ML20215G105
| ML20215G105 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 10/01/1986 |
| From: | Farrar D COMMONWEALTH EDISON CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| 2172K, NUDOCS 8610170106 | |
| Download: ML20215G105 (19) | |
Text
- _ ~
m) Commonwealth Edison
.1 - 72 West Adams Street, Chicago, Ilhnois V
Kidress Reply to: Post Office Box 767 Chicago, Ilknots 60690 - 0767 October 1, 1986 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555
Subject:
Byrcn Station Unit 2 Deferral of Limited Aspects of the Initial Test Program NRC Docket No. 50-455
Dear Mr. Denton:
This is a request for schedular relief for a few items contained in the Byron Station Initial Test Program. This program is described in Chapter 14 of the Byron /Braidwood FSAR. In accordance with our commitment in FSAR Section 14.2.5, we intended to complete all preoperational testing and evaluation of test results prior to fuel load. We believe over 97% of the preoperational testing and results approval will be completed at the time Byron Unit 2 is ready for fuel load. However, it has become apparent that-certain limited aspects of the test program will not be fully completed at that time. We believe these items can be temporarily deferred beyond fuel load with no impact on safe plant operations.
4 Thirteen systems are involved in the test program deferrals. The testing and results approval for twelve of these systems will'be completed i
by the operational. mode in which the Technical Specifications require the system to be operable. These systems, the status of testing, and the nature of the deferrals are described in Attachment A of this letter.
1 The thirteenth system, auxiliary building ventilation, is presently being tested and balanced. However, the testing and results approval for this system will not be completed by the time Unit 2 would enter.the operational mode in which the system is required to be operable by Technical Specifications. The status of testing, the present functional capability of the system, and the technical justification for temporarily operating the plant without the testing completed are contained in Attachment B of this letter. Attachment B also contains a temporary technical specification to address interim operation of the auxiliary building ventilation system.
4
\\
f[
y" I
~8610170106 861001 I>
PDR ADOCK 05000455 -
-.a,.
_A.
l j
_ _,.,. _ _ _ _,. _. _ _ _ _ _, _ _ _, _ _ _ _. _. ~. _ _ _. _. _ _ _ _ _.,. - _ _ __. _, - - _ _ _, _ _
Mr. H. R. Denton October 1, 1986 As noted earlier, commonwealth Edison believes the deferrals 2
outlined above will have no impact on safe plant operation. This is because the items in Attachment A will be completed by the operational mode in which the Technical Specifications require the systems to be operable. Inherent in the Technical Specifications system operability requirements is an accepted determination of safety importance of systems in relation to the operational mode of the plant.
In addition, all of the items listed in Attachment A will be completed prior to initial criticality (Mode 2).
In modes 6 through 3, a shutdown boron concentration sufficient to assure shutdown under all conditions will be maintained in the reactor coolant system. Because the reactor will never become critical during either the fuel loading process or the precritical testing sequence, it follows that no fission product source terms will be generated. Therefore, the Attachment A deferrals will have no safety impact.
Attachment B addresses the safety aspects of plant operation with the auxiliary building ventilation system not fully operable. The analysis in Attachment B shows that the plant can operate up to essentially 30% power without exceeding offsite or control room radiological dose limits. Since these limits would not be exceeded, this proposed deferral will have no safety impact.
Commonwealth Edison believes that deferral of the items in Attachments A and B should be provided as schedular relief under 10 CPR 50.57(b) because facility construction is substantially completed.
The proposed deferrals have resulted from a variety of reasons, including delays in the delivery and subsequent installation of equipment, and design change improvements suggested by operating experience at Byron Unit 1.
Nonetheless, Commonwealth Edison has been continually working toward completing the preoperational test program. At the time the plant is ready for fuel load, the test program will be approximately 97% complete. Under these circumstances, the failure to complete a few items late in the licensing process in spite of a good faith effort to do so constitutes special circumstances justifying the relief sought.
In addition, failure to grant this relief would result in undue hardships and costs far in excess of those contemplated when Commonwealth Edison committed to the preoperational test schedule. Although Commonwealth Edison recognized that elements of the preoperational testing program could be delayed beyond fuel load, nonetheless, Commonwealth Edison committed to completing its program prior to fuel load in reliance on the long standing l
l
)
Mr. H. R. Denton October 1, 1986 NRC practice of granting schedular relief for deferrals which do not present safety issues. Under these circumstances, the undue hardships which woulo result from a rigid adherence to the program's schedule would constitute special circumstances further justifying the relief sought.
Finally, deferral of the items in Attachments A and B are authorized by law, do not violate a statute or regulation of another agency, and do not affect common defense and security.
For the reasons stated above, Commonwealth Edison believes schedular relief should be granted for the items on Attachments A and B.
One signed original and fifteen copies of this letter and attachments are provided for NRC review.
please address any questions regarding this matter to this office.
Very truly yours,
_-_=-_q Dennis Farrar Director of Nuclear Licensing im Attachments cc: Byron Resident Inspector 2172K
ATTAGGENT A SYSTEMS WICH WILL NOT BE OPERABLE BY FUEL LOAD (MODE 6),
BUT WILL BE OPERABLE WEN REQUIRED BY TECHNICAL SPECIFICATIONS SYSTEMS WITH TEST RESULTS NOT YET APPROVED TECH SPEC COMPLETION NATURE TEST / DESCRIPTION APPLICABILITY REQUIRED OF DEFFERAL STATUS AS OF 09/30/96 EF 26.60/ Engineered Safety Features-Mode 4 Prior to Mode 4 See Note B (1)
Testing couplete 08/15 Safeguards EF 26.62/ Engineered Safety Features-Mode 4 Prior to Mode 4 See Note B (2)
Testing complete 08/07 Logic and Time Response FW 34.60/ Main Feedwater Mode 4 Prior to Mode 4 See Note B (3)
Testing complete 09/04 SX 76.60/ Essential Service Water Mode 4 In Mode 4 See Note B (4)
Retest pending Available to DG,CC During Modes 6 & 5 VQ 94.60/ Primary Containment Purge Mode 4 Prior to Mode 4 See Note B (5)
Retest pending l
CV 18.67/ Chemical & Volume Control-Mode 3 Prior to Mode 3 See Note B (6)
Retest to complete 10/09
{
IHF RP 68.60/ Reactor Protection - Time Mode 2 Prior to Mode 2 See Note B (7)
Retest complete 09/14
Response
VD
.61/VD, VE, VX Ventilation-Non-Mode Prior to Mode 2 See Note B (8)
Retest pending Integrated
ATTACWENT A - Cont'd SYSTEMS WITH TEST RESULTS APPROVED - MINOR DEFICIENCIES REMAIN DEFICIENCY TECH SPEC COMPLETION NATURE TEST DESCRIPTION APPLICA8ILITY REQUIRED OF DEFFERAL STATUS AS OF 09/30/86 AB 1.60 2AB03P pump perf. curve, and Mode 3 Prior to Mode 3 See Note B (9)
Retest pending Boric Acid 2AB04F filter inlet pressure Processing RY 69.67 PZR level and pressure Mode 3 In Mode 3 See Note B (10)
Instruments have been Reactor Coolant indication & controls recalibrated. Perfor1n Pressurizer - IHF reverification retest during Mode 3 i
SI 73.63 Leakage of one RCS pressure Mode 2 Prior to Mode 2 See Note B (11)
Valve seating surface i
ECCS Chk Valve isolation check valve not has been repaired. Per-Oper, & Leakage less than 1 gpn form retest during Mode 3 AF 3.60 Retest of dual pung start Mode 2 Prior to Mode 2 See Note B (12)
Perforin retest during Mode 3 Auxiliary Feedwater for flow to S.G.'s i
i PS 61.60 CASP, Aux Bldg. equip drn Non-Mode Mode 2 See Note B (13)
Retesting to canplete 10/09 Primary Process tank, BTRS demin., letdn.
