ML20031C440
| ML20031C440 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 10/02/1981 |
| From: | Delgeorge L COMMONWEALTH EDISON CO. |
| To: | Schwencer A Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0519, RTR-NUREG-519 NUDOCS 8110070146 | |
| Download: ML20031C440 (24) | |
Text
.
'N Commonwealth Edison
) one First National P! za. Chictqo, Illinois O
C 'J Address Reply to: Post Office Box 767
- hf[
3(j Chicago, Illinois 60690 e
all/(r b CT lh October 2, 1981 '
2 O
G 7 5 y% #o8/A -
)
Mr. A. Schwencer, Chief Licensing Branch 2 9
Division of Licensing y/fg g Q U.S. Nuclear Regulatory Commission Washington, DC 20555
Subject:
LaSalle County Station Units 1 and 2 Request for Clarification to NUREG-0519 NRC Docket Nos. 50-373/374
Dear Mr. Schwencer:
The purpose of this letter is te request clarification of the LaSalle County SER (NUREG-0519).
The specific areas of interest are delineated below, along with the attachment to this letter in which the issue is discussed:
1.
Section 4.6.2 (Pg. 4-30) - CRD Return Flow (Attachment 1).
2.
Section 6.2.1 (Pg. 6-33) - Secondary Containment In-Leakage (Attachment 2).
3.
Section 6.4.1 (Pg. 6-56) - Control Room Emergency Filters (Attachment 3).
i 4.
Section 9.4.6 (Pg. 9-12) - Emergency Switchgear Heat Removal (Attachment 4).
5.
Section 22.II.E. (Pg. 22-66) - Isolation of Essential and Non Essential Systems (Attachment 5).
Ear.h of these items is discussed in detail in the attachment noted.
We request that you advise us whether the SER requires revision based on the information provided.
In the event further information is required of us in order to complete your review, please advise us.
Furthermore, in the event you determine that no change in the SER is required or is justified please advise us in writing of the basis for that conclusion so that your formal interpretation can be used to resolve any future inquiries from the regional office of Inspection and Enforcement.
If you have any questions, please direct them to this office.
Sool Ve y ur,
5 5
.* DelGeorge l (
L. O Director of Nuclear-Licensing Attachments no.' one unn* Inspector - LSCS 6110070146 811002 POR ADOCK 05000373 E
.PDR
CRD Return Flow (Section 4.6.2, Page 4-30)
To resolve the Concrol Rod Drive Return Line (CRDRL) nozzle cracking problem, LaSalle implemented the total removal of the CRDRL and capping cf the nozzle in accordance with NUREG-0619.
This implementation and alternatives (GE CRDRL study recommendations) to the CRDRL concern are documented in NUREG-0619, "BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking -
Resolution of Generic _ Technical Activity A-10," November 1980.
The staff's conclusions and recommendations for implementation are presented in Part II of NUREC-0619.
The staff hes concurred with the " cut and cap" action plan for LaSalle (251-inch BWR/5) but has required plant-specific testing to assure proper system operation and return flow capability.
In considering the removal of the return line, the NRC Staff expresscd concern with regard to the effect of the change opon the makeup flow capability..., as stated in NUREG-0619, "The major Staff concern regarding the final recommendation was the loss of a portio-of the high-pressure return-flow capability to the reactot vessel.
Based on this concern, the staff has concluded that the GE ' cut and cap' recommendation is only acceptable for certain classes of BWRs, and only for these after specific modifications have been made and operability testing completed.
Goerability testing should include flow-capability testing in the form of a demonstration of simultaneous two-pump coerability during which flow measurements are recorded."
l The NRC Staff recognizes that the presence of the CRD system's high-pressure flow capability has not been directly assumed in previous safety analyses and no credit has been taken in safety analysis.
Also, the water makeup capability is not a design requirement of the CRD system.
CRD water makeup capability is insignificant in terms of reactor vessel water inventory makeup when I
overall water makeup availability, i.e.,
normal vessel inventory and ECCS capability is considered.
l The major deterrent to the NRC granting permission initially _ for the removal of the CRDRL was the Brown's Ferry Unit 1 fire incident in 1975.
Today, we believe a recurrence is incredible because of improvements to BWRs like LaSalle...
i.e.,
i
... personnel procedural controls to restrict open flame usages."
... flame retardance, fire barriers, and cable trays -
redundancy and separation."
w... administrative procedures specifically to cover BF-1 l
fire scenario in addition to safe shutdown,"
These improvements and concepts are documented in the NRC review and evaluation of the LaSalle fire protection progra:s (LaSalle SER 9.5).
However, acknowledging the Staff's NUREG-0619 criteria which were based upon the Browns Ferry fire scenario, i e.,
require by test to demonstrate adequate return flow capability to keep the core from uncoverning; we believe from the previous discussion that normal reactor water level can be maintained up to three hours or more, taking credit for the accepted fire protection improvements.
Given this postulate, at least three hours will pass before
[
complete loss of normally available RPV it.ventory makeup capability 9
can occur.
