ML20069M121
| ML20069M121 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 11/12/1982 |
| From: | Devincentis J PUBLIC SERVICE CO. OF NEW HAMPSHIRE, YANKEE ATOMIC ELECTRIC CO. |
| To: | Knighton G Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.E.3.1, TASK-2.G.1, TASK-TM SBN-367, NUDOCS 8211230069 | |
| Download: ML20069M121 (52) | |
Text
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m m su m IPUBLIC SERVICE r;z., onk.:
Companyof New Hempshire 1671 Worcester Road Fromingham, Massachusetts 01701 (617) 872 - 8100 November 12, 1982 SBN-367 TF B7.1.2 United States Nuclear Regulatory Commission Washington, D. C. 20555 Attention:
Mr. G. W. Knighton, Chief Licensing Branch 3 Division of Licensing
References:
(a) Construction Permits CPPR-135 and CPPR-136, Docket Nos. 50-443 and 50-444 (b) PSNH Letter SBN-289, dated July 2, 1982, "FSAR Section 8, Open Item List - Power Systems Branch (Electrical)
J. DeVincencis to F. Miraglia (c) USNRC Memorandum, dated October 14, 1982, " Notice of Meeting Regarding Open Items in the Safety Review",
L. L. Wheeler to J. D. Kerrigan
Subject:
Resolution of Open items - Power System Branch (Electrical)
Dear Sir:
Transmitted herewith are additional responses to certain open items listed in Reference (b) and (c). These open items were discussed with PSB representatives during a telephone conversation on 9/23/82.
Very truly yours, YANKEE ATOMIC ELECTRIC COMPANY J. DeVincentis Project Manager GTs/dsm C) 8211230069 821112 PDR ADOCK 05000443 A
y OPEN ITEM 10 Bus transfer capability.
RESPONSE
Additional information on the above subject is provided in l-the attached marked-up FSAR page, 8.3-25.
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SB 1 & 2 FSAR brings the system into operation including portions of the protection system and transfer of p0wer among various of fsite and onsite power supplies will be tested.
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3.
Criterion 5 - Sharing of Systems or Components Between Units Electrical structures, systems and components important to safety in the onsite power system are not shared between the two units, except in the instances discussed below.
This sharing will not impair their ability to perform their safety function, including, in thq event of an accident in one unit, an orderly shutdown and cooldown of the remaining unit.
($)
The seismic Category I s'ervice' water pump" house provides
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one room and raceway system for the Train A equipment of both units, and another room and s'eparate raceway system for the Train B equipment of both units (see Figure 1.2-46).
A single failure cannot af fect both trains of either unit.
(b)
Individual ducts for underground feeders of Train A of both units share the same Class 1E duct bank and manhole system for a portion of the run from the service water
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intake structure to the Unit 1 area. Again, a single failure cannot affect both trains of either unit.
(c)
Individual ducts for underground circuits of Train A of both units for the control room makeup air system share the same Class 1E duct bank and manhole system where the circuits travel through the opposite unit to the makeup air pits. The same is true for the Train B circuits. A single failure cannot af fect both trains of either unit.
(d)
Each of the two Train B service water cooling tower fans is capable of being connected to the Train B emergency bus of either unit (Figure 8.3-15).
The two breakers connecting each fan to the two buses are mechanically and electrically interlocked to ensure that only one breaker can be closed ac any time. A single failure cannot af fect both trains of either unit.
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4.
Criteria 33, 34, 35, 38, 40, 41, 44 l
t Onsite electric power interconnections and transfers are designed so that the safety functions of the reactor cool-ant makeup system, the residual heat removal system, the l
emergency core cooling system, the containment heat removal l (
system, the containment atmosphere cleanup system and the t
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During reactor operation, the capability to transfer power from the unit auxiliary transformer source to the reserve auxiliary transformer source is continuously monitored. Alarms are provided in the main Control Room to alert the operator if synchronism is lost between the switchgear and the reserve source or if control power is lost to the reserve source circuit breaker.
Transfer of power from the unit auxiliary transformer soucce to the reserve auxiliary transformer source is not periodically tested at power because such transfers may introduce unwarranted challenges to the Electric Power System that may result in a plant trip.
For a discussion of transfer of power initiated by operation of the generator breaker, see Section 8.2.1.6.
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OPEN ITEM 12a: Position 1 of BTP PSB-1
Response
In order to close out open item 12a, compliance.with Position 1 of'BTP PSB-1, we agreed to submit a list of equipment which would be available t
in the event of degraded grid voltage for the plant toLreach hot shutdown.
The response to RAI 420.38, which is in the final stages of review, will include a list of equipment required to reach hot and cold safe shutdown.
We have determined through a review of. the preliminary response and through discussion with tne cognizant. individuals, that the response will
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list equipment which will be available in the event of degraded grid voltage j
for the plant to reach hot' shutdown.
This equipment will be available in the event of degraded grid voltage because of one or more of the following reasons:
1.
Not powered by the degraded power source 2.
Does not rely on electric power 3.
In standby and therefore not connected to the degraded source 4.
Equipment will run under degraded voltage conditions because of margin between equipment rating and duty 5.
Not sensitive to degraded grid voltage (resistive load) 6.
Time is available for corrective action.
The response.to RAI 420.38 will be submitted before January _1, 1983.
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CPEN ITEMS 13b, Complicac3 with IEEE standards.
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and 32 k.
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RESPONSE
See revised response to RAI 430.3 (attached).
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430.3 Section 8.1.5.2 of the FSAR indicates that the Seabrook design is
'4 (8.1) in conformance with IEEE Standard 387-1972. Section 8.1.5.3 of
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the FSAR, on the other hand, implies that the Seabrook design is
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also in conformance'with 387-1977 by reference to Regulatory Guide 1.9 (Revision 1).
Similar inconsistencies exist between g
Sections 8.1.5.2'and 8.1.5.3 of the FSAR.for IEEE Standard 308-1971 and Regulatory Guide 1.32.(Revision 2), IEEE Standard 317-1972 and Regulatory Guide 1.'63:(Revision 2), IEEE Standard 1
384-1974 and Regulatory Guide 1.75 (Revision 2), IEEE Standard 333-1975 and Regulatory Guide 1.118 (Revision 2), IEEE Standard 484-1975 and Regulatory Guide 1.128 (Revision 1), and IEEE 1
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Standard 450-1972 and Regulatory Guide 1.129 (Revision 1).
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f Correct the irkonsistencies and describe and justify each l T exception taken to IEEE Scandards 308-1974, 387-1977, 317-1976, b ;s 384-1974, 318-1977, 484-1975 and 450-1975.