t Sampling ht. exch.
a a
1
ATTAQWENT A - NOTES 8 (1) EF 26.60/ Engineered Safety Features - Safeguards. Preoperational testing of the system has been conpleted and the test results are under review. The test objective is to verify proper operation of the safeguards actuation circuits. Some potential exists for minor retesting. The Engineered Safety Features Actuation System is not required prior to Mode 4.
B (2) EF 26.62/ Engineered Safety Features - Logic and Time Response. Preoperational testing of the system has been conpleted and the test results are under review. The test objectives are to verify operation of logic output and response time from is.put con 6inations in conjunction with each possible interlock logic state. Evaluation and approval of this test is dependent on EF 26.60 results which may require retesting as noted above. The Engineered Safety Features Actuation System is not required prior to Mode 4.
i B (3) FW 34.60/ Main Feedwater. Preoperational testing of the system has been completed and the test results are under review. The test objectives are to demonstrate proper response time and actuation logic of isolation and control valves in the system. The main feedwater system and feedwater isolation function are not required prior to Mode 4.
1 B (4) SX 76.60/ Essential Service Water. (a) Preoperational testing and results review and evaluation have been completed. The results evaluation detennined that f.aw data to various heat exchangers was less than the values expected. Subsequently, the heat exchangers were disassen61ed, inspected and any siltation removed. A retest of the flows was conpletei Septenber 18, 1986. The retest results have received a preliminary engineering review which indicates design modifications to incorporate flow limiting orifice plates at selected locations may be necessary to rebalance flows. The flows existing without these flow balancing alterations are under the expected values. Any modifications and retesting resulting from the completed evaluation will be i
conpleted prior to Mode 4.
(b) Due to insufficent data obtained during the Byron Unit 2 integrated hot functional testing, additional data is required. Proper heat load conditions on cooling tower (s) are required to obtain this data. These conditions, which involve heatup to and cooldown from operating temperature, are not available until Mode 4.
The essential service water pump discharge tenperatures were demonstrated to be less than the values specified in Technical Specification 3.7.5.d.
Therefore, no j
relief from Technical Specification 3/4.7.5 (Mode 4 requirements) is required.
B (5) VQ 94.60/ Primary Containment Purge. Preoperational testing was completed September 22, 1986. During the course of testing, containment purge isolation valve 2VQ001A did not respond to open or close signals. The valve hydraulic operator was disassenbled and seals within the operator require replacement. Replacement seals are expected to be delivered by Septenber 26, 1986.
]
Reassembly of the valve operator and retesting will be accomplished before entry into Mode 6 or during Mode 5.
During Mode 6, each purge isolation valve must be closed if isolation (closure) can not occur upon an ESF test signal. (ref: Technical Specification 3/4.9.9)
I
l ATTAQWENT A - NOTES. Cont'd I
8 (6)
CV 18.67/ Chem. & Volume Control - IHF. Preoperational testing and results review and evaluation have been completed. The results evaluation detemined that the 2CV110A valve control loop requires tuning. Valve 2CV110A is a control valve in the boric acid supply line to the boric acid blender. The instrument loop tuning will be co gleted before Septad er 30, 1986. A retest of the j
valve flow control will be perfomed by borating the volume control tank. This retest will be coglete by approximately October 9, 1986. The boration flow path from the boric acid storage system is not required prior to Mode 3 if an alternative flow path from the refueling water storage tank is operable. Therefore, no relief from Technical Specifications 3.1.2.1, 3.1.2.5 and i
3.1.2.6 is required.
B (7)
RP 68.60/ Reactor Protection - Time Response. Preoperational testing has been co g leted and the test results are under review.
l The test objectives are to verify the time response from input to reactor trip breaker operation. Some potential exists for minor retesting. Per Technical Specification 3.3.1 Tables 3.3-1 and 3.3-2, the inputs included in the scope of Test RP 68.60 are not required prior to Mode 2.
8 (8)
VD 86.61/VD, VE, a Ventilation - Integrated. Preoperational testing and results review and evaluation have been cogleted. The i
results evaluation detemined that air flows in certain ductwork extensions were less than the values expected. Subsequently, the fan blade pitch on the diesel generator 28 fan was changed to increase the developed airflow. A retest of the cod ined VD, VE, VX airflows was completed Septed er 22, 1986. The retest results have received a preliminary engineering review which indicates an i
additional iteration involving a fan blade pitch change may be necessary to increase air flows. The flows existing without this l
change are under the expected values. Any modifications and retesting resulting from the cogleted evaluation will be cogleted
]
prior to Mode 2.
1 B (9)
AB 1.60, 2A803P pump performance curve, 2A804F filter inlet pressure. verification. Preoperational testing and results approval have been cogleted. Subsequent to preoperational testing, maintenance work on pump seals was perfomed to correct a minor test l
deficiency regarding seal leakage. During post maintenance testing, the pressure and flow from the pump did not meet the
)
manufacturer's pug curve. The pump did, however, develop the required pressure and flow to meet the system operating and design requirements.