Test of CR0 makeup capability at LaSalle demonstrates a e#
CRD makeup flow capability of approximately 130 gpm.
Based on these facts, analysit demonstrates that approximately five hours will pass from the time of complete loss of normal water inventory makeup capability before water level reaches the top of active fuel.
(See Table 1).
We feel that this allows ample time for operator actions ta depressurize the RPV allowing even greater makeup flow throug., the CRD drives (260 gpm at 0 1
pressure verified by test).
In addition, other corrective actions to make normal sources of makeup water available can be taken during this time period.
Therefore, we conclude that LaSalle CRD system makeup capability satisfies the intent of NUREG-0619.
Table 1 Assumed Time (To)
Time (I TAF) at which Time (TTAF-To) for Loss of All Makeup Water level at Top of Available for Except CRD Active Fuel Corrective Action I hour 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> 45 minutes 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 46 minutes 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 6 hours 2 minutes 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 2 minutes 3* hours 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 25 minutes 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> 25 minutes
- minimum of 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> assumption justified based on fire protection i
program implementation.
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In as much as the return flow capability discussed in SER Section 4.6.2 is not a design basis raquirement, it is judged that the actual flows determined by test at LaSalle County Unit 1 are adequate and satisfy the concern raised in NUREG-0619.
NRC Staff concurrence in this interpretation is required.
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K Secondary Containment In-Leakage r
e p-(Section 6.2.1, Page 6-33)
The NRC Staff required *that "a leak test of the Secondary d'
Containment to verify the inleakage assumption and the drawdown time in reestablishing the -0.25 inches of water guage" be done at LaSalle County (O. D. Parr letter-to L. O. DelGeorge dated March 7, 1979).
The purpose of this requirement-was to demonstrate the caoability of the Standby Gas Treatment System to overcome the
" combined inleakage" while maintaining a vacuum 0.25 inch w.g. at N
its design flow of 4000 cfm 110%.
Moreover the SBGTS was to be capable of drawing down the Secondary Containment to -0.25 inch w.g.
' in 5 minutes.
This capability shall be. demonstrated prior to fuel loading on LaSalle County Unit 1 as a part of Preoperational test
,g PT-VG-101 (FSAR Table 14.2-20).
Analysis for LaSalle County, documented in FSAR Section
- =
6.5.1.1..l.c.2, indicated that the Secondary Containment combined in-leakage rate,is 2000 cfm at still wind conditions with ad P of
-0.25 inch w.g.
Verification of this specific number can not be done emperically in as much as there is no way to' measure actual in-leakage, However, performance of the preoperational test discussed previously does verify that the combined inleakage can be
^
accommodated and that the appropriate vacuum can be restored within 5 minutes.
g In as much as the actual inleakage number is of no importance if the SBGTS' test is satisfactory, emperical verification e
of the inleakage assumption need not be addressed further.
Staff concurrence in this interpretation is required.
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i Control Room Emergency Filters
_(Section 6.4.1. Page 6-56)
The description of the control room emergency filters discussed in the referenced section of NUREG-0519 appears to be in 3
error, i.e..it is inconsistent with information provided in the LaSalle' County FSAR.
A suggested clarification to the effected SER pages and the supporting FSAR pages are provided to facilitate your review.
NRCLStaff concurrence in this observation is required.
There are certain editorial changes provided to pages 6-53 to 6-55 of NUREG-0519 which are also attached for your consideration.
These '
were discussed with Mr. A. Bournia in April, 1981 and not reflected in Supplement 1 to NUREG-0519.
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TARALLGl.
% n-smsr coil, prefilter, upstrea high efficiency particulate air filter, t;; f;r inch deep carbon adsorbers i W and a downstream hi,t efficiency particulate t
air filter.
The equi;, ment and components are seismic Category I design and are located in a seismic Category I structure.
For our evaluation of the engineering cafety feature filter systems in Section 15 of this report, we have assigned the standby gas treatment system filter /adsorber trains with a removal efficiency of 99 percent for all forms of radioiodines and 99 percent for particulates as recommended in Regulatory Guide 1.52 for greater than or equal to 4 inches deep charcoal beds.
We determined that the standby gas treatment system is designed in accordance with the guidelines of Regulatory Guide 1.52 and is capable of maintaining suit-able control of gaseous effluents following a design basis accident.
We, therefore, find the design of the system acceptable.
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6.6 Inservice Inspection of Class 2 and 3 Components Criteria 36, 39, 42 and 45 of the General Design Criteria require, in part, that the Class 2 and 3 components be designed to permit appropriate periodic inspectica of important component parts to assure system integrity and capability.
Section 50.55a(g) of 10 CFR Part 50 defines the detailed requirements for the preservice inspection programs for light water cooled nuclear power facility components.
Based upon the construction parmit issuance date of September 10, 1973, this section of the regulations requires that a preservice inspection program be developed for Class 2 and 3 components and be implemented using at least the Edition and Addenda of Section XI of the American Society of Mechanical Engineers Code in effect six months prior to the date of issuance of the con-struction permit.