A RESPON$h, The basic commithnts to various industry standards pertinent to j
E Jehe electrical' design of the Seabrook project have been established at the PSAR stage. We have and plan to adhere to
' these commitments as shown in Section 8.1.5.2.
It should be pointed out that SEEE Scandard 338-1975 and IEEE 379-1972 are additions to the.FSAR listed standards as are the upgrading of IEEE Standard 344-1971 to 344-1975.
Over and above'rhese commitments, it was decided at the FSAR stage to evaluate the extent of compliance to the most recent Regulatory i
Guides a'nd Branch Technical Positions at the date of issuance of
.the FSAR. This evaluation is reflected in Sections 8.1.5.3 and 8.5.1.4 and in'no way should be construed to reflect a commitment to the Regulatory Guide. We are aware that the listed Regulatory i
Guides of ten reference a later issue of an industry standard; however, Regulatory Guides such as 1.30, 1.40, 1.41, 1.53, etc.,
continue to reference superseded issues of industry standards in spite of the publication of revised standards. This unfortunate lack of coordination will always exist unless simultaneous revisions are made by industry and the NRC.
It is obvious that the problem is not easily' resolvable, and for this reason, we have established the ground rules reflected in the discussion above.
l The evaluation of the Seabrook design against the most recent Regulatory Guides provides an indication of how Seabrook design complies with recent NRC requirements.
In addition, an evaluation of the Seabrook design has been performed to the standards listed below having more recent issue dates than those listed in the_FSAi. The purpose of the evaluation was to determine if-there are any requirements of safety significance that the Seabrook design does not meet and which are included in the standards. 'The results of the j
evaluation are outlined below:
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a IEEE 308-1974: It is cur enginacring judgment that tha S2sbrook design meets the requirements of this standard. On test intervals for batteries and diesel generators, cur design (Technical Specifications) exceeds the requirements of this standard because the intervals have been specified to meet even more recent industry standards such as IEEE-387 on diesel generators and IEEE-450 on batteries.
IEEE 387-1977: The basic difference between the 1977 version of the standards and the 1972 version is the incorporction of type testing requirements 'for diesel generators.
Because the FSAR commits to the type test program of IEEE 387-1977, it is our engineering judgment that the Seabrook design meets this standard.
IEEE 317-1976: All major electrical containment penetrations were manufactured to meet the 1976 version of the standard.
Some minor electrical penetrations, 3/4" to 1" size which are associated with the personnel air lock, the equipment hatch and the containment recirculation sump isolation valve encapsulation tank, were manufactured to meet the 1972 version.
It is our engineering judgment that although these minor penetrations might not meet all the most recent requirements, they will perform their safety function.
IEEE 384-1974: The Seabrook design meets the requirements of this standard.
IEEE 338-1977: The Seabrook design meets the requirements of this standard.
IEEE 484-1975: Refer to RAI 430.31 for comments on this standard.
IEEE 450-1975: The Seabrook design meets the requirements of this standard.
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OPEN ITEM 14 Non-safety. leads powered from the Class lE ac distribution system
RESPONSE
The response to this item was reflected in attachment L of our: letter to the NRC, SBN-289, dated July 2, 1982.
Apparently the reviewer did not receive this attachment.
.Another copy of'this attachment is transmitted herewith.
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430.44 Redundant Buses E5 and E6 are interconnected through Bus ES2, (8.3.1)
MCC 523, Charger ED-13C-2B,125 V Bus 128, UPS ED-I-2B and Bus E63.
(8.3.2)
Either eliminate or justify this interconnection.
Provide the results of an analysis that identifies and justifies all electrical interconnections between redundant divisions.
RESPONSE
The interconnection decribed above is imaginary and from a (11/82) practical standpoint is nonexistent. For this interconnection to exist, multiple failures of diverse equipment would have to occur first,3 followedbyatransformationfrog,a3-Wire {hystemtoa 2-Wire /hystem and back again to a 3-WireA ystem. The safety S
divisions which are purported to be interconnected are in addition effectively isolated through the inherent blocking feature between input and output of the chargers, the current limiting feature of the UPS and the charger and the various protective. devices. If this imaginary interconnection is to be given any credence, then so must the equally imaginary interconnection that could exist between redundant divisions through the common ground path and ground grid at the station.
We have performed an analysis to identify all other electrical interconnections between redundant divisions and it remains our position that there are no electrical interconnections between -
redundant divisions.
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OPEN ITEM 18 Battsry cupporto - interfcca b2twe n bitt;ry rack 3 and battery cells.
RESPONSE
See revised response of RAI 430.33 (attached).
Rev. Nov. 1982 430.33 Recent operating experience has shown that an incompatibility between the battery rack and the tattery may cause cracking of the battery case. The cracking may be caused in part by the improper support at the battery stress points. Describe the battery stress points and their relationship with battery rack supports.
RESPONSE
The Seabrook batteries utilize cells that have a plate support bridge that is molded separately from the cell jar. This permits even distribution of element weight across the entire bottom surface of the jar. This design results in a relatively stress-free plastic molding with the weight of the elements equally distributed across the container bottom.
In the racks, the cells, which have a base of 14-9/16 x 14-1/2 inches, sit on three 1.63-inch wide steel stringers located under the cell center line and + 5 inches from the center. This evenly distributes the cell weight to minimize stress on the cell.
Concerns regarding the location of the rack steel stringers in relationship to the battery plate support bridges are addressed in the attached letter from the battery manufacturer.
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Gould inc Industrial Battery Division e * * '2050 Cabot Boulevard Wxt. L ngnorns, P2.19047 GOULa
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CABLE: GOULNATBAT LANGHORN E, PA.
Sectrottics & Electrical Products TWX: GOULD LAHN. 510-667 2056 GOULD LANG. 51(M67 2066 August 13, 1982 United Engineers & Constructors Inc.
30 South 17th Street P.O. Box 8223 Philarialphia, PA 19101 Atta:
G.M. Aggarwal Supervising Electrical Engineer
Subject:
Public Service Company of New Hampshire Seabrook Station - Units 1 & 2 Storage Batteries 9763-006-137-1 SBU-58840
Dear Mr. Aggarwal:
In response to the subject regarding NRC's concern on container stress between the container and battery, we wish to respond as follows:
(1) Both the cell design and associated rack design were-governed by Gould and represent the current state-of-the art with respect to element and/or cell support.
1 (2) The battery and rack interface was tested as designed through our seismic qualification program; ref: Wylie Test Report No. 44681-1 of 10-27-81.