B (10) RY 69.67, Pressurizer level and pressure indication and controls reverification. Preoperational testing and results review and evaluation have been completed. The post test instrument accuracy checks found that at the high range of two instrument units, the cardinal points had drifted and therefore, were out of calibration. This invalidated the recorded test data. The instrisnent i
recalibration has been completed, but Mode 3 pressurizer conditions are required to retake data. The instrument inaccuracies were i
at points outside the scope of Technical Specification 3/4.3.2 Table 3.3-4 items 1.d and 9.a.
Therefore, no relief from Technical i
Specification 3/4.3.2 Table 3.3-3 items 1.d and 9.a (Mode 3 requiments) is required. Retesting will be perfomed during Mode 3.
4 1
ATTADOENT A - NOTES, Cont'd 4
l B (11) SI 73.63, Vaive 25189480 leakage not less than or equal to I gpm. Valve 25189480 is a reactor coolant system (RCS) pressure isolation check valve. The valve was tested for leakage at RCS operating pressure and temperature during the Unit 2 integrated hot functional test. Leakage was greater than 1 gpm. The valve seating surface has been repaired subsequent to hot functional testing. Mode 3 plant conditions are required to establish the necessary parameters for testing. Technical Specification 4.4.6.2.2 requires all RCS pressure isolation valves to be tested for leakage prior to entering Mode 2.
The retest will be completed in Mode 3.
B (12) AF 3.60, Auxiliary Feedwater, retest of dual pump start. As a result of a preoperational test deficiency, a modification has been installed on the AFW pung suction lines from the condensate storage tank. The final test of this modification will consist of a dual pump start during heatup prior to initial criticality. The steam generators will be at no load pressure and temperature for j
this test. Testing at cold conditions, simulating steam generator back pressure, has been successfully completed. The retesting will be perfonned during Mode 3 conditions. The auxiliary feedwater pump (s) discharge pressure and flow were demonstrated to be i
greater than the values specified in Technical Specification 4.7.1.2.1.
Therefore, no relief from Technical Specification 3/4.7.1.2 (Mode 3 requirements) is required. The retest will be perfonned in combination with Technical Specification surveillance requirement 4.7.1.2.2 which is required prior to entering Mode 2.
B (13) PS 61.60, CASP, Aux. Bldg. equip. drn. tank, BTRS demin., letdown ht. exch. Preoperational testing and results approval have been i
completed. Various modifications, some of which are being made to correct minor test deficiencies, are being installed on the I
system to increase sample flowrate and incorporate changes from Unit 1 operating experience. A retest of these changes will be performed upon completion of installation.
5 i
I 1
2172K i
t i
I
ATTACHMENT B i
One system, auxiliary building ventilation (VA), will not be fully operable for Unit 2 when required by Technical Specifications. The VA system is common to Units 1 and 2 serving the auxiliary building and fuel handling building. The system is designed to filter radioactive contaminants from the air exhausted from the buildings to assure offsite radiation levels are within acceptable limits. The system provides balanced supply and exhaust airflows throughout the buildings to maintain them at negative pressures with respect to atmosphere. The system also provides cooling to the Units 1 and 2 angineered safety features (ESP) equipment cubicles as well as general areas of the buildings that are shared by both units. Proper temperature is maintained in each ESF cubicle by a cubicle cooler which operates independently of the VA system's main fans. The cubicle cooler subsystem is not a subject of this schedular relief request.
A complete description of the VA system is discussed in FSAR Section 9.4.5.1.
Portions of the VA system serving Unit 1 and common areas of the auxiliary building and fuel handling building have been operable since July 1, 1985. The remaining portions of the VA system, serving Unit 2 areas of the auxiliary building, are the subject of this schedular relief request. These portions must become operable to support the operation of Byron Unit 2.
Ventilation requirements for the Unit 2 areas of the auxiliary building will vary with each stage of the Unit 2 startup.
Radiation levels and temperatures inside these areas will be dependent upon the plant operating mode and reactor power level. Filtration and cooling needs of the Unit 2 areas will increase to maintain existing margins of safety as the unit approaches full power operation. At the time of Unit 2 fuel load, only the ESF equipment cubicle coolers need to be operable.
Later, at high power levels, HEPA and charcoal filtration of Unit 2 area exhaust air is required and room negative pressures must be established to ensure plant safety.
At present, the Unit 2 portion of the auxiliary building ventilation system is constructed and supply and exhaust airflows are functioning.
These airflows are maintaining the Unit 2 areas at a negative pressure with respect to atmosphere. The activities remaining are the adjustment of individual balancing dampers to obtain 1/4 inch water gauge negative pressure in each individual subccmpartment, and then performance of the integrated testing. During these iterations, it is possible that individual subcompartment airflows may change to the point that the pressure in the I
subcompartment may become positive in relation to other subcompartments.
It is highly unlikely it will become positive in relation to atmosphere since the building total exhaust airflow will be greater than the total supply airflow.
l
l l,
l Interim Operating Plan An interim operating plan for the Unit 2 portions of the auxiliary building ventilation system has been established. This plan will assure adequate ventilation for the auxiliary building and fuel handling building during Unit 1 operation and Unit 2 startup. During the implementation of this plan, the Unit 1 and common portions of the VA system will meet the requirements of Technical Specifications 3/4.7.7.
The plan follows:
A.
Initial Startup Phase During the period between Unit 2 fuel load and initial criticality, fission products will not be produced from the Unit 2 reactor.
Radiation levels inside Unit 2 areas of the auxiliary building will be at safe levels and there will be essentially no risk of releasing radioactive contaminants to the environment. Consequently, there will be no need for HEPA and charcoal filtration of Unit 2 area exhaust air at this time. Also, these areas do not need to be maintained at negative pressures with respect to atmosphere.
Excessive heat load will normally not exist in the Unit 2 areas of the auxiliary building during this period. However, temperature rises may occur in the ESP equipment cubicles when operating the ESF equipment.
Cooling of the ESF equipment cubicles will be provided by their respective cubicle coolers which will be operable. Operation of the coolers will provide the necessary heat removal in any plant operating condition.
B.
Power Operation Beyond initial criticality, fission products are produced at a rate directly proportional to the reactor's thermal power level. The potential for radioactive releases from the auxiliary building is due mainly to fission products from emergency core cooling system (ECCS) equipment leakage. During normal plant operation at all power levels, ECCS equipment leakage will be minimal since most ECCS equipment will not be operating. Therefore, radiation and airborne radioactivity levels in the auxiliary building will be normal and will not affect offsite and control room doses.
However, during an accident condition, more ECCS equipment may operate and leakage may increase.