Also, the initial inservice inspection program must comply with the requirements of the latest Edition and Addenda of Section XI of the American Society of Mechanical Engineers Code in effect 12 months prior to the date of issuance of the operating license, subject to the limitations and modi-e ficaticas licted in Section 50.55s(b) of 10 CFR Part 50.
C.6.1 Evaluation of Compliance for Unit No. 1 to 10 CFR 50.55a(g)
A preservice inspection program for Urdt No.1 based on the 1974 Edition through Summer 1975 Addenda of Section XI of the American Society of Mechanical Engineers Code was submitted by the applicant in a letter dated March 15, 1979.
Specific written relief from Code requitements was requested and supported by information in letters dated Deceb er 12, 1980, February 3, 1981, and February 4, 1981.
We have determined that certain American Society of Mechanical Engineers Code Section XI examination requirements defined in 10 CFR 50.55a(g)(2) are impractical.
We have evaluated the American So:iety or Mechanical Engineers Code required exminations that have been determined to be impractical and, pursuant to 10 CFR Part 50.55a(a)(2), have allowed deviations from the requirements that have been determined to be impractical and that if implemented would result in hardships
/
or unusual difficulties without a compensating increase in the level of quality and safety.
Based on the granting of relief from these preservice examination requirements, we conclude that the preservice inspection prograr for Unit No. 1 6-56 L~
LSCS-FSAR AMENDMENT 50 OCTOBER 1980 TABLE 6.5-1 (Cont'd)
TYPE, QUANTITY AND NOMINAL CAPACITY NAME OF EQUIPMENT (per component)
- f. Charcoal Adsorber Bed Type Verticcl gasketless 3 %
Quantity 4 - 8 in, thick I
,)
Media Impregnated Char-coal Iodine Removal Efficiency 99.8 on Methyl
(%)
Iodine (Purchased) 90.0 (Operational Requirement) 99.9 oa Elemental Iodine (Purchased) 90.0 (Operational Requiremen t)
Quantity of Media (Ib) 3800 Media Density (lb/ft )
30 Depth of Bed (in.)
8 Residence Time for 8 in. bed (sec) 2.0 Charcoal Ignition Temperature (*C) 340 Static Resistance (in. H 0) 4.6 l
2 g.
Standby Cooling Air Fan Type Centrifugal Quantity 1
Drive Direct Capacity (ft / min) 200 Static Pressure (in. H O) 5 l
2 B.
Secondary Containment Isolation Dampers l.
Equipment Numbers IVR04YA&B, IVR05YA&B 2VR04&A&B, 2VR05YA&B 6.5-19
i
.t I
t the results of these tests when they are submitted.
In the interim period, we i
believe there is a sufficient technical basis to permit licensing af La Salle, i
This conclusion is based on:
(1) The existence of safety margins between available and required spray flow indicated by preliminary analyses and measurements, (2) The existence of counter-current-flow-limiting phenomena should provide a steam / water layer on top of the core and force even distribution cf the core spray, (3) The timely confirmation cf the spray flow margin presently believed to e.<ist which should be provided by the aforementioned tests and information.
G.3.5 Conclusions Wa reviewed piping and instrumentation drasings and the description of the q
emergency core cooling system presented in the Final Safety Analysis Report.
4 4
We find the design of the system acceptable becau:e it conforms to the pertinent regulatory guides, Standard Review Plan and General Design Criteria.
r In addition, based on the discussion above, we find the performance of the emergency core cooling system acceptable because it conforms with the require-j ments Of 10 CFR Sect 4n 50.46.
j e
3 6.4 Control Room Habitability Systems a
~
We reviewed the applicant's control room habitability systems with respect to i
Criterion 19 of the Ger,eral Design Criteria.
The following describes our evaluation in terms of radiological and toxic gas hazards.
R 6.4.1 Radiological Protection 1
The applicant proposes to meet Criterion 19 of the General Design Criteria by use of concrete shielding and by installing redundant 4000 cubic feet per minute cnce-through charcoal filters in the habitability ventilation system.
The habit-J ability system consists of two separate ventilation envelopes, one servicing l
the control rocm area and the other seivicing the auxiliary electrical equipment In the event of high radiation detection-4* the cut 3ide air intakes, 7
crea.
the radiation monitoring system will initiate emargency ventilation by automa-i tically isolating the habitability system and routing some nutside air flow
-l through a once-through emergency charcoal filter train. A portion of the filtered airflow (1500 cubic feet per minute) will be directed to the control j
room heating, ventilating and a.ir conditioning system in order to pressurize 1
the control room.
The remaining 2500 cubic feet per minute are used to pres-surire the heating, ventilating and air conditioning system for the auxiliary electric equipment room.
The radiation monitors h the outside air intakes provide the option of manual selection of the out side air intake with the lower
^
airborne radioactivity level.