(3) The seismic loads experienced during this test were l
quite significant as indicated by the required and test response spectra. The batteries and racks 1
used at Seabrook Unit 1 were typical of the cells
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and racks tested.
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(4) Thc distribution of stresses at the bridge / container bottum as well as those at the cell bottom and rack rail interface have been successfully accomodated as evidence by the tests aforementioned.
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Mr. G.M. Aggarwal August 13, 1982 Page 2 (5) The Gould design feature of the separated bridge concept has done away with the potential stress problem that existed prior to our M & N line design initiated in 1970.
We trust the above satisfies those concerns referenced in your letter of July 21, 1982.
9 OTBGR S 0%7 6 C-T t4eT RCL ATED To (mae autreq cow ceau s j
Q ry truly yours, (RaynondR. Fletcher
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Manager, Marketing Utilities RRF:dh cc: Bob Bevan - Brown Boveri Electric Ramesh Desai - Gould
.B. Hopewell - Gould G Morris p U.E. & C.
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3 OPEN ITEM 20 Tha NRC rsquratsd that the load profilco for the safety-related batteries be included in the FSAR.
RESPONSE
The FSAR has been revised accordingly to incorporate these profiles.
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OPEN ITEM 22 Submerged electrical equipment as a result of a LOCA.
RESPONSE
For additional information on the above item see attached revised RAI 430.62.
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t Revision November 1982
~ 430.62 Identify all electrical equipment, both safety and non-safety, that may become submerged as a result of a LOCA. For all such equipment that is not. qualified for service in such an environment provide an analysis to determine the following:
1.
The safety significance of the failure of this electrical equipment (e.g. spurious actuation or loss of actuation function) as a result of flooding.
2.
The effects on Class lE electrical power sources serving this
- equipment as a result of such submergence; and 3.
Any proposed design changes resulting from this analysis.
- RESPONSE:
All electrical equipment that may become subme.,ed as a result of a LOCA is listed in Table 430.62-1.
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TABLE 430.62-1 EQUIPMENT OR COMPONENT OF EQUIPMENT LOCATED IN THE CONTAINMENT BELOW THE FLOOD LEVEL OF (-) 20'-8" VALVE LIST V-VITAL EQUIPMENT OR TAG NV - NON-VITAL COMPONENT SUEMERGED APPLICATION CS-V-59 NV Solenoid & Limit Switch RCP LD Seal Water Return CS-V-145 NV Solenoid & Limit Switch Letdown HK-E-2 to HX-E8 CS-V-170 NV Solenoid & Limit Switch Letdown HX-E-3 to RCDT C3-V-175 NV Solenoid & Limit Switch Excess Letdown Line CS-V-176 NV Limit Switch Excess Letdown Line CS-V-177 NV Limit Switch HX-E2 to Cold Leg 4 CS-V-180 NV Solenoid & Limit Switch HX-E2 to Cold Leg 1 CS-V-185 NV Solenoid & Limit Switch HX-E2 to Pressurizer CS-V-168 V
Motor Operator RCP Seal Water Isolation 1
NG-V-17 NV Solenoid & Limit Switch Accumulator 9A Ni Line NG-V-19 NV Limit Switch Accumulator 9B Ni Line NG-V-21 NV Solenoid & Limit Switch Accumulator 9C Ni Line NG-V-23 NV Limit Switch Accumulator 9D Ni Line RC-LCV-459 NV Solenoid & Limit Switch Letdown Isolation Valve RC-LCV-460 NV Solenoid & Limit Switch Letdown Isolation Valve RC-LCV-81 NV Solenoid & Limit Switch RC Loop 3 Letdown to HX-E2 RMW-V-28 NV Solenoid RMW-TK 12 to RC TK 11 RMW-V-180 NV Solenoid RC-P-1B Seal Pressurizer Equalizing Valve RMW-V-181 NV Solenoid RC-P-1A Seal Pressurizer Equalizing Valve RH-V-27 V
Solenoid & Limit Switch HX-E-9B Header Test RH-V-28 V
Solenoid & Limit Switch HX-E-9A Header Test RH-V-49 V
Solenoid & Limit Switch HX-E-9A Injection Test RH-V-54 NV Solenoid & Limit Switch SI-P-6A Discharge Test RH-V-55 NV Limit Switch SI-P-6B Discharge Test SI-V-03 V
Stem Mounted Limit Switch Accumulator Iso. Valve Stem Limit Switch SI-V-04 NV Limit Switch Accum. Test Valve SI-V-15 NV Solenoid & Limit Switch Accum. Fill Valve
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SI-V-17 V
Stem Mounted Limit Switch Accum. Iso. Valve Stem Limit Switch SI-V-18 NV Solenoid & Limit Switch Accum. Test Valve SI-V-23 NV Solenoid & Limit Switch Accum. Fill Valve SI-V-32 V
Stem Mounted Limit Switch Accum. Iso. Valve Stem Limit Switch
TABLE 430.62.1 (Continued)
VALVE LIST.
V-VITAL EQUIPMENT OR TAG NV - NON-VITAL COMPONENT SUBMERGED APPLICATION SI-V-33 NV Solenoid & Limit Switch Accum. Test Valve SI-V-38 NV Limit Switch-Accum. Fill Valve SI-V-47 V
Stem Mounted Limit Switch Accum. Iso Valve - Stem Limit Switch SI-V-48 NV Solenoid & Limit Switch Accum Test Valve SI-V-53 NV Solenoid & Limit Switch Accum Fill Valve SI-V-131 V
Limit Switch SI - Cold Leg Test SI-V-132 NV Limit Switch SI - Hot Leg 3 Test SI-V-133 NV Limit Switch SI - Hot Leg 2 Test SI-V-134 V
Solenoid & Limit Switch SI - Hot Leg Test SI-V-158 V
Solenoid Charging Pump Test.
SI-V-160 V
Limit Switch SI Pump - Test Line Iso. Valve WLD-FV-1403 NV Solenoid & Limit Switch RCDT Transfer Valve T
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INSTRUMENTATION LIST SAFETY-RELATED INSTRUMENTS THAT MAY BECOME SUBMERCED TAG DESCRIPTION ACTION NI-NE-41 A/B Power Range Neutron Detectors Not required post-LOCA NI-NE-42 A/B Power Range Neutron Detectors Not required post-LOCA NI-NE-43 A/B Power Range Neutron Detectors Not required post-LOCA i
NI-NE-44 A/B Power Range Neutron Detectors Not required post-LOCA RM-RM-6535A-Manipulator Crane Radiation Monitor Not required post-LOCA RM-RM-6535B Manipulator Crane Radiation Monitor Not required post-LOCA RC-FT-414, 415, 416 RC System Loop 1 - Flow Will be raised above the flood level 424, 425, 426 RC System Loop 2 - Flow Will be raised above the flood level 434, 435, 436 RC System Loop 3 - Flow Will be raised above the flood level 444, 445, 446 RC stem Loop 4 - Flow Will be raised above the flood level a
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-TABLE 430.62-1 (Continued)
NON-VITAL INSTRUMENTS THAT MAY BECOME SUBMERGED TAG DESCRIPTION CAH-TE-5640'- 5647 NI Detector Wall Temp.