If this condition occurs, the auxiliary building ventilation system's HEPA and l
charcoal filters, which are operable, will ensure that offsite and I
control room doses remain within 10 CFR 100 and GDC 19 limits.
i l
l l
l l
_..-_,_-.___._.._,-,,....__.,_-_,_.._-.._m,
.,,,,,,,,..~,,r,
_ The radiological consequences of ECCS component leakage in the auxiliary building are discussed in FSAR Section 15.6.5.6.
This analysis was based on methods described in Regulatory Guide 1.4. and assumes component leakage of approximately 1 gallon per hour. Analysis of offsite and control room doses for Byron Station can be found in FSAR Sections 15.0 and 6.4, respectively. The FSAR dose analysis assumes the auxiliary building ventilation system to be operable. Expected doses are provided in FSAR Tables 15.0.11 and 6.4.1.
Offsite and control room doses have been reanalyzed assuming the auxiliary building ventilation system is inoperable (i.e., without HEPA and charcoal filtration). Details of this reanalysis are provided in Attachment I_ entitled "Offsite and Control Room Doses - Byron Staton".
In the reanalysis, the assumed ECCS equipment. leakage is that stated in Standard Review Plan Section 15.6.5, Appendix B and is larger than the leakage values used in the PSAR analysis. Under these conditions, the reanalysis demonstrates that Byron Unit 2 can operate at a maximum power level of essentially 30% with the auxiliary building ventilation system inoperable without violating offsite and control room dose limits.
I The offsite and control room dose analysis discussed above assumes continuous ECCS leakage of 1 gallon per minute.
If this ECCS leakage is reduced, Byron Unit 2 can safely operate at higher reactor power levels without HEPA and charcoal filtration and negative pressure. The enclosed Figure 1 indicates the allowable Unit 2 power level at any given Unit 2 BCCS leakage rate without filtration of Unit 2 area exhaust air and with acceptable low population zone (LPZ), exclusion area boundary (EAB) and control room doses. This curve is based on the assumptions noted on the figure.
It is clear from Figure 1 that Byron Unit 2 can potentially operate at a reactor power level of 100% with the auxiliary building ventilation system inoperable without exceeding offsite and control room dose limitations.
l Abnormal temperatures are not expected to exist in Unit 2 areas of the auxiliary building prior to the completion of final air balancing. The l
Unit 2 cubicle coolers will maintain the ESP equipment at proper tempera-l tures when operated. The VA system's main supply and exhaust fans will generate airflow through other Unit 2 areas of the building while the balancing is in progress.
Based on the discussion above, balanced airflows and negative l
pressures may not exist in Unit 2 areas of the auxiliary building prior to l
low power operation of Byron Unit 2.
In addition, the preoperational testing of the Unit 2 portions of the VA system may not be completed prior to low power operation.
I
_ - Prior to exceeding 30% reactor power operation of Byron Unit 2 with Unit 2 portions of the VA system inoperable, the leak rate from Unit 2 BCCS equipment in the auxiliary building will be determined. This leak rate will be used to determine, from Figure 1, the available range of~ reactor power operation for that condition. Byron Unit 2 will operate within the available power range until the Unit 2 portions of auxiliary building ventilation system becomes operable. At that time, these operating limits will no longer apply. Because of the expected short duration of these limitations, no ongoing surveillance will be conducted once the leak rate has been determined for a given power level.
Prior to operation of Byron Unit 2 outside the limits determinad from Figure 1, balanced airflows and negative pressures will exist in the Unit 2 areas of the auxiliary building. Also, the preoperational testing to verify proper functioning of Unit 2 portions of the VA system will be completed.
After completion of the above activities, the auxiliary building ventilation system will function to ensure safe operation of Byron Units 1 and 2 at all power levels.
Technical Specification surveillance requirement 4.7.7.d.3 states that a negative pressure of greater than or equal to 1/4 inch water gauge must be maintained in the ECCS equipment rooms at certain system flowrates.
Since adjustment of individual balancing dampers and integrated testing may cause a temporary loss of the 1/4 inch water gauge negative pressure in some ECCS equipment rooms, a temporary technical specification allowing for this situation to exist until July 1, 1987 is being proposed. The proposed technical specification is contained in Attachment III. All other surveillance requirements of Technical Specification 4.7.7 will be met for the Unit 2 portions of the auxiliary building ventilation system prict to entering Mode 4.
Conservatism of FSAR Dose Analysis considerable conservatism is introduced into the FSAR analysis of ECCS leakage by assuming that elemental iodine becomes airborne after release from the core. This conservatism results in radiological releases which are much larger than expected. Attachment II is an analysis entitled
" Iodine and Cesium Releases due to ECCS Leakage" performed by Fauske and Associates, Inc. This analysis addresses the concentration of iodine and cesium within reactor cooling water and its diffusion into the auxiliary building atmosphere at Byron Station. The analysis generates more realistic values of the radiological release which would exist during a loss-of-coolant accident based on component leakages of 1 gpm for 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> and 50 gpm for 30 minutes. These more realistic numbers are small compared to those l
generated by the PSAR analysis.
7,
_..v.
v_
,_...-.,_.,,._m_
w..
- m..m m,.r_._...
. _. The Fauske analysis demonstrates that with the conservatism removed, the radiological releases from the auxiliary building can be reduced at least by a factor of 10.
This reduction in source term is equivalent to that resulting from the operation of the auxiliary building ventilation system filters. As a result it should be possible to operate the reactor at 100% power with the auxiliary building ventilation system inoperable (i.e., no HEPA and charcoal filtration) without exceeding offsite and control room dose limitations. The Fauske analysis provides additional assurance that Byron Unit 2 can operate safely per this interim operation plan.
Summary Unit 1 and common portions of the auxiliary building ventilation system at Byron currently operate to support Unit 1 operation. Balancing and testing of Unit 2 portions of the system will be completed at various stages in the Unit 2 startup program. Adequate filtration and cooling provisions will exist in the auxiliary and fuel handling buildings to support Unit 1 operation during Unit 2 startup. prior to the completion of the Unit 2 startup program, the entire auxiliary building-ventilation system will be integrally balanced and preoperationally tested as required. A temporary technical specification is proposed to allow operation of Byron Unit 2 while testing of Unit 2 portions of the auxiliary buidling ventilation system is being completed.
i i
i 2172K
s 100.
s 90.-
\\
g LPZ L'IMIT > 100 l
s s
s s
.x 80.-
\\
ERB LIMIT s
a
's]
s s
s 70.-
N s, W
s J
's, m
6 0. --
~
E CONTROL ROOM E
00SE LIMIT 50.-
n E
co o
~ ~ ~,-
cc 40.-
W M
E 3 0..'