Redundant, full-f' ow (20,000 cubic feet per cinute) recirculating charcoal filters are provi fed in the c trol room heating, ventilating and air conditioning systen for smoke an odor removal.
These filters are normally bypassed, but will be used in the c trol room z 4,340 66IM6 t
l 6-53 i
i 6.5.2 System Description and Evaluation 6.5.2.1 Control Room Heating, Ventilation and Air Conditioning System i
The function of the control room heating, ventilation and air conditioning system is to supply non-radioactive air to the common control room for Unit Nos. I and 2 after a design basis accident and to pressurize the control room
)
and the auxiliary electrical equipment room to a slight positive Jressure.
l This system will perait operating personnel to remain in the control room and electrical equipment area follo@g a design basis accident.
The emergency cckwp air filter train of the control room heating, ventilation and air con-ditioning system has redundant active components with an intake design capacity of 4,000 cubic feet per minute.
Each train contains the following components:
a fan, a heating coil, a demister, a prefilter, two high efficiency particulate air filters, and a two inch deep charcoal adsorber.
The normal air supply and return portion of the control room heating, ventilation and air condition system consist of redundant trains with an intake and recirculating design capacity of 20,000 cubic r minute.
Each train contains the following components:
a high efficienc culate air filter, a two inch deep charcoal adsorber, a humidifier, a he ing and a cooling coil, and a fan.
The equipment and components sre seismic Category I design and are lo:sted in seismic Category I structure.
Following a design basis accident the pressuriza-tion and recirculation system will be automatically activated by a signal from radiation monitors located 4e the inlet ducts or be activated '<1nually from the centrol room.
t hm For our evaluation of the emergency safety features filter systems in Section 15 m
of this report, we have assigned the control room heating, ventilation and air e
conditioning system filter /adsorber trains with a removal efficiency of 95 percent for all forms of radioiodine and 99 percent for particulates as recommended in Ic,;ulatory Guide 1.52, " Design, Testing, and Maintanance Criteria for Engineered hfety-Feature Atmosphere Cleanup System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants," for two inch deep charcoal beds.
We determined that the control room heating, ventilation and air conditioning system is designed in accordance with the guidelines of Reguiatory Guide 1.52 and is capable of maintaining a suitable control room environment following a design basis accident. We find the design of the system to be acceptable.
6.5.2.2 Standby Gas Treatment System Tha function of the standby gas treatment system is to prod ge and maintain a slightly negative pressure (-0.25 inches of water gauge) in the secondary con-tainment and to provide control of the releases of radioactive materials in gaseous effluents following a design basis accident, La Salle Unit Nos. I and 2, each have a system that is automatically activated by a high drywell pressure signal, low reactor water level, and airborne radiation monitors in the reactor building ventilation system, or can be manually controlled from the control room.
Each system has a treated exhaust capacity of 4,000 cubic feet per minute of air.
Interconnecting lines and controls are provided to permit either standby gos treatment system to be used with either Unit Nos. 1 or 2.
Each standby gas treatment system filter train consists of a fan, desister, electric heating 6-55
Emergency Switchgear Heat Removal
($ection 9.4.b, Page 9-12)
The description of the Emergency Switchgear Heat Removal System given in the referenced section of N'JRfG-0519 is not correct.
The description of the LaSalle County system is provided in FSAR Pages 9.4-34 through 36 and Figure 9.4-10 Sheet 1 which are attached.
Also attached is a proposeo clarification of Page 9-12.
NRC Staff concurrence in this interpretation is required.
l l
_\\
Based on our review, we conclude that the design of the fuel building ventilation system meets the recommendations of Regulatory Guides 1.13 and 1.52 m u is, therefore, acceptable.
9.4.5 Diesel-Generator _ Building Heating and Ventilation System The diesel generator building heating and ventilation system is designed to
}
maintain a suitable environment for the operation of tne diesel generators,
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the high pressure core spray pumps, and their auxiliary components during all modes of plant operation, including accident conditions.
Independent diesel-generator heating and ventilation systems and air supply and exhaust systems are provided for each of the five diesel generators and the two high pressure core spray diesel driven pumps to satisfy the required environmental conditions and combustion air requirements during diesel operation.
The diesel generator room ventilation system is designed to (1) seismic Lategory I W
requirements; and (2) maintain the diesel generator rooms below 122 degrees Fahrenheit whenever the diesel generators are in operation, The combustion air supply is drawn from the room ventilation air supply.
Neither reteorological 6- -
changes nor accida conditions can affect all-diesel air supplies.
The outside H
v air intakes and exnausts are tornado-missile protected.
3 5
g.i_W We reviewed the diesel generator building heating and vent'ilation systems design I
and conclude that they are acceptable.
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9.4.6 Emergency Switchgear Heat Removal System u
a g.
wk The ventilation system for the emergency switchgear area provices air to the i
hP emergency s it hgear rooms and the battery rooms for heat removal.