CAS-AE-8815 Hydrogen Analyzers in RC TK 55 Area CAS-AE-8816 Hydrogen Analyzers in RC CS-E-3 Area CAS-AE-8817 Hydrogen Analyzers in Cntant. Valve Rm.
CAS-AE-8818 Hydrogen Analyzers in CS-E-2 Area COP-PT-1787 CS-PT-124 Excess Letdown HX, CS-E-3 Outlet Pres.
CS-TE-126 Regenerative HX, CS-E-2 Charging Line Temp.
CS-FT-154 - 157 RCP Low Range Leakage Flow CS-FIS-191 - 194 RCP Seal Flow LD-LT-8333 Sump B Level RC-LT-9405 Press. Relief Tank Level RM-RX-6578 - 6581 Radiation Monitor SF-LT-2629 Refueling Canal Level SM-XS, XT-6701 Seismic Monitor SM-XS, ST-6709 Seismic Monitor WLD-LT-1403 RCDT, TK 55 - Level WLD-TE-1403 RCDT,-TK 55 - Temp.
WLD-FT-1406 RCDT, TK 55 - Flow WLD-FT-1411 RCDT, EX E-43 Inlet Temp.
WLD-TE-1413 RCDT, EX E-43 Outlet Temp.
WLD-PT-1412 RCDT, EX E-43 Pump P33 A/B Discharge Pres.
WLD-PT-1420 RCDT, EX E-43 Pump P33 A/B Section Pres.
WLD-LSH-6266 Containment Drains Sump A Level WLD-LSH-6267 Containment Drains Sump B Level t
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TABLE 430.62-1 (Continued)
. MISC. NON-VITAL EQUIPMENT THAT MAY BECOME SUBMERGED PUMPS TAG DESCRIPTION RC-P-271 Pressurizer Relief Tank, TK ll, Recire. Pump
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SF-P-272 Refueling Canal Drain Pump WLD-P-5A, 58 Containment Sump A Pumps WLD-P-SC, SD Containment Sump B Pumps WLD-P-33A, 33B RCDT, TK 55, Pumps 4
INSTRUMENT RACKS Individual instruments located on these racks have been identified above.
CONTROL PANELS TAG DESCRIPTION 1
WLD-CP-280 Containment Sump A - Control Panel 1
WLD-CP-281 Containment Sump B - Control Panel i
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TAG DESCRIPTION ED-X-16F Transformer Supply for Panel Ll7 ED-X-16A Transformer Supply for Panels PP-8B and EL 13 ED-X-16J Transformer Supply for Panel L41 Panel L41 Lighting Panel TERMINAL BOXES TAG DESCRIPTION X45 Pressurizer Heater Backup Group C X46 Pressurizer Heater Backup Group D
1.
The followieg analysis discusses the safety significance of the failure as a result of flooding of the electrical equipment listed in Table 430.62-1.
A.
Valves - Safety-Related (1) Stem-mounted Limit Switches for SI Accumulator Valves Stem-mounted limit switches only provide an alternate valve position indication. Failure of these switches could cause loss of the alternate valve position indication circuits but would not affect valve operationj valve position indication will still be provided by the normal valve limit switches.
(2) RCP Seal Water Isolation Valve, CS-V-168 This motor-operated valve is driven closed upon a containment isolation signal, therefore, this valve would fail in its safe position. Submergence will not change the fail-safe position. Motor-operated l
valves are powered from individual circuits of a motor control center so failure of this circuit would not affect the remaining loads on the motor control center.
Failure of the limit switch internal to the operator could effect the valve position indication at the control switch on the MCB and also the post-accident monitor (PAM) light indication of this valve.
This circuit will be modified by adding an interposing relay to the circuit of CS-V-168. The relay will be located in the control building thus 1
not affected by flood and will prevent loss of the i
remaining PAM monitor circuits upon flooding of the valve.
(3) SI Pump Cold and Hot Leg Test Line Isolation Valves SI-V-131, and SI-V-160 and Charging Pump Test Line Isolation Valve, SI-V-158 These Train B air-operated valves are closed on a containment isolation signal and submergence will not alter their position.
Loss of power on this circuit will also cause loss l
of power to two other safety-related valves, SI-V-174 and SI-V-70, which are powered f rom the same circuit. These valves which are not subject
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to submergence are already closed on safety injection or containment isolation signals, therefore loss of power results only in loss of valve position indication at the MCB control
r switches. The limit switches for valves SI-V-131, V-158 and V-160 are also used in the Train B PAM monitor light circuits, and this circuit.may also be lost.
These circuits will be modified by adding interposing relays to the circuits of SI-V-131, SI-V-158 and SI-V-160.
The relays will be located in the control building thus not affected by flood l
and will prevent loss of the remaining Train B PAM monitor circuits upon flooding of the above valves.
(4) SI Hot Leg Test Line Isolation Valves, SI-V-134 This valve is a normally closed test valve that also receives a containment isolation signal to ensure that it is closed. Valve SI-V-134 is a Train A valve, and loss of power on this circuit, because of submergence, will cause loss of power to four other safety-related valves which are not subject to submergence, SI-V-165 and V-173 (already l
closed on safety injection) and SI-V-62 and V-157 (already closed on containment isolation). Valve position indication at their respective MCB control switches will be lost.
Limit switch contacts for SI-V-134 are also used in the PAM monitor light circuits. An interposing relay will be added to SI-V-134 circuit to prevent loss of the remaining PAM monitor circuits upon-flooding of SI-V-134.
(5) RHR Test Valves RH-V-27 and RH-V-49 These valves are normally closed Train A test valves that receive a containment isolation sig nal. The open contact of the containment isolation signal isolates the solenoids.
If the containment isolation signal is reset, the circuit for RH-V-27 and V-49 could fail resulting in loss of power. This circuit failure would also cause loss of power to safety-related valve RH-V-16, and solenoids for FY-618-1 and HCV-606. Va17e RH-V-16 is already closed on containment isolation. Loss of power to FY 618-1 and HCV-606 solenoids will result in full flow of the RH System A loop through the RHR heat exchanger E-9A and the closing of the bypass line. This scenario is addressed in response to RAI 440.133. Valve position indication for these valves at the MCB control switches RH-V-27 and V-49 will be lost.
l Limit switch contacts for RH-V-27 and RH-V-49 are also used in PAM monitor light circuits.
