / Auxiliary Building Air Cleanup System to be operating or OPERABLE ensures that radioactive material released to these buildings will be filtered through the HEPA filters prior to release to the atmosphere.
In the event the systems are not restored to OPERABLE status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />,.the Technical Specifications require the suspension of any liquid or solid radioactive waste handling, the release of which could exceed 50% of the instantaneous release rate limits for gaseous effluents specified in Section 2.1.2 of the Appendix B Technical Specifications. These restrictions correspond to solid radioactive waste with a total activity of greater than 1 curie of particulate with half-lives greater than eight (8) days and liquid radioactive waste with an activity greater than 0.5 curies of particulate with half-lives greater than eight (8) days.
3/4.9.13 ACCIDENT GENERATED WATER These specifications are provided to ensure compliance with the Commission's Statement of May 25, 1979, and the Commission's Statement of Policy and Notice of Intent to Prepare a Programmatic Environmental Impact Statement of November
~21, 1979, which prohibit these actions pending evaluation of the environmental impacts of such actions. The PEIS issued in March 1981, deferred a decision on the ultimate disposal of processed water. Further Commission action is necessary prior to release of ACCIDENT GENERATED WATER.
However, the Commission has recognized that there may be emergency situations, not at this time foreseen, which could require rapid action.
In these situations, the Commission has indicated its intention to consult with the Council on Environmental Quality to the extent practicable.
THREE MILE ISLAND - UNIT 2 B 3/4.9-1
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ATTACHMENT P
-4410-87-L-0167 NRC'COMENT'6 Proposed specification 3.9.12,thould be updated to reflect current plant Lconditions and operations. The reference to, gases can be~omitted.
l' Radioactive liquids and solids should be addressed in a quantitative manner.
L GPU NUCLEAR RESPONSE'
'Per the-above NRC. Comment, attached is a proposed revision to the Action Statement and Basis for Specification 3.9.12.1, " Fuel Handling Building Air Cleanup System',"1and 3.9.12.2, " Auxiliary Bui'lding Air. Cleanup Exhaust System." This proposed revision deletes the references to gaseous radioactive (wastes and proposes'a quantitative. criteria for liquid and solid radioactive waste's based on the instantaneous release rate for gaseous effluents in l
Section 2.1.2 of the Appendix B Technical Specifications.
Specifically, the action statements of Specification 3.9.12.1 and 3.9.12.2
-have been. revised to require, if either the' Auxiliary Building (Aux. Bldg.') or Fuel Handling Building (FHB) Air Cleanup System is'not restored to operable
'i status within four-(4) hours, the suspension of operations involving movements of liquid and solid radioactive wastes in the respective building the release of which could exceed 50% of the Appendix 8 Technical Specification instantaneous release rate for gaseous effluents (i.e., 1.5 E-1 UC1/sec for particulate with half-lives greater than eight (8) days).
The current exception for sampling evolutions required by the. Technical Specifications or Recovery. Operations Plan would still apply.
The basis for this proposal is an engineering analysis that was performed to l
determine the: radiological effects of a release of one (1) curie (C1) of solid radioactive' waste'and an arbitrarily assumed 200 gallon spill of Reactor l
Coolant System (RCS) grade water and Reactor Building (RB) basement water.
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The' details of-this analysis are described below and are summarized in the f
attached table. The results of this analysis indicate that, for example, the following activities could continue:
o Transfers of defueling waste containing up to 1 Ci of activity of particulate with half-lives greater than eight (8) days Movement of RCS grade water containing a total activity of up to 0.5 Ci j
o of particulate with half-lives greater than eight (8) days However, based on the analysis, movement of R8 basement water would be
. prohibited in the Aux. Bldg. or FH3 if the Air Cleanup Exhaust System in that Ebuilding is not operable and it is not. restored to operable status within four (4) hours.
The below sections describe the detailed analysis for each of the three (3) assumed scenarios (i.e., movement of 1 Ci of defueling waste trash, a 200 gallon spill of'RCS grade water, and a 200 gallon spill of RB basement water). The release fractions used were consistent with similar scenarios in the GPU Nuclear Corporation Seismic Design Criteria (reference GPU-Nuclear letter.4410-85-L-0077, dated April 16, 1985). The release rate for a one (1) hour release was calculated for each scenario.