The system f
n4 consists of 100 percent-capacity seismic Category I ventilation systems W
for each switchgear room The battery rooms receive air from the switchgear
- k*k i The battery rooms are provided with separate exhaust fans so that they t
rooms.
can be maintained at a negative pressure with respect to the switchgear rooms.
The switchgear heat removal system removes heat from the switchgear rooms to maintain a temperature range of 65 degrees Fahrenheit to 104 degrees Fahrenheit.
r (t
Based on our evaluation, we have determi ed that the design of the ventilation ARE sorrectessrty stPARATED system for the emergency switchgear are :=t:N sfficist ;xpxxt Td" g
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"-- to meet the sin e failure criterion so that ventilation is assured during accident conditions r..&, therefore, the system is acceptable.
9.4.7 Emergency Core Cooling System Equipment Area Cooling System The emergency core cooling system equipment area cooling system consists of a fan-coil unit for each emergency core cooling system equipment cubicle except P
for the residual heat removal service water pump cubicles located in the diesel building.
Each system is seismic Category I and powered from the essential buses serving the cubicle f~om which the equipment is powered.
Full edundancy exists throughout the entire emergency core cooling system equipment area.
The seismic Category I core standby cooling water system is circulated in the e
cooling coils in order to limit the maximum room temperature to 148 degrees 3
Fahrenheit after a design basis accident.
Ventilation air for the emergency core cooling system equipment cubicles is provided by the redundant reactor building heating, ventilation, and air conditioning system.
1 9-12 W
LSCS-PSAR 9.4.5.2 Switchgear Heat Removal System The system serves the essential switchgear areas and battery rooms under normal and abnormal station conditions.
Outside air is provided for each unit via independent systems for switchgear heat removal.
Discharge air is vented to the air shaft which connects with the outside atmosphere through a missile protected louver.
9.4.5.2.1 Design Bases The system removes equipment heat to maintain temperatures in accordance with equipment requirements.
The system is an engineered safety feature.
9.4.5.2.1.1 Safety Design Bases u.
The switchgear heat removal systems for Units 1 and 2 are designed to operate under all plant operating conditions and to limit switchgear room temperatures to 104' F maximum and 650 F minimum.
Each system is designed to conform to Seismic Category I requirements.
b.
Each heat removal system is powered from the essential bus serving its associated essential switchgear.
The controls, instrumentation, and power supply for the system are designed to meet IEEE 279 and IEEE 308.
c.
The system exheusts sufficient air f rom the battery rooms of Unit e 1 and 2, to preclude the possibility of the formation of an explosive hydrogen atmosphere.
9.4.5.2.1.2 Power Generation Design Bases The ventilation systems for the ESF systems provide the same l
functions for both normal and abnormal conditions; thereby initiating the design bases for both safety and non-safety l
functions.
Consequentir, the power generation design bases are identical to the safety design bases, shown in Subsection 9.4.5.2.1.1.
9.4.5.2.2
System Description
I a.
The schematic diagram of the switchgear heat removal systems for Units 1 and 2 is shown in Figure 9.4-10.
Equipment parameters of principal. system components are listed in Table 9.4-17.
l l
b.
The heat removal systems for Units 1 and 2 circulate air continuously to maintain temperature within the approximate limits of a maximum of 104' F and minimum of 650 F.
9.4-34
LSCS-FSAR AMENDMEMT 42 FEBRUARY 1979 c.
Each system is designed to operate during a design-(
basis accident and is connected to the essenti.al bus serving the equipment in its respective area.
d.
Each heat removal system is designed to admit 100%
outside air, but vatside air dampers and return air dampers are modulated to maintain temperatures within the limits.
Any excess air is elieved,through the exhaust air riser.
e.
One exhaust fan is provided for purging each battery room of Units 1 and 2 of hydrogcn generated during battery charging.
The battery rooms are maintained at a negative pressure with respect to the switchgear room.
f.
Fire dampers with fusible links are provided in any dact penetrations and any ventilation openings in fire walls.
- g. _All switchgear heat removal system equipment is
_pnysically seoaratea Dv virtue of its location within the separate switchgear areas.
h.
The heat recovery unit consists of a high efficiency filter, heat recovery coil, and two 50% fans for each Unit 1 and 2.
The air, thus induced by heat recovery
-(
fans is either recirculated or discharged to the atmosphere.
i.
Controls and instrumentation:
1.
Each switchgear heat removal system has a control panel containing all starters, relays, and controls necessary for the operation and monitoring of its respective air handling unit.
2.
Ionization smoke detectors are located in each switchgear room to provide main control room annunciation in the eve: t of detection of combustion products-3.
Handswitches for manual starting of heat removal-units and exhaust fans are provided locally.
4.
Controls are electric or electronic.
9.4.5.2.3 Safet7 Evaluation a.
All power and control circuits for each package heat removal unit meet IEEE 279 and IEEE 308 criteria.
k 9.4-35
LSCS-FSAR AMENDMENT 42 FEBRUARY 1979 b.