Interposing. relays will be ' added to these valve circuits to prevent loss of the remaining PAM monitor circuits upon flooding of the above valves.
(6).RHR Test Valve RH-V-28 Valve RH-V-28 is a normally closed Train B test valve that also receives a containment isolation signal. The open contact of the containment isolation signal isolates the solenoid. This valve is similar to RH-V-27 (above).and failure of this circuit would result in loss of power to safety-related valve RH-V-17.and solenoids for FY-619-1 l
and HCV-607. This will result in full flow through l
RHR heat exchanger E-9B.
Valve position indication for these valves at the MCB control switches will be lost.
Limit switch contacts for RH-V-28 are also used in-PAM monitor light circuits. An interposing relay will be added to this valve circuit to prevent loss of the remaining PAM monitor circuits upon flooding of the above valve.
B.
Valves - Non-Safety Non-safety-related, air-operated valves that may be submerged following LOCA, are listed in Table 430.62-1.
Failure of the stem-mounted limit switch or pilot solenoid may cause the entire valve circuit to lose power with the results that any other-non-safety-related valves on that particular circuit may de-energize to their fail safe position. Valve position indication will also be lost.
C.
Instrumentation i
All safety-related instrumentation, with the exception of the excore neutron detectors-and maniulator crane radiation monitors is located or will be relocated above the flood level. The excore neutron detectors and manipulator crane radiation monitors are not required following a LOCA.
Non-safety-related instrumentation located below the i
flood level may f ail.
None of this instrumentation is required following a LOCA.
4
.D.
Miscellaneous Non-Safety-Related Equipment (1) ' Pumps t
Those pump motors that may.become submerged and-fail are either protected by redundant Class lE breakers or are de-energized during normal plant operation. Failure of the particular circuirlwill not affect other circuits. These non-safety-related motors are not required following a LOCA.
(2) Control Panels l
The sump pump control panels are not required following a LOCA.
(3) Lighting i
The lighting system inside containment ils normally off. Control of the system is at control stations a
located outside the personal air lock. Failure of lighting equipment due to flooding will not affect other circuits.
i (4) Pressurizer Heater Terminal Boxes Flooding of the terminal boxes for pressurizer heater backup groups C and D, will cause de-energization of these heaters. Pressurizer heater backup groups A and B which are part of the i
NUREG-0737 requirements, will not be affected-l '
because their terminal boxes are above the flood level. Backup groups A and B are fed from the j
diesel generators.
The above analysis demonstrates that there is no safety
{
significance from the failure of the equipment described l
above as a result of flooding.
In areas where safety i-concerns were identified, the modifications mentioned above will be implemented to alleviate these concerns.
l 2.
A.
There is no detrimental effect on the Class lE electric.
power sources as a result of equipment submergence.
All submerged equipment can be divided into the following three categories:
Category 1: Class 1E load supplied by a Class lE power supply.
Category 2: Non-Class lE load supplied by a Class lE power supply.
Category 3: Non-Class lE load supplied by a non-Class lE power supply.
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For Category.1 and 2 equipment, Class lE power supply is
-protected by the Class lE protective device which feeds the circuit of the submerged equipment. Only the circuit which feeds the submerged equipment is-de-energized. For Category 3 equipment there is no connection to the Class lE power supply; therefore, there is no effect due to ribmergence.
B.
Instrumentation Safety-related and non-safety-related instruments are powered through separate instrumentation panels. These panels are powered from separate distribution circuits.
Protective devices of the internal power supplies further isolate the individual circuits from the distribution system. Therefore, the Class lE pcwer sources will not be affected by submergence of any instrumentation.
C.
Miscellaneous (1) Pumps The non-vital pumps that may be submerged are powered from the non-Class lE power system.
Failure of these motor circuits will not affect the Class lE sources.
(2) Control Panels The control panels for the containment sump pumps are powered from non-Class lE sources and their
^
failure will not affect the Class lE system.
(3) Lighting System The lighting system inside the containment is normally de-energized during plant operation; therefore, the Class lE power sources will not be affected.
(4) Pressurizer Heater Terminal Boxes Backup heater groups C and D are not powered from the Class 1E power system and their failure will not affect the Class 1E system.
3.
Proposed Design Changes A.
Safety-Related Valves As discussed in Item 1, interposing relays will be added to certain valve circuits to electrically isolate the valve PAM monitor light circuit from the "alve limit switch circuit. This modification prevents submergence
4 of the valve limit switch from tripping the entire PAM monitor light circuit. No additional changes are proposed.
B.
Non-Safety-Related Valves No design changes are proposed.
C.
Instrumentation Safety-related reactor coolant flow transmitters will be raised above the flood level.
No additional design changes are proposed.
D.
Miscellaneous Equipment
+
No design changes are proposed.
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OPEN ITEM 27 Complicaca with tha guidalinne of NUREG-0737.
RESPONSE
The attached revised RAI 430.53 provides the requested information.
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11/82 Revisicn 430.53 Describe how the Seabrook design complies with the guidelines of NUREG-0737, Items II.F.3.1 and II.G.I.
i
RESPONSE
Item II.E.3.1 The Seabrook design complies with the guidelines of NUREG-0737 and the " clarification" to NUREG-0737. A description of the pressurizer heaters is provided in FSAR Section 5.4.
One pressurizer heater bank can be supplied from the Train A diesel generator and one bank can be supplied from the Train B diesel generator during loss of off-site power. Each bank can establish and maintain natural circulation at hot standby conditions. Each bank can be supplied from either of f-site power or from one diesel generator.
As demonstrated in FSAR Table 8.3-1, the standby power supply has the capacity to supply the pressurizer heaters without load shedding.
Changeover of the pressurizer heaters from normal off-site power to emergency on-site power can be accomplished manually in the I
i Control Room.
Motive and control power connections to the Class lE buses for the pressurizer heaters are through Class 1E devices.
Because of our design (see RAI 430.149), the pressurizer heaters are not automatically shed from the emergency buses upon the occurrence of a safety injection actuation signal.
Item II.G.1 The decign complies with the guidelines of NUREG-07'
' the
" clarifications" to NUREG-0737.
Motive and control components of the PORVs and the PORV block valves can be supplied from the off-site power source or the on-site power source.
f Motive and control power connections to the Class 1E buses for the PORVs and block valves are through Class lE devices.