RVRILRBLE r
~
POWER x
RANGE ct 20.-
r
~
10.-
~
O.
,,,,i...,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,
D.
5.
10.
15.
20.
25.
30.
35.
40.
45.
50.
55.
60.
RUXILIRRY BUILDING ECCS LERK RATE (GRL'/HR)
FIGURE 1.
MAXIMUM POWER LEVEL VS. ECCS EQUIPMENT LERKRGE Assumptions:
- 25 CFM Control Room Unfiltered Inleakage
- Thyroid Dose Calculations Include Contributions From:
- 1. Containment Leakage;
- 2. Massive ECCS Leakage (50gpm for 30 min. at T=24 hours);
- 3. Thirty (30) day ECCS Continuous Leakage ac shown on curve.
x
ATTACHMENT I OPPSITE AND CONTROL ROOM DOSES BYRON STATION The operation of the auxiliary building ventilation (VA) system directly affects the radiological doses both offsite and in the main control room. Air exhausted from the auxiliary building is filtered and released to the atmosphere. Potential outside air leakage into the control room boundary through normal mode and purge mode intake dampers which are closed during emergency operation, is essentially precluded by the existance of bubble tight dampers. However, unfiltered inleakage from areas surrounding the control room is conservatively assumed.
The radiation levels of the air exhausted from the auxiliary building is directly proportional to the power level of reactor operation and the assumed ECCS equipment leak rate. The HEPA and charcoal filters in-the auxiliary building ventilation system will operate to maintain radiological releases at a minimum during a loss-of-coolant accident at high power levels.
It is not necessary for these filters to operate at low power levels due to reduced fission products. Calculations using regulatory source terms have demonstrated that the control room dose and offsite dose can be maintained within acceptable limits without the auxiliary building ventilation system operable at low reactor power levels. The following is a summary of those calculations:
A.
Offsite Doses The offsite radiological consequences of a design basis loss-of-coolant accident have been calculated using the methodology described in the NRC Standard Review Plan, Section 15.6.5, Appendices A and B.
The results of the Exclusion Area Boundary (EAB) and Low Population Zone (LPZ) doses were given in FSAR Table 15.0-11.
The meteorology of Table 15.0-13 was used.
If the VA system is not tested and certified to remove postaccident radioiodine, Appendix B of SRP 15.6.5 requires different ECCS leakages to be considered. The leakages are as follows:
Applicant assumption 3910 cc/hr or about (FSAR Table 15.6-15a) 1 gallon / hour NRC assumption with VA 1 gallon / minute non-operable: continuous leakage, 30 days Massive leakage 50 gpm for 30 minutes at t = 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1
's The resulting thyroid doses at the EAB and LpZ are as follows:
Thyroid dose, rem EAB LPZ With Applicant's values and no VA 115 16 With NRC assumptions and no VA 579 117.5 10 CFR 100 limit 300 300 Thus according to the NRC assumptions, the power level would be limited to 300/579 or about 52% to meet the 10 CPR 100 Itnit of 300 rem at the EAB.
B.
Control Room Dose Analysis of the control room dose during a loss-of-coolant accident is discussed in Section 6.4.4.1 of the Byron /Braidwood FSAR. This analysis assumed all engineered safety systems to be operable and is the base case for NRC safety evaluation.
If the auxiliary building ventilation system is assumed to be inoperable (no HEpA and charcoal filtration is provided), then the above analysis is repeated with the following assumptions:
1.
Containment leakage as identified in the Byron /Braidwood FSAR; 2.
10% of iodine released from ECCS equipment leakage is assumed airborne.
3.
Fluid leakage from the er.ergency core cooling system (ECCS) is assumed as follows:
- 1 gallon per minute (gpm) leak for 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> (continuous leakage);
- 50 gpm leak for 30 minutes at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (massive leakage).
(If the VA system is operable, the only leakage term is a continuous 1 gallon per hour.)
.-.-.--,,_.-~v
- All calculations were performed per the Murphy-Campe paper. The results are shown in Figure 1.
These figures present thirty day control room dose at 100% power as a function of unfiltered inleakage for each of the three contributing source terms: containment leakage, continuous ECCS l
leakage (both 1 gph and 1 gpm are shown), and massive ECCS leakage. Figure I combines the doses for the three source terms (containment leakage, continuous ECCS leakage and massive leakage) and, in addition, shows maximum allowable power levels that can be achieved without exceeding the 30 rem dose limit. Figure 1 shows that with unfiltered inleakage of 25 cfm, the reactor is limited to essentially 30% power.
l l
l l
l 21~12K
877/1(.//MEAf7 1
4 i
i FIGURE 1.
POSTACCIDENT CONTROL ROOM DOSE (calculated using USNRC assumptions) 104 MLtow en b
~
Sh Maiwd + e um recs t kw
"",%}
1 I wm tee s 2-
_ _ _ r _ %_ _
103
_- deh:n. + + iMkEcc5+)bsw
_ _ _ _ _ _ 5_ _% -
5-E C rki= =,4
~
a:
lo sa n
2-E
_ _ _22_E.
o w
10g Bo g
f' _
m 3
x
_ _ ___ _ _ r_ _ A _
r e,
m 5-O er leo O/o -
2-10I -
.}ia,< ;ve 5-
\\
2-I $PL Ece s 100
' y ' l 'i'g s ilel 1 ' y ' l ' i 'g I IIl 2 'f'i'i'gisiij 3
i ' i 'g i s i i 10 10 10 103 104 UNFILTERED IN-LERMAGE RRTE. CFN
l s
l l
1 i
ATTAQ9 mrt II 0
9 IODINE AND CESIUM RELEASES DUE TO ECCS LEAKAGE
/*
Pauske and Associates, Inc.
Sargent & Limdy f
/
t e
e e
h 2172K l
.f FAlla4 5g' l
(
)
Fauske & Associates, Inc.
16WO70 West 83rd Street Burr Ridge, Illinois 60521
~
(312) 323-8750 IODINE AND CESIUM RELEASES DUE TO ECCS LEAKAGE Sargent & Lundy December,1984
(
J.
14010M9(o
1.0 INTRODUCTION
Licensing requirements for the Byron Nuclear Power Station required an assessment of the potential leakage of iodine and casium into the auxiliary building environment for design basis accident conditions.
The leakages to be considered are 1 gpm for a 30 day period and 50 gpm over a 30 minute interval as a result of potential leaks in charging pumps. RHR pumps, etc.