All equipment and surrounding structures are designed
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for Seismic Category I, c.
The loss of any single heat removal unit does not arrect tne sare shutdown capau12Acy vt une station i
s,ince independent units are provided for each, division of switchgear.
d.
A failure analysis is presented in Table 9.4-18.
i 9.4.5.2.4 Inspection and Testing
~
All equipment is factory inspected and tested in accordance with the applicable equipment specifications, quality assurance requirements, and codes.
System ductwork and erection of equipment are inspected during various construction stages for quality assurance.
Construction tests are performed on all mechanical components and the system Js balanced for the design air flow and system operating pressure.
- Controls, l
interlocks, and safety devices on each system are cold checked, adjusted, and tested to ensure the proper sequence of operation.
The maintenance its performed on a basis generally in accordance with the equipment manuf acturer's recommendations and station practices.
Equipment and systems operation are demonstrated during normal
(_
plant operation since this system operates continuously.
9.4.5.3 ECCS Equipment Areas Cooling System This system serves the emergency core cooling system (ECCS) equipment cubicles whenever the ECCS equipment is required for service.
Each Unit 1 and 2 is provided with independent ECCS area cooling systems.
9.4.5.3.1 Design Bases This system removes equipment heat from ECCS equipment areas and i
maintains temperatures within equipment limits.
The BCCS equipment areas cooling system is designated an engineered safety feature.
9.4.5.3.1.1 Safety Design Bases a.
The ECCS equipment area cooling system for both Units 1 and 2, consisting of fan-coil units for each ECCS equipment cubicle, is designed for removing equipment heat under all normal and abnormal station operating conditions with the exception of residual heat removal (RHR) service water pump cubicles, located in the diesel building.
The cooling is provided by
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outside air.
9.4-36
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SWITCHGEAR HEAT REMOVAL SYSTEM JSMJ
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j Isolation of Essential and Non-Essential Systems (Section 22.11.E, Page 22-66) r The referenced section of the SER discusses isoletion requirements.
Although not treated explicitly in Section 22.II.E.4.2.2 questions have been raised by the NRC Staff relative to review of the reset logic associated with ESF actuation logic.
This question is discussed in FSAR Appendix L. and in tha response to Q31.285.
The reset of the ESF logic for the Feedwater Inlet Check Valves has been determined not to require modification.
The basis for this determination is provided in the draft revised response to.Q31.285 which is attached.
The purpose of this attachment is to assure an accurate understanding of the meaning of SER Section 22.II.E.4.2.(b).(2) as shown on the attached' sheet.
NRC Staff concurrence in this interpretation is required.
1 4
4 i
i 2632N
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(i) Clean condensate; service air; clean condensate to refueling bellows and reactor well bulkhead drain--all these lines are provided with locked closed valves.
(ii) Recirculation pump seal injection supply--these lines are equipped with check valves to prevent f'ow out of containment.
(iii) Combustible gas rett.rn, provided with closed valve during all modes of operation (see item 5 below).
(b)
- t. ~ ial lines.
Since these lines are vital to litigate the consequences of an accident, the basis upon which these lines are designed is found to be commensu ate with the safety importance of isolating the lines.
They include:
]
i 4
(i) Lines that are provided with remote manual isolation capability to permit isolation from the main control room if necessary, including icw pressure coolant spray, high pressure coolant spray, residual heat removal (low pressure coolant injection),
reacter core injection coolant suctions from the appression poo'; reactor core injection coolant pump minimum flow lines, low pressure coolant spray minimum flow and test lines and residual heat removal safety / relief valve discharge.
(ii) Lines that are provided with check valves to prevent reverse flow in case of a line break in addition to valves that have remote manuai isolation capability, including:
low pressure coolant injection, high pressure coolant spray, residual heat removal (low pressure coolant injection) injection lines-SGE N A'5 0 reactor core i ;jection coolant turbine exhaust lines;(Feactor t
Rouisson ro C}31.28Cfeedwater liEp and reactor core injection coolant vacuum pump 4 7FA cus o.
discharge.
(iii) Lines that are provided with an automatic isolation signal in addition to remote manual capability including:
main steam and drain lines; control rod drive insert and withdrawal; residual heat removal minimum flow and test lines; and reactor core injection coolant turbine exhaust breaker lines.
~
(iv) Other lines that have isolation provisions that ar i not de-scribed above, including the standby liquid control system (check valves); reactor core injection coolant, high pressure coolant spray, low pressure coolant spray and residual heat re n val safety relief valves discharge and vacuum breakers (process signals); automatic pneumatic supply-(is required to be open during all modes).
(3) Isolation of nonessential systems.
As discussed previously, all nonessential systems that provide a possible open path out of the primary containment were found to be I
either isolated by diverse isolation signals, by check valves that 22-66
LSCS-FSAR AMENDMENT 54 JANUARY 1981 QUESTION 031.285 "Several instances have been reported where automatic closure of the containment ventilation / purge valves would not have occurred because the sa'ety actuation signals were either manually over-riden or bypassed Iblocked) daring normal plant operations.