ATTAC + TENT 4410-87-L-0167 Scenario I - Defueling Waste Trash This scenario assumes that one (1) Ci of defueling waste trash is released airborne at a release fraction of 10-4 The resultant release rate over a one (1) hour period is:
(1.0 C1) (1.0 E-4) (1.0 E-6 uC1) 2.78 E-2 UCi
=
(
Ci
)
sec 3600 sec Thus, since the release rate is less than the plaposed Tech. Spec, limitation of 1.5 E-1 uC1/sec (i.e., 50% of the Appendix B Tech. Specs. instantaneous release rate for particulate with greater than eight (8) days), this activity l
is permissible when either the Aux. Oldg. or FHB Ventilation System is out-of-service for a period greater than four (4) hours.
l Scenario 2 - 200 Gallon Spill of RCS Grade Water This scenario assumes a hypothetical 200 gallon spill of RCS grade water with approximately 5.0 E-1 Ci total inventory of particulate with half-lives greater than eight (8) days.
This inventory is based on the radionuclides inventory of RCS sample 87-23112 on October 12, 1987.
The total inventory of this sample, which includes Cs-134, Cs-137, and Sr-90 is 6.39 E-1 uCi/ml which for a 200 gallon spill converts to:
(6.39 E-1 uCi/ml)(
1 mi
)(200 gallons)( 1 Ci )
approx. 0.5 C1
=
(2.64 E-4 gallons)
(TO M )
The resultant release rate based on a release fraction of 10-3 is:
(0.5 C1) (1.0 E-3) (1.0 E6 uCi/cc) 1.4 E-1 UCi/sec
=
3600 sec The above release rate is less than the proposed Tech. Spec. limitation of 1.5 E-1 uCi/sec, thus, this activity is permissible when either the Aux. Bldg.
or FHB Ventilation System is out-of-service for a period greater than four (4) hours.
In the specific case of the Defueling Water Cleanup System (DWCS), the NRC approved DWCS Technical Evaluation Report (TER), reference GPU Nuclear letter 4410-87-L-0051, dated April 19, 1987, notes that a line break in the Reactor vessel Cleanup : System portion of the DWCS could result in a spill of up to 1000 gallons. Therefore, since the limiting factor in determining the release rate is the total Ci inventory, RCS processing is permissible provided the total curie inventory of 1000 gallons of RCS water does not exceed 0.5 Ci.
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ATTACH 404T-4410-87-L-0167 Scenario'3 - 200' Gallon' Spill of RB Basemens Water This' scenario _ assumes'a 200 gallon' spill of RB basement water with approximately.l.63 Ci~ total' inventory. This inventory-is based on the i
L Jradionuclide inventory of a sample 87-22887 of WDL-T-88 on.0ctober 4, 1987.
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lThe. total' inventory of;this sample which includes H-3,' Sr-90,.Cs-134, and Cs-137 is 2.15.uC1/mllwhich, for.a 200 gallon spill, converts to:-
l-1;63 Ci (2.15 uC1/ml) (
1 ml.
)(200 gallons)(
l_' Ci- )
=
L
- (2.64.E-4 gallons)
(100 uC1)
The-resultant release rate based on a release fraction of 10-3 is:
t
'(1.63 Ci) (1.0 E-3)
(1.0 E6 uCi/cc).= 4.53 E-1 uC1/sec 3600 sec
.The above release rate exceeds the proposed Tech. Spec. limitations of i
.1.5'E-1 uCi/sec, thus, the movement of this water would be prohibited in'the~
1 Aux / Bldg. or FHB whenever.the' associated Ventilation System is out-of-service
'for-a period greaterLthan four.(4) hours.. (NOTE: H-3: activity for this sample was 8.3 E-2 uCi/ml.
Subtracting this activity from the total activity would give a release ~ rate of 4.36 E-1 uCi/sec for particulate with half-lives greater than eight (8) days.)-
In ' conclusion, GPU Nuclear believes.that the proposed limitation for liquid and solid radioactive wastes in. Specifications 3.9.12.1 and 3.9.12.2 are conservative and will ensure that any potential release will be we11'within
' Technical' Specifications and regulatory limits.
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ATTACHMENT l.
4410-87-L-0167 1
l TABLE 1 Summary of Release Rates for Accident Scenarios Described in TSCR 53 - Revision 2 Release Rate Scenario Conditions Release Fraction (uC1/sec)
Defueling 1 Curie, Note (1) 10-4 Note (2) 2.78 E-2 i
Waste No HVAC, Airborne No HEPA (Trash)
Filtration j
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.200 Gallons, Note (3) 10-3 Note (2) 1.4 E-1 Airborne i
' Filtration RB Basement 200 Gallons, Note (4) 10-3 Note (2) 4.53 E-1 l
Water Spill Airborne No HVAC, No HEPA Filtration l
(1) From Defueling Waste Isotopic Update, reference Radiological Engineering memorandum 9240-87-4207, dated October 19, 1987.
(2) From SER 4430-87-7322-851, "GPU Nuclear Corporation Seismic Design Criteria," GPU Nuclear letter 4410-85-L-0077 dated April 16, 1985, (3) From RCS sample 87-23112 (4) From Sample 87-22887