The re-leases of iodine and cesium within the core are those assumed for Itcen cases, i.e. half of the core inventory of each of these radioactive elements.
For these analyses, the chemical state is considered to be iodine gas and elemental cesium as opposed to the dominant and less volatile states of casiu iodide and cesium hydroxide.
Consequently, a considerable conservatism is introduced into the analysis as a result of these assumed chemical states.
4 The radioactive elements are assumed to be lost from the fue deposited within the emergency cooling water, which includes the primary.
systen water as well as the refueling water storage tank (RWST).
Considering half of the iodine to be released from the core, about 0.06 kg/ moles of fadine (7.5 kg) would go in the solution with 1.8 x 106 kg (484.000 gallons) of water.
The questions addressed will be the concentration of iodine and cesium within the water and its removal from the water by diffusion into the circu-lating air within the cell.
This is addressed for both leakage conditions.
It is recognized that elemental cesium could not be in solution with water.
3 Therefore, evaluations are also carried out for the most likely chemical forms of cestus iodide and cesium hydroxide dissolved in the water.
i l
--e-,r-e,----.
--n-,,-.-.n,,
2.0 BASIC CHARACTERISTICS FOR CESIUM AND IODINE In the initial analyses carried cut in this this report, iodine and cesium will be assumed to behave as elemental species I 2 and Cs.
The vapor pressures for these species are given by i
Iodine:
InP = - 6119/T + 24.81 (2.1)
Cesium:
InP = - 8513/T + 20.35 (2.2) where the pressure (P) is in Pascals and T is in degrees K.
These will be used to determine the partial pressure of these elements in an aqueous solu-tion.
The masses of the two elements considered are 7.5 kg of 12 and 83 kg of Cs which is typical of an equilibrium core cycle for a 1000 We plant.
amount is assumed to be released into 1.8 x 106 This kg of water during the opera-tion of ECCS functions forming an aqueous solution of the elemental specie The effective partial pressure of the iodine or casiism in the aqueous solution can be estimated through Raoult's law.
The expression for the partial pres-sure (PP) of a dissolved element (f) in solution is given by N
PP, = Psat(T)
(2.3) where P,,g(T) is the saturation pressure of tne element at the solution temperature (T). N9 are the moles of iodine in the solution and N total moles.
T are the At equilibritse. the partial pressure of the dilute species in the gas phase is equal to the effective partial pressure of the species in aqueous solution.
Genarally, this has _been characterized as a partitioning coeffi-cient (H) defined as the ratio of the material concentration in the liquid (C ) divided by the concentration in the gas phase (C ).
g g
H=C/C g g (2.4)
i The concentration in the gas phase can be characterized as the mass of the dilute species divided by the total gas volume, and assuming the species behaves as an ideal gas, this can be expressed as PP, Mw, C
=
g g
(2.5) where Mw is the molecular weight of t$ species, R in the universal gas g
constant and T is the absolute temperature.
Similarly/, the concentration in the liquid phase is the mass of the material in aqueods solution divided by the volume of the water as expressed by N Mw o CL*
M w ww (2.6) where N,, M, and o, are the number of moles, the molecular weight and density of water respectively. Using the Raoult's law for the effective pressure of the material in solution, and assuming N, = N, the partitioning coefficient T
under equilibrium conditions can be expressed as This predicted behavior can be compared to the measurements in the CSE experiments [1] in which elemental and particulate fodine was injected into a steam atmosphere along with cesium. uranium and ruthenfun.
The materials were i
accumulated in the sump water due to steam condensation, gravitational set-j t1ing and direct vapor condensation.
After about 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> the ratio of the concentration in the sump water to the concentration in the gas base remained essentially constant. This value can be compared with the prediction from Eq.
(2.7) to demonstrate the v'fability of using solution chemistry to evaluat'e the effective partial pressure of the various species in aqueous solution.
Such a comparison is given in Table 2.1 and is seec to be in genersi agreemer.t with the measured values, with the deviation between the experiments and predic-tions showing the material to be more tightly bound in the solution than predicted by the simple model.
This is generally attributed to reactions in the water that make the fodine less volatile than the consideration of
~..
J Table 2.1 COMPARISON WITH CSE IODINE RESULTS h
T Time C
C
,est g
g H
H K
Hrs.
u /l u /l Experimental Predicted g
g A-1 356 4
1.8 (-3)*
1.8 (2) 1 (5) 8.1 (4) 8 5 (-4) 1.2 (2) 2.4 (5) 8.1 (4) 12 3 (-4) 8 (1) 2.7 (5) 8.1 (4) 24 1 (-4) 4 (1) 4 (5) 8.1 (4)
A-2 358 4
7 (-2 1.3 (4) 1.9 (5) 7.4(4) 8 4 (-2 9(3) 2.2 (5) 7.4 (4) 12 3 (-2 7 (3) 2.3 (5) 7.4 (4) 24 1 (-2) 4 (3) 4 (5) 7.4 (4)
A-5 396 4
1.3 (-1) 2 (4) 1.5(5) 1.6 (4)~
8 7(-2) 1.3 (4) 1.9 (5) 1.6(4) 12 6 (-2) 9 (3) 1.5 (5) 1.6(4) 24 4 (-2) 7 (3) 1.8 (5) 1.6 (4)
A-ll 395 4
1.9 (-1) 1.4 4) 7.4(4) 1.6(4) 8 8
1.3 4) 1.6 ( 5) 1.6 (4) 12 5
1.2 4 2.4 1 5) 1.6(4) 1 (4))
24 3
3.3(5) 1.6 (4)
- 1 8 x 10~3 l
l l
e elemental iodine.
As a result, the effective partial pressure of the species in an aqueous solution would be from simple solution chemistry.
It should te remembered that in this experiment, elemental fodine was injected directly into the gas base as opposed to the postulated accident case in which traterial removed from the fuel matrix would have the dominant chemical for iodide and cesium hydroxide.
Therefore, the comparison with the CSE expert-ments should be viewed as a qualification of the solution chemistry approach and its conservatisms, but not an indication of the dominant chemical states.
A similar calculation can be carried out assuming that cesium could exist in solution, which is conserv&tive, but not physically possible.
For the CSE tests the gas phase did not contain measurable cesium vapor, but cesium hydroxide in particulate form.
With this approach to the effective vapor pressure, the driving force for diffusion of vapor /s from the ECCS leakage can be estimated for the conditions of interest.