In addition, a related design deficiency with regard to the resetting of engineered safety feature actua-tion signals has been found at reveral operating facilities where, upon the reset of an ESF signal, certain safety re-lated equipment would return to its non-safety mode.
"Specifically, on June 25,
.978, Northeast Energy Company discovered that intermittent containment purge operations had been conducted at Millstone Unit No. 2 with the safety actua-tion signals to redundant containment purge isolation valves (48 inch butterfly valves) manually overriden and inoperable.
The isolation signals which are required te automatically close the purge valves to assure containment integrity were manually overriden to allow purging of containment with a high radiation signal present.
The manual override circuitry designed by the plant's architect / engineer defeated not only the high radiation signal but also all other isolation sig-nals to these valves.
To manually override a safety actua-(~#)
tion signal, the operator cycles the valve control switch to the closed position and then to the open position.
This action energized a relay which blocked the safety signal and allowed manual operation independent of any safety actuation signal.
This circuitry was designed to permit reopening of certain valves after an accident to allow manual operation of required safety equipment.
"On September 8, 1978, the staff was advised that, as a mat-ter of routine, Salem Unit No. 1 had been venting the con-tainment ventilation system valves to reduce pressure.
In certain instances this venting has occurred with the contain-ment high particulate radiatina monitor isolation signal to the purge and pressure-vacuum relief vlaves overridden.
The override of this containment isolation signal was accom-plished by resetting the train A and B reset buttons.
Under these ci rcumstances, six valves in the containment' vent and purge syscams could be onened with the radiation isolation signal prcrant.
This override was performed after verifying that the actual containment particulate levels were accept-able for venting.
The licensee, after further investigation of this practice, determined that the reset of the particu-late radiation monitor alarm also overrides the containment isolation signal to the purge valves such that the purge valves would not have autocatically closed on ta energency core cooling system (ECCS) safety injection signal.
gs
(/
031.285-1 L
)
LSCS-FSAR AMENDMENT 54 JANUARY 1981
()
A related design deficiency was discovered during a review of system operation following a recent unit trip and subse-quent safety injection at North Anna No. 1.
Specifically, it was found that certain equipment important to safety (for example, control room habitability system dampers) would re-turn to its non-safety mode following the reset of an ESF signal.
"In addition, many utilities do nst have safety grade radia-c tion monitors to initiate containment isolai Jn.
SAFETY SIGNIFICANCE "The overriding of certain containment ventilation isolation signals could also bypass other safety actuation signals and thun prevent valve closure when the other isolation signals are present.
Although such designs may be acceptable, and even necessary, to accomplish certain reactor functions, they are generally unacceptable where they result in the unneces-sary bypassing of safety actuation signals.
Where such by-passing is also inadvertent, a more serious situation is cre-ated especially where there is no bypass indication system to alert. the operator.
"Where the resetting ESF actuation signals, such as safety
(~')
injection, directly causes equipment important to safety to return to its non-safety mode, protective actions of the af-fected systems could be prematurely negated when the associ-ated actuation signal is reset.
Prompt operator action would be required to assure that the necessary equipment is re-turned to its emergency mode.
"The use of non-safety grade monitor to iniciate containment isolation could seriously degrade the reliability of the iso-lation system.
STAFF POSITION "It is our position that, in addition to other applicable cr'iteria, the following should be satisfied for all operating license applications currently under review:
1)
The overriding" of one type of safety actuation signal (e.g., particulate. radiation) should not cause the block-ing of any other type of safety actuation signal (e.g.,
iodinc radiation, reactor pressure) for those valves that have
.w function other than containment isolation.
l 2)
Physical features (e.g.,
key lock switches) should be provided to ensure adequate administrative controls.
O Q31.285-2 i
LSCS-ESAR AMENDMENT 54 JANUARY 1981
()
3)
A system level annunciation of the overridden status should be provided for every safety system impacted when any override is active.
(See R.G.
1.47).
4)
The following diverse signals should be provided to ini-tiate isolation of the containment purge / ventilation system:
containment high radiation, safety injection actuation, and containment high pressure (where contain-ment high pressure is not a portion of safety injection actuation).
5)
The instrumentation systems provided to initiate contain-ment purge ventilation isolatlon should be designed and qualified to Class IE criteria.
b 6)
The overriding or resetting of the ESF actuation signal should not cause any equipment to change position.
"Accordingly, you are requested to review your protection system design to determine its degree of conformance to these criteria.
You should report the results of your review to us describing any departures from the criteria and the correc-tive actions to be implemented.
Design departures for which no correcrive action is planned should be justified.
"The following definitions are given for clarity.
aOverride:
The signal is still present, and it is blocked in order to perform a function contrary to the sig-nal.
bReset:
The signal has come and gone, and the circuit is being cleared in order to return it to the normal condition."
