ML23251A014
| ML23251A014 | |
| Person / Time | |
|---|---|
| Issue date: | 08/31/2023 |
| From: | NRC/OCIO |
| To: | - No Known Affiliation |
| Shared Package | |
| ML23251A034 | List: |
| References | |
| FOIA-2023-000163 | |
| Download: ML23251A014 (1) | |
Text
AC-REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)
ACCESSION NBR g 8409140165 DOC ~ DATE: 84/09/13 NOT'ARI'ZEDS YES, DOCKET FACIL:50 410 Nine= Mile Point Nuclear Station< Unit 2i Niagara Moha 05000410 AUTH, NAME AUTHOR AFFILIATION MANGANiC,VN . Niagara Mohawk Power Corp, RECIP ~ NAME~ RECIPIENT AFFILIATION SCHNENCER i A, Licensing Branch 2
SUBJECT:
Forwards responses to SER Open Items 421 3i4i lOF 13F15F18y20>>
23<25i27g28i34i36i37,42i43>>44 L 47ito aid in NRC revi,ew of application for license; Info will be included in next'SAR amend ~
DISTRIBUTION CODEF B001D COPIES RECEIVED:LTR .ENCL SIZE':
TITLE': Licensing Submittal: PSAR/FSAR Amdts 8 Related Correspondence I
NOTES:FNL icy FSAR'S & AMDTS ONLY, LleasW@l Glk. 05000410 RECIPIENT RECIPIENT COPIES ID CODE/NAME LTTR ENCL' ID CODE/NAME LTTR ENCL NRR/DL/ADL NRR LB2 BC ,1 0 NRR LB2 LA 1 HAUGHEYFM 01 1 1 INTERNAL; ADM/LFMB 1 ELD/HDS3 1 IE FILE" IE/DEPER/EPB 36 3 IE'/DEPER/IRB 35 1 IE/DQASIP/9 AB21 1 NRR/DE/AEAB 1 NRR/DE/CEO 11 1 NRR/DE/EHEB 1 NRR/DE/EQB 13 2 NRR/DE/GB 28 2 NRR/DE/MEB 18 1 NRR/DE/MTEB 17 1 NRR/DE/SAB 24 1 NRR/DE/SGEB 25 1 NRR/DHFS/HFEB40 1 1 NRR/DHFS/LQB NRR/DHFS/PSRB NOTES'OPIES 1 1 32'RR/DL/SSPB 1 NRR/DS I/AEB 26 NRR/DSI/ASB 1 NRR/DSI/CPB 10 1 NRR/DSI/CSB 09 1 NRR/DS I/ICSB 16 1 NRR/DS I/METB 12" c 1 NRR/DS I/PSB 19 1 N .L/ B 22 1 NRR/DS I/RSB '23 R F1 04 1 RGN1 3 M I/MIB 1 EXTERNAL) ACRS 4$ BNL(AMDTS ONLY)
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V MIIASAIRA 9 PgQO@g~g NIAGARA MOHAWK POWER CORPORATION/300 ERIE BOULEVARD WEST, SYRACUSE, N.Y. 13202/TELEPHONE (315) 474-1511 September 13, 1984 (NMP2L 0152)
Mr. A. Schwencer, Chief Licensing Branch No. 2 U.S. Nuclear RecpQatory Conmission Washington, D.C. 20555 Re: Nine Mile Point Unit 2 Docket No. 50-410
Dear Mr. Schwencer:
4 Enclosed for your use and information are the Nine Mile Point Unit 2 responses to the Nuclear RecpQatory CoIImission's Safety Evaluation Report open items. This information has been previously discussed with your staff and is submitted to aid your review of the Unit 2 licence application for the resolution of these open items. This information includes Safety Evaluation Report open items 421.3, 421.4 421.10, 421.13, 421.15, 421.18, 421 20 I 421 23'21 25I 421 27 421.43, 421.44, 421.47.
421 28 g 421 34 g 421 36 g 421 37 '21 42 g The enclosed will be included in the next Final Safety Analysis Report Aaendrrent.
Very truly yours, C.V. Mangan Vice President Nuclear Engineering 6 Licensing NRL-ga Enclosure xc: Project File (2)
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter of Niagara Mohawk Power Corporation ) Docket No, 50-410 (Nine Mile Point Unit 2)
AFFIDAVIT C.V. Mangan, being duly sworn, states that he is Vice President of Niagara Mohawk Power Corporation; that he is authorized on the part of said Corporation to sign and file with the Nuclear Regulatory Commission the documents attached hereto; and that all such documents are true and correct to the best of his knowledge, information and belief.
Subscribed and sworn to before me, a Notary Public in and the the State of Maryland and County of Montgomery, this 13 day of September 1984.