Since the spills onto a cubicle floor would be removed to holding tanks via floor drains, the key element of the analyses is the rate dependent process of diffusion from the liquid into the gas phase over the time interval of' interest, i.e. 30 minutes fcr a large spillage and 30 days for the 1 gpm leak rate.
O
(*
.-----..---.m.-.-
e m
3.0 _ RATE DEPENDENT PROCESSES The diffusion from the spilled liquid can be estimated from PP N /A = h y
(3.1) where N is the rate of moles diffused for the t$ species, A is the surface 9
area for diffusion. D is the diffusivity in the gas phase, R is the universal gas constant, T is the absolute temperature of the liquid, PP is the partial g
pressure of the species in aqueous solution and 4 is the diffusion boundary layer.
In this calculation, the diffusion rate dependent process in the l
liquid phase is ignored as is the partial pressure of the species in the surrounding gas phase. Both of these make the analysis somewhat conservative, i.e. the diffusion rate will be overestimated.
Table 3.1 gives the asstamp-tions of material quantities for core inventory, water inventory and the concentration of the iodine should complete ionization occur. Table 3.2 lists l
the assumptions in the FSAA analysis (905 retention by the liquid and 90%
filtration by the auxiliary building ventilation system) to estimate the quantities of iodine mass lost to the environment.
This value of 0.15 g of iodine can be compared to the calculations resulting from evaluating the rate dependent processes.
Figure 3.1 shows an assumed configuration for a continuous leakage of 1 gpm for 30 days in which the stream would pour onto the floor and into the floor drains.
As a result of the leakage, the stream can be viewed as a continually resupplied column of liquid such that the concentration of the j
dissolved species is constant over the interval of interest.
For these calculations, a diffusion boundary layer of 2 as is assumed in the gas phase, and the material pouring onto the floor is assumed to have an average ve'ocity of 1 m/sec to account for run off on equipment and running onto the floor.
With this velocity, the effective diameter of the stream is about I cm and heat transfer coefficients show that the stream temperature would decrease only slightly during its resonance time in the equipment cell.
Consequently, a temperature of 373'K is assumed for the water during its residence interval.
Table 3.3 lists the saturation pressure of fodine at this temperature, the
[
\\
1 Table 3.1 ASSUMPTIONS Core inventory 1 - 15 kg Cs - 166 kg Released from Fuel 1 - 7.5 kg Cs - 83 kg Water inventory 64,740 ft3== 484,000 gal.
6 5
1.8 x 10 kg
=
10 kg moles Moles of l~ = 0.058 kg moles l
Concentration = O 0 *
~7 j05
= 5.9 x 10 i
Spillage ~ 1000 gal. =. 3636 kg
= 202 kg moles
~4
= 1.2 x 10 kg moles l'
/d m-
i Table 3.2 FSAR ANALYSIS
-5 Westinghouse FSAR Analysis
= 1.2 x 10 kg moles released to the auxiliary building atmosphere
= 1.2 x 10-6 kg moles released to the environment
= 1.5 x 10-4 kg 0.15 g
=
i i
--- -...-..-- - ~ -
f CONTINUOUS LEAKAGE i
4 l
V (1 gpm)
(30 days)
Fig. 3.1 Assumed configuration for continuous leakage.
E]\\
J
[
Table 3.3 e
CONTINUOUS LEAKAGE (1 com for 30 days)
Assumed Surface Area = 0.063 m Diameter of Water Stream ~ 0.01 m interval = 2.6 x 106 seconds T = 373 K P (T) = 4470 Pa g
-3 lodine PP (T) = 2.6 x 10 Pa g
N/A = 4.2 x 10-12 2
kg moles /m /sec rii, = 3.4 x 10~I I kg/sec
-5 Amg = 8.7 x 10 kg = 0.087 g Cesium lodido P,g(T) = 2.1 x 10* Pa PPcl(T)
=
1.2 x 10-18 Pa
~
N/A = 2 x 10-28 2
kg moles /m j,,,
-27 di ;
= 3.3 x 10 kg/sec c
-21 Amg=
8.5 x 10 kg e
+
v
partial pressure of the material in solution, the diffusion rate per unit area, the rate of iodine lost to the atmosphere and the total lost over tr,e entire 30. day interval.
As shown, this value is comparable to..but less tnan the value used for the FSAR analysis when the building ventilation system is credited for removing 90% of the airborne fodine.
This calculation demon-strates that the solution chriistry is very effective in retaining the fission products and would not release substantial quantities to the cubicle environ-ment, even when elemental f adine is assumed as the chemical state.
Table 3.3 also illustrates the influence of the chemical state by demonstrating the amount of mass lost for cesium todide held in solution, which essentially shows only negligible quantities to be released.
Obviously, the actual release would be larger than that calculated assuming cesium todide as a result of secondary reactions in the water, but the release value would be less than that calculated assuming elemental fodine to be in solution, Release calculations for cesium and cesium hydroxide are listed in Table 3.4, 1
and these are well within the tolerable levels.
i Figure 3.2111ustrates the configuration for a large spillage of 50 gpm i
over a 30 minute interval. The assumption is that the spill rete is suffi-cient to accumulate a layer of water on the cubicle floor before it runs off into the floor drain.
As a result, the surface area for diffusion is equal to the cubicle floor area and is taken to be 30 m2 in this analysis. While the surface area is larger, the time available for diffusion equals the spill time of 1800 seconds and Table 3.5 lists the assumptions and the results for assuming the chemical state is iodine and cesium fodide.
The net result of assuming elemental iodine and solution is a release quantity which is approxi-mately 205 of that calculated in the FSAA and the assumption of cesium todide results in an extrasely small amount of material released.
Table 3.6 provides calculations,fer the assumptions of elemental cesium and cesium hydroxide under the same conditions, and as shown, reveal that cesium would be tightly bound in the water regardless of the assumed chemical state.
It should again be noted that elemental cesium could not be in solution with water and hydroxide is the domirant chemical state.
These analyses demonstrate that the assumed spill rates do not result in substantial release of iodine or cesium to the cubicle environment even if
Table 3.4 CONTINUOUS LEAKAGE T = 373 K Cesium P (T) = 0.084 Pa PP (T) = 5.3 x 10-7 Pa g
N/A = 8.5 x 10-16 kg moles /m /sec 2
A = 7.1 x 10-15 kg/sec g
Am = 1.9 x 10-8 kg e
Cesium Hydroxide CSON(T) = g 1 x 10*8 P
C50H(T) = 5.7 x 10*I4 PP Pa N/A = 9.2 x 10-23 gg,,j,,f,2/sec AC50H = 7.7 x 10-22 kg/sec-0*Cson
- 2 x 10*II kg
+
i
f
\\
f LARGE SPILLAGE t
U A
A A
A 11 V
50 gpm 30 min.
i Fig. 3.2 Assumed configuration for a large 1
spill on a cubicle floor.