The pressurizer level indication instrument channels are powered from the vital instrument buses. The vital buses can be powered from the of f-site power sources or on-site power sources.
The design of the PORV block valves provides the capability to close the valves and retains the capability to open the valves.
The motive and control power for the block valves is from a different emergency bus from the source supplying the PORVs.
3
Chnngaevar of power to tha PORV cnd bleck valvac frco off-site power to on-site power can be accomplished from the Control Room.
Additional information regarding compliance with NUREG-0737 has been provided in our letter SBN-212 to the NRC dated February 12, 1982, to the attention of:
Mr. F. Miraglia, Chief, Licensing Branch #3.
- -. - - - - -. ~ - -
n.
.OPEN ITEM'29' EPS'Tes' ting - Start of the cooling tower pump during the diesel-generator load sequence testlng.
RESPONSE
The attached marked-up FSAR page 8.3-18 addresses the above Concern.
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1:
A SB 1 & 2 Amendment 47 FSAR September 1982 emergency bus parameters such as voltage and incoming line circuit breaker positions (open or closed). Depending upon whether an accident condition (SI) is also present, the EPS provides appropriate contact outputs to the various safety related loads to start them in a programmed time sequence.
Momentary signals are provided to circuit breakers and starters of loads which are required to start at a specific time
(" definite start loads" see Tables 8.3-1 and 8.3-2, Sheet 1 of 2); maintained permissive contacts are provided for loads whose starting is also dependent upon the presence of a pro-cess signal (" indefinite start loads," see Thbles 8.3-1 and 8.3-2, Sheet 2 of 2).
Indicating lights for the sequencing steps are provided on the main control board to assist operation. Loading is started when the diesel generator reaches rated speed and voltage and the generator circuit breaker closes (approximately 10 seconds af ter the diesel start signal).
Table 8.3-1 shows the order and time at which the loads are automatically and sequentially applied to the diesel generator during a combined loss of of fsite power and accident condition.
Table 8.3-2 shows the order and time at which the loads are
('
automatically and sequentially applied to the diesel generator during a loss of offsite power.
Whenever a tower actuation (TA) signal is received:
the cool-ing tower pumps receive an automatic start signal and, the I
service water pumps are automatically tripped and locked out.
As noted on Tables 8.3-1 and 8.3-2, either the cooling tower pump or the service water pump, but not"both, will be loaded on the diesel generator. Upon loss of off-site power, all service water pumps and cooling tower pumps receive a trip signal. At sequence interval 37 seconds (step 5), both the cooling tower pump and the service water pump receive a start permissive from the EPS.
If a TA signal is also present, the cooling tower pump will start, otherwise the service water pump will start.
l l
After step 5, automatic starting of the cooling tower pump is blocked by the EPS until step 8 (52 seconds), such that, if a TA signal is received between 37 to 52 seconds the cooling tower pump will start at 52 seconds. The cooling tower pump will start immediately if a TA signal is received any time after sequence interval time 52 seconds.
l The diesel generator has been tested and/or analyzed to demon-l strate its ability to successfully start a load larger than the l
41 1
8.3-18
During the diesel generator load sequence testing, which is performed at least every 18 months, the design accident load will be tested by sinw.lating a loss of off-site power with a
-safety injection signal and a tower actuation signal.
In this way, the test will simulate the load of Table 8.3-1 as close as practical.
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OPEN ITEM'30- Additional information on bypass and inoperable status indication.
RESPONSE
The attached marked-up FSAR pages,'8.1-6 and 8.1-5, provide additional.information on the above subject.
i, 5
i
,,96'lI By ed a no le us In tio or 3) clea ower ant S ty Sys s"
" Application of the Single-Failure Criterion to (6/73)
Nuclear Power Plant Protection Systems" -- Refer to Subsection-7.1.2 RG 1.62
'Hanual Initiation of Protection Actions" -- Refer to (10/73)
Subsection 7.3.2 RG 1.68 "Preoperational and Initial Startup Test Program for (Rev 2)
Water-Cooled Power Reactors" -- Refer to Subsection 14.2.6 RG 1.73 ~-
" Qualification Tests'-of Electric Valve Operators -
~~
(1/74)
Installed Inside the Containment of Nuclear Power Plants" -- Refer to Section 3.11.
" Shared Emergency and Shutdown Electric Syst' ems for (Rev 1)
Multi-Unit Nuclear Power Plants."
" Qualification of Class IE Equipment for Nuclear (11/74)
Power Plants" - Refer to Section 3.11 RG 1.93
" Availability of Electric Power Sources" -- Refer Cc (12/74) to Subsection 16.3/4.8 RG 1.100
" Seismic Qualification of Electric Equipment for (Rev 1)
Nuclear Power Plants" -- Refer to Section 3.10 RG 1.106
" Thermal Overload Protection for Electric Motors on (Rev 1)
Motor Operated Valves" -- Refer to Subsection 8.3.1.1 RG 1.118
" Periodic Testing of Electric Power and Protection (Rev 2)
Systems" RG 1.128
" Installation Design and Installation of Large Lead (Rev 1)
Storage Batteries for Nuclear Power Plants" RG 1.129
" Maintenance, Testing and Replacement of Large Lead j.
(Rev 1)
Acid Storage Batteries for Nuclear Power Plants"
- Re fer to Subsection 8.3.2 l
" Qualification Tests of Electric Cables, Field (8/77)
Splices, and Connections for Light-Water Cooled l
Nuclear Power Plants" - Refer to Subsection 8.3.1.4 b.
The design of the electric power system is in accordance with the following regulatory guides, with clarifications as noted:
I l.
8.1-5 i
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SB 1 & 2 Amendment 45 FSAR June 1982 RG 1.9
" Selection of Diesel Generator Set Capacity for
('
(Rev_2)
Standby Power Supplies" Position C.4 requires that frequency should be restored to d'
within 2% of nominal in less than 60% of each' load sequence time interval.
For diesel generators, frequency can only recover when load accelerstion is completed. Load acceleration may, on rare occasions, exceed 60% of the load sequence time interval, therefore, frequency recovery in this time period is not possible. This fact presents no problem. The ef fect of extended load acceleration times has been considered in the diesel generator loading calculation. Voltage and frequency profiles, developed from the loading calculations and actual load test data, confirm that capability for successful load sequencing is not compromised by extended load acceleration times. Re~fer to Subsection 8.3.l.
-~ -
Position C.14 requires that the engine run at, full load for 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br /> following 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> at short. time rated load. For Seabrook, a " Load Capability Qualification" test was performed per IEEE 387-1977. The engine was run at full load for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after reaching equilibrium temperature, but before 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> short time rated load test.