Notary Public in an for Montgomery County, Maryland My Commission expires:
Nine Mile Point Unit 2 FSAR QUESTION F421.3 (7.1, 7.2, 7.3, 7.4, 7.5, 7.6) 1. 10 Identify any "first-of-a-kind" instruments used in or 1. 12 providing inputs to safety-related systems. Identify each , 1.14 application of a microprocessor, multiplexer or computer 1.15 system where they are in or interface with safety-related systems. 1.16 RESPONSE 1.18 Per k>~.
The Unit 2 transient analysis recording system utilizes the 1.20 validyne remote signal multiplexer, MC3TOAD-Q2, to provide 1.21
'solation of lE signals from non-1E equipment. The 1.22 multiplexer unit a d associated plug-in signal conditioning modules provide he signal conditioning, multiplexing, and 1.23 A/ conversion to rocess and transmit up to 32 channels of nput data.
The following components describe the multiplexer and its 1.25 associated components:
- 1. MC370AD-Q2 Remote multiplexer/module case 1.28
- 2. AB295,-Q2 Ana1og multiplexer board 1.29
- 3. AD296-Q2 A/D converter hoard 1.30 PS294-Q2 Multiplexer/AID power supply brand 1.31
. 5.
6.
7.
8.
PS171-Q2 PS324-Q2 CD173-Q2 BA332-Q2 Signal/conditioning power, supply Remote DC power supply High gain carrier Buffer amplifier demodulator 1.32 1.33 1.34 1.35
- 9. BA332-150-Q2 Buffer amplifier 1.36
- 10. DI338-24-Q2 Digital encoder plug-in module 1.37 For details of testing against EMI, short circuit failures, 1.40 voltage faults and/or surges, and the summary of performance 1.41 characte risti cs, see the Environmental Qualification document. p gs ~~,.g ~g Pei F i~s ~~ cc ~I'~c I +i~~]~ iS Pace 6>> g g< @~A fi] P(Lg P ~seg~4>> ay~ ~
The Unit. 2 digital radiation monitoring system (DRMS) l. 43 supplied by Kaman Instrumentation provides isolation of lE digital, analog, and communication signals from non-1E 1.45 equipment.
The following modules describe the DRMS isolators. 1. 47 KESIM Kaman safety radiation monitoring: 1.50 system isolation module provides electrical isolation between the serial data lines of the Class 1E .1.51 data acquisition units and the non-lE redundant microcomputers.
Amendment 9 .QEcR F421 3-1
~ March 1984 ch1217718fqr9c 02/06/84 112 NOR 8u091u0165 u10 Nine Vii i 1
Nine Mile Point Unit 2 FSAR
- 2. KEI-D Kaman digital isolation module 1.54 provides isolation between the Class 1E and non-1E digital signals.
3 . KEI "A Kaman analog isolation module 1.57 provides isolation between the Class 1E and non-1E analog signals.
For details of testing against EMI, short circuit failures, 2.1 voltage faults and/or surges, and the summary of performance 2.2 characteristics, see the Environmental Qualification 2.3 document.
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Ge/9o NINE MILE POINT 2 FSAR OUESTION 421.3 Identify any "first-of-a-kind" instruments used in or providing inputs to safety-related systems. Identify'each application of a mircroprocessor,.multiplexer or..computer system where they are in or interface with safety-related 7.3 systems. P I
7.S) 7.6)
RESONSE hiss
) There are no "first of-a -k<-nd" instruments ued $n or provSdfng
'nputs to safety-related systems.
- 2) Microprocessors are used as an integral part of.the Redundant Reactivity Control System (RRCS). Four microprocessors (2/division) receive input signals, (e.g. low water level, high
.dome pressure, APRM downscale), process them against a time base
.formula, and generate output signals (e.g. ARI, recirc pump trip, feedwater runback) to other systems. Details of RRCS operation are discussed in Section (7.6.1.X). In addition to these data processing microprocessors, the RRCS has 4 microprocessors 2/division) for monitoring power supply status, 2 microprocessors
~
1/division) for assisting in the calibration of RRCS process
~ ~
instrumentation, and 2 microprocessors (1/division) to perform
~
automatic on-line testing. of the safety related RRCS system.
Hardware failures are annunciated and faults localized via use of a local keyboard/display.
The performance monitoring system (PMS), which interfaces with safety-related systems, is a non-safety-related system. Isolation of safety-related inputs to the PMS is shown functionally in the
-- logic diagrams and elementary diagrams listed in Table 1.7-1 and provided to the NRC.
KC:pes:pc/1188-3.2 4/26/84
Nine Mile Point Unit 2 FSAR (g cj QUESTION F421.4 (7.1, 7.2, 7.3, 7.4, 7.5, 7.6) 1.10 Section 7.1.2.3 of the FSAR provides a brief discussion on 1.11
- conformance to Reg. CQide 1.47. Discuss in detail the 1.13 design of the bypassed and inoperable statue indication 1.14 using detailed achamacica. Include Cho following 1. 15 information in the discussion:
- 1. Compliance vith the recommendations of Reg. Guide 1.