=
(
O
,,,,-n
-,---n.n
i Table 3.5 LARGE SPILLAGE (50 com for 30 minutes)
RHR Pump Cubicle 20' long x 15' wide Area = 300 ft2=
30 m2 Mechanism - Diffusion from the water pool with an 2
area of 30 m and an interval of 30 min.
4 i
T = 373 K lodine P (T)
= 4470 Pa g
PP (T) = 2.6 x 10-8 g
Pa
-12 2
N/A = 4.2 x 10 kg moles /m j,,,
-8 h
= 1.6 x 10 kg/sec g
-5 Amg = 2.9 x 10 kg = 0.029 g
~18 Cesium lodide Pag (T) = 2.1. x 10 Pa i
~I8 PP,g(T)
= 1.2 x 10 Pa
-28 2
N/A =
2 x 10 kg moles /m f,,,
-24 ag = 1.5 it 10 kg/sec W
m Amci *
- I * ' O kG k
l l
1
^
Table 3.6 LARGE SPILLAGE T
= 373 K Cesium P,(T) = 0.084 Pa
~I PP,(T)
= 5.3 x 10 Pa
-16 2
N/A = 8.5 x 10 kg moles /m f,,,
-12 l
rhc
- 3.4 x 10 kg/sec Am, = 6 x 10-9 kg Cesium Hydroxide PCsOH(T) = 0.1 x 10 ' Pa
~
-14 PPCsOH(T) = 5.7 x 10 Pa
-23 N/A = 9.2 x 10 2
kg moles /m 7,,,
thCsOH = 4 x 10-18 kg/sec
-N AmCsOH = 7 x 10 kg.
i
")
i
elemental form is assumed to be in aqueous solution. Several conservatisms are inherent in the analyses as carried out and these are delineated in Tab'e t
3.7.
The first is that an equilibrium core cycle was assumed and this como overstate. the amounts of fission product material available in the early stages of the Byron plant operation by a factor of 2 to 10.
In addition, it is assumed that 50% of the iodine and cesium fission product material are re-leased from the fuel matrix, which for a design basis accident could overstate the fission products in aqueous solution by a factor of 10 to 100.
The
~
comparisons with the CSE experiments demonstrate that the partition coeffi.
cient is larger than calculated by simple aqueous solution chemistry, which could potentially decrease the ratec of materials lost to the enviroment by a factor of 2 and perhaps as much as 10. Lastly, the dominant chemical states, which is not an independent change from the partitioning coefficient listed in Item 3 decrease the release by orders of magnitude.
Considering low and of the significance of these conservatisms and neglecting the dominant chaeical state, the analyses are conservative by at least a factor of 60 if more realistic assessments were applied to the actual core cycle, the material-released from the fuel matrix and use of an experimentally detemined parti-tioning coefficient from the CSE tests.
However, the analyses already show that the efficient retention of fission products in aqueous solution are equal l
to the releases considered in the FSAR, hence further refinement of the calculations would not appear to be warranted.
i j
1 I
._____.----__.-_-___,__,.m_--
[
Table 3.7 CONSERVATISMS IN THE ANALYSES Type Significance 1.
Equilibrium Core Cycle 3 - 10 2.
50% 1 and Cs Released From 10 - 100-the Fuel l
l 3.
Neglect Partitioning Coefficient 2 - 10 for lodine 4.
Chemical State Col and CsOH Several Orders of Magnitude
(
)
-w w,_,e---,-.-.e.,,ne,.en.,__e___.,,.wv.w,,-,,-
._.,.,_m_,,
w.,,..,,___,,_,.,,,_
4.0 REFERENCE 1.
R. K. Hilliard and L. F. Coleman. " Natural Transport Effects on Fission Product Behavior in the Containment Systems Experiment," BNWL-1457, Decemoer.1970.
I e
o e
r i
-- ~.,.,,,, - -.. ~..,
,--.,_,_r,,__,,-_-,.,_---.-_ve-,--,._,,,_w_,,
-,n
.n,.
D ATTACHMENT III TEMPORARY TECHNICAL SPECIFICATION 4.7.7.7.d.3 1
)
i 2172K
PLANT SYSTEMS SURVEILLANCE REOUIREMENTS (Continued) 3)
Verifying a system flow rate of 66,900 cfm i 10% through the train and 22,300 cfm 10% per bank through the exhaust filter plenum during operation when tested in accordance with ANSI N510-1980; and 4)
Verifying that with the system operating at a flow rate of 66,900 cfm' 10% through the train and 22,300 cfm *10% per bank and exhausting through the HEPA filter and charcoal adsorbers, the total bypass flow of the system and the damper leakage is less than or equal to 1% when the system is tested by admitting cold DOP at the system intake and the damper leakage rate is determined by either direct measurements or pressure decay measurements at a test pressure of 2 inches of water and the auxiliary building exhaust fans are operating at their rated flow.
c.
After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation, by verifying, within 31 days after removal, that a laboratory analysis of a repre-sentative carbon sample obtained from each bank of adsorbers of the train in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing cri-teria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978, when the average for a methyl iodide penetration of less than 1% when tested at a temperature of 30*C and a relative humidity of 70%.
d.
At least once per 18 months by:
1)
Verifying for each filter bank of the train that the pressure drop across the combined HEPA filters and charcoal adsorber banks of less than 6.0 inches Water Gauge while operating the exhaust filter plenum at a flow rate of 66,900 cfm + 10% through the train and 22,300 cfm 210% per bank; 2)
Verifying that the exhaust filter plenum starts on manual initiation or Safety Injection test signal; and
- 3) M Verifying that the system maintains the ECCS equipment rooms at a negative pressure of greater than or equal to 1/4 in. Water Gauge relative to the outside atmosphere during system operation while operating at a flow rate of 66,900 cfm 10% through'the train and 22,300 cfm 210% per bank.
e.
After each complete or partial replacement of a HEPA filter bank, by verifying that the exhaust filter plenum satisfies the in place penetration testing acceptance criteria of less than 1% in accord-ance with ANSI N510-1980 for a DOP test aerosol while operating at a flow rate of 66,900 cfm : 10% through the train and 22,300 cfm 10%
per bank; and 4
% + a p he a le. k Unit 2. unHI July I, In7.
BYRON - UNITS 1 & 2 3/4 7-20