" Periodic Testing of Protection System Actuation i'
(Rev 0)
Functions" 45
(
The design is in accordance with RG 1.22 as supplemented by Regulatory Guide 1.108, Rev. 1, entitled " Periodic Testing of Diesel Generator Units Used as Oneite Electric Power Systems at Nuclear Power Plants".
Refer to Subsection 8.3.1 RG 1.32
" Criteria for Safety-Related Electric Power Systems l
(Rev 2) for Nuclear Power Plants" Physical Independence of electric systems is in acco'r-dance with Attachment "C" of AEC letter dated December 14, 1973, entitled " Physical Independence of Electric Systems" (Appendix 8A). Refer to Subsection 8.3.1 RG 1.63
" Electric Penetration Assemblies in Containment (Rev 2)
Structures for Water-Cooled Nuclear Power Plants" The electrical penetration assemblies are designed to withstand, without loss of mechanical integrity, the maximum fault current vs. time conditions that could occur as a result of single random failures of circuit overload devices.
In addition to the 15 kV switchgear breakers, the medium voltage 15 kV penetrations are also protected by fuses inserted in the feeders outside containment. These fuses are (s
8.1-6
" Bypassed and Inoperable Status Indication for (5/73)
Nuclear Power Plant Safety Systems" With the exception of the Emergency Diesel Generator System, the Electric Power System is not required to have inoperable status indication because it is not expected to be bypassed or deliberately induced inoperable more frequently than once per year.
Reference regulatory position C.3(b).
Inoperable status indication is provided for the Emergency Diesel Generator System as a result of data provided by the NRC indicating that Diesel Generator Systems have been declared inoperable more frequently than once per year.
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.t OPEN ITEM 33 compliance with Regulatory Guide 1.63.
4
RESPONSE
Our submittal SBN-322 dated September 9,1982, addressed the above subject; additional.information is provided for the following items:
a)
Penetration protection for the reactor coolant pumps motor feeders.
b)
Testing of protective devices used to satisfy Regulatory Guide 1.63.
c)
Penetration protection for 460 volt distribution panels located inside containment.
d)
Use of thermal overload relays in penetration protection.
e)
Capability of penetrations to handle long duration overcurrent.
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a) - Penetration protection for the reactor coolant pumps motor feeders.
9 Additional.information on the above subject is provided in the FSAR.
See-attached marked up page, 8.3-3.
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4 SB 1 & 2 Amendment 45 FSAR June 1982 I-pump motors. These fuses are located in a seismic Category I building and are part of the protection for the containment electrical penetrations as required by Regulatory Guide 1.63.
In addition, a measure of backup protection is provided by the reactor coolant pump circuit breaker and the 13.8 kV bus incoming line circuit breaker.
DC control power from
[(
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separate battery sources is provided for these breakers to preclude the loss of a single de source from preventing the tripping of both the RCP and the incoming line breaker.$ For the penetration protection coordination curve, see Fig're u
8.3-47.
- 5
- b.. 4160 Volt Distribution System l.
Arrangement The 4160 volt distribution system for each unit is shown in Figures 8.3-1, 8.3-7, 8.3-8, and 8.3-9.
For each unit, the system consists of four buses, two of which are the. redundant Class lE emergency buses supplying the redundant engineered safety features loads. These safety 1(ads are divided into two separatc and independent Trains A and B, as shown on Figures 8.3-8 and 8.3-9.
The preferred power supply to each 4160 volt bus is from a UAT.
An alternate source is available to each bus through a RAT. A standby power supply, consisting
(,'
of a diesel generator, is available to each emergency bus.
Buses E5 and E6 are the equipment designations of the redun-dant Class lE buses.
Redundant Class IE Buses E5 and E6 are located in completely separate, but adjacent rooms in the seismic Category I control building, cs shown on Figure 8.3-27.
Buses E5 and E6 are connected to the auxiliary transformers via non-Class 1E non-segregated phase bus duct.
The bus duct is supported by seismically qualified supports in the control building. Taps in the bus duct provide the power to non-safety-related Buses 3 and 4 from the bus duct runs to Buses E5 and E6, respectively.
The tie between the non-safety related bus ducts and the Class IE switchgear is through Class IE air circuit breakers.
2.
Switchgear All Class IE switchgear has identical electrical ratings:
)
(a) Buses - 2000 ampere continuous rating, braced for 80,000 amperes momentary.
(b)
Incoming line breakers - 2000 ampere continuous rating, 350 MVA nominal interrupting capacity.
I l
8.3-3
Although these breakers are not Class 1E, the construction of the 13.8 kV switchgear is similar to the construction of the Class IE 4 kV switchgear. Periodic testing of these breakers according to the Technical Specifications further verifies their reliability.
A fault on one of the RCP motors assuming failure of its switchgear will be isolated by the fuses provided. Backop protection is provided by the incoming feeder to the switchgear; credit is taken for the 13.8 kV breakers mentioned above to provide backup protection because if it is assumed that the seismic event damages the nonseismic qualified 13.8 kV switchgear, then it will have to be assumed that the nonseismic qualified power source will also be damaged by the same event; thus, the circuit will be deenergized.
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b) Testing of protective devices used to satisfy Regulatory Gaide 1.63.
i..
The attached marked-up FSAR page, 8.3-22 providee information on the above subject.:
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Amendment 45 SB 1 & 2 FSAR June 1982 i
is NEMA Class B as a minimum, with the actual insulation for motors class selected on the basis of environment and service conditions g
in which the motor is required to operate. The factors taken gE into consideration in selection of the insulation system are resistance s
to radiation, resistance to moisture, resistance to chemicals, g
y ambient temperature and pressure. The motor enclosure is selected dy to protect against adverse environmental conditions. Winding temperature detectors and bearing thermocouples are provided on 97j large motors to alarm high temperature conditions.
J ys The motor suppliers are required to verify that actual test Jj data confirm that the torque margin is eqdal to or greater than 9
that of the calculated data.. A further check of motor capability k,
is the preoperational testing conducted at the site under plant j
light load conditions, to simulate the maximum voltage practically j'
obtainable, and under plant heavy load conditions, to simulate the 30 wj minimum voltage practically obtainable (reference.Section 14.2.6, exceptions to Regulatory Guide 1.68).
s V
j.
Provisions for Periodic Testing and Maintenan3 The onsite ac distribution system for engineered safety features J
loads is designed and installed to permit >eriodic inspection and testing in accordance with General Design Criterion 18, IEEE d7 J
Standard 308-1971, Regulatory Guide 1.118, Rev. 2 and IEEE 338-1977
[- -
J to ensure:
U 1 O
1.