RESPONSE
A review and reverification of the LSCS ESF systems has been conducted with the following results:
1)
For those valves which have no function other. than contain-ment isolation, the overricing of any single safety signal doec not block the receipt and appropriate response to any other type of safety actuation signal.
In addition, these valves which I
close upon receipt of isolation signals will remain closed upon reset of the isolation signal, until manual operation action is
(
taken to reopen them.
2)
Where necessary, key lock switches have been provided to ensure adequate administrative control of override capability.
, (LJ Q31.285-3
LSCS-FSAR AMENDMFl1T 54 / 7 JANUARY)d81
()
3)
A system level annunciator is provided for every safety system impacted when any override is activated.
4)
Diversity of containment purge isolation is provided by the following initiating conditions:
containment high radiation, reactor building high radia tion, reactor low water level, and containment high pressure.
5)
The instrumentation systems provided to initiate containment purge isolation are designed and qualified to c. ass lE criteria.
6)
A review of the LSCS Unit 1 schematic D'3 rams resulted in the identification of various electrical services which currently revert to their rarmal mode on reset of ESF actuation signals.
This review is in addition to that discussed in Item II.E.4.2 of Appendix L.
These equipment include:
m
/MM87 h Schematic Diagram Equip: rent Nos.
)
e 1E 6-~4432AB ~ ~ ~
TCGTCe 2
^
lE
-4432AD
/
OV 2YA, OVC05YA, pFZ-VE049X.,
OF XVE049XB
/
lE 432AG OVC05YB,, OVC05 'lB lE-0-4432AH QF Z-VC 04'9 XA,
FZ-VC049XB,
- OVCO2YB, VCO3YB Q--
lE-1-4018
- 1CM022A, CM024A, 1CM025A, j
/
ICM0' B, iCM023B, 1CMO::6B
.)
lE-10 074AB lVG00 N
f
\\
lE-4074AA \\
IVG C
lE- -4200AF Valves 1B21-F032A&B
~i r_Will closo nn ramet of Isol Sig).
The installed instrumentation and controls at LSCS will be veri-fied by test to be consistent with the schematics.
Any safety-related equipment which actually does not remain in its emergency configuration on reset of an ESF signal will be identified and a proposed system modification, design change or other corrective action will be determined or the current design justified.
Implementation of any necessary modifications will be completed prior-to fuel load subject to availability of qualified equipment.
/W32 T&
(efhd.)
o V
Q31.285-4
LSCS-fy)-g
/NSTEAT & :
lE-0-4432AA OVC03CA lE-0-4432AB OVC03CB lE-0-4432AD OVC05YA, OVC52YA lE-0-4432AE OVC08YA, OVC10YA, OVCl4YA, 0VC53YA, OVCllYA, OVCl2YA, 0VCl3YA lE-0-4432AG OVC05YB, OVC52YB lE-0-4432AH OVC08YB, 0VC10YB, 0VCllYB, OVCl2YB, OVCl3YB, OVCl4YB, 0VC53YB lE-0-4432BD OVC0lYA, OVC0lYB lE-0-4434AC OVE03YA, 0VE04YA, OVE05YA, OVE07YA, OVE08YA, 0VE09YA, OVE30YA lE-0-4434AE OVE03YB, OVE04YB, OVE05YB, OVE07YB, OVE08YB, OVE09YB, OVE30YB lE-1-4018AD 1CM022A, 1CM024A, 1CM025A, 1CM021B, 1CM023B, 1CM026B lE-1-4074AA lVG01C lE-1-4083AA lVR0lCE, IVR02CA lE-1-4083AB IVR01CB, lVR02CB 1E-1-4083AC lVR01CC, lVR02CC lE-1-4083AF lVR04YB, lVR05YA lE-1-4083AG lVR04YA, IVR05YB lE-1-4200AF 1B21-F032A, 1B21-F032B INSEKT & :
With respect to the above listed equipment, system designs have been modified in all instances but one to ensure that safety related equipment remains in its emergency configuration and does not automatically revert to its normal operating mode upon reset of ESF actuation signals.
The single case that does not require a design change is that of the Feedwater Inlet Check Valves, 1B21-F032A&B.
These valves, as noted in Table 6.2-21 Note 17, are provided with actuators for testing and ensuring leak tightness at low differential pressures.
Upon receipt of an ESF actuation signal, the valve actuator will be energized to close, howe-ever, since it is not capable of completely closing the valve against flow, the valve will remain open.
Once reverse flow is established, the valve will close and be securely seated by either the flow itself or by the actuator if the dif feren-tial pressure across the valve is low.
Once the valve is seated, the actuator performs no other function and its continued energization through the ESF actuation signal is not necessary.
The line will remain isolated, irrespective of the presence or absence of the ESF actuation signal, until process line conditions dictate otherwise.
Therefore, even though the present design allows the isolation signal to the valve actuator to clear immediately with reset of the ESF actuation signal, the valve will not automatically revert to its normal operating mode.
An such, the present system design is acceptable.
Q31285-5
- - - - -