The operability and functional performance of the components j0 V
of the system, and d
The operability of the system as a whole under design conditions.
j 2.
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Switchgear and accessories for the auxiliary power system are U
3,,}
easily accessible for inspection and testing.
J av5 The 13.8 kV, 4160 volt and 480 volt switchgear circuit breakers 74 v individual equipment is deenergized. The I
may be tested when the breakers can be placed in the test position and tested functionally.
j j
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The first and second level undervoltage schemes (see Subsection 8.3.1.1.b.4) are designed to permit periodic testing during normal c-
"j plant operation.
Breakers for engineered safety features auxiliaries are exercised on a schedule similar to that for the auxiliaries controlled by
], h,$
Transfer schemes can be exercised during normal Mf the breakers.
J operation, or by simulation of the necessary conditions. Timing Protective relays d
checks can be performed on transfer schemes.
V are provided with test plugs or test switches to permit testing d
and calibrating the devices.
w 8.3-22
l I
i c) Penetration protection for_460 volt-distribution panels located inside containment.
j i
1' The attached marked-up FSAR pages, 8.3-9 and 8.3-9a provide information on the above subject.
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/
FSAR September 1982
<n c
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device may possibly be af fected because of its proximity.
"f However, in this instance, no penetration damage can 4 I occur because all short circuit current flow will be d }'
diverted to the new fault located at the protective device.
3
,c Therefore, there is no conceivable electrical failure gj that could prevent both the protective devices from 4 y operating and at the same time allow the fault current to flow through the penetratian t/
460 volt loads inside the containment,'wTildn' arc fed 13 J 5
directly from the 480 volt unit sub* station, satisfy the hg*y requirements of Regulatory Guide 1.63 by utilizing the
-4 g
load breaker as primary protection and the unit substation
[] J incoming feeder breaker as backup protection.~ See Figure 8.3-48 for an example.
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3 d -(s f
machines, welding receptacles, etc.) are not provided
,dya with redundant protection because their circuits are de-
~
g jy energized and padlocked at the unit substation or motor a3-o control center during normal plant operation. Verification Tf,"
of the circuits being de-energized is part of the Technical
'> l 7 Specifications. Though some of these circuits may be
(
ah)I,j, required f r brief durations during plant operation such 0
6 as prior to or af ter refueling outages, lack of redundant protection is justified because of the very limited usage
.T 5' # g in this mode and the fact that such usage will be under i
Yi lh5m Technical Specification requirements.
Control circuits powered from 120 V ac or 125 V dc h
distribution panels have dual protective devices (circuit
( '4 j breakers and/or fuses) to provide penetration protection 9_<a)/'
in accordance with Regulatory Guide 1.63.
4 ds t
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Control circuits powered from limited capacity power sources such as control power transformers (maximum capacity 150 VA) and instrumentation circuits do not require dual protection because the short circuit versus time capacity of 'their power sources is within the penetration capabilities.
47
(
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8.3-9a
SB 1 & 2 Amendment 47 FSAR September 1982 I
7.
Special 480 Volt Criteria (a) Protection of Containment Electrical Penetrations The Class-1E and non-Class 1E 480 volt unit substations, 460 volt motor control centers and the distribution panels which feed loads inside the containment are all qualified to meet Class 1E requirements and are located in seismic Category I structures. 460 volt loads inside
.the containment are fed from distribution equipment with special provisions to satisfy thy requirements of-Regulatory Guide 1.63 for containment electrical penetration protection. These provisions are outlined below.
460 volt loads inside containment which are fed from motor control centers or distribution panels are provided with one of the following special arrangements to insure that the penetration integrity is maintained.
(1) Circuits of motors 5 hp and less are provided with-two identical combination starters.
Both units are located in the same compartment of the MCC.
See Figure 8.3-hY'for typical coordination curves.
(2) Circuits of motors greater than 5 hp are provided with a thermal magnetic breaker in series with a combination starter. Both the breaker and the combination starter are located in the same compartment of the MCC.
See Figure 8.3-ks for typical coordination curves.
'f1 (3s Feeder circuits, including the pressurizer heater circuits, are provided with'two identical thermal
- magnetic breakers. Both' breakers are located in 4
l the same compartment of the MCC or panel.
l The motor control centers containing these special protective devices are located in the Control Building l
switchgear area, with one exception,'the panels for the pressurizer heater circuits are located in the i
electrical penetration area outside the containment.
Both the primary and the backup protective devices are qualified IE devices.
There are no high or mode' rate energy lines in the above i
areas, therefore, only faults within the electrical devices could conceivably damage these protective devices.
If a protective device fails catastrophically j (
while clearing a short circuit, the second protective
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8.3-9
7_
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d) Use of thermal overload relays it. penetration protection.
Refer to FSAR Figure 8.3-50; it can be seen in this typical protection scheme that the thermal overload relays identified in the figure by numbers (2) and (4) are utilized to protect the penetration from a " low grade" fault whose magnitude will not trip the magnetic part of the protective device.
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1 e) Capability of penetration to handle long duration overcurrents.
(R4(evuu.a @AT. A3o.56)
The manufacturer of the penetrations has furnished damage curves /which establish the duration and magnitude of overcurrents that.the penetrations can sustain. Typical curves are shown on FSAR Figures 8.3-47, 8.3-48, 8.3-49 and 8.3-50.
It can also be seen in these figures that the curves of the protective devices are, in all cases, to the lef t of the penetration damage curves; thus, the protection provided will assure that long or short duration overcurrents that are capable of damaging the penetration will be interrupted before they cause damage.
s
In regard to the following open items appearing in Reference (c) of the cover letter, we are providing the following response:
OPEN ITEM - Physical Independence (GDCt7) a.
Independence between Class 1E and non-Class 1E.
c.
Identification of safety related associated circuits.
RESPONSE
a.
Our response to RAI 430.149, transmitted via our letter SBN-347, dated October 26, 1982, provides information on this subject.
c.
The open item refers to safety related associated circuits; there are no safety related associated circuits at Seabrook; however, as it has been explained in RAI 430.149 and in the FSAR, the associated circuits in the Seabrook design meet the requirements placed on Class lE circuits but they are not safety related.
The physical identification of the associated circuits is addressed in Section 8.3.1.3 of the FSAR.
Regarding the identification of associated circuits on documents; there is no requirement for the associated circuits to have specific identification on documents i.e.,
Train association, etc., and therefore none is provided.
Our method of identification of safety related circuits is in accordance with IEEE Standard 494. Documents clearly identify what is nuclear safety related and what is not.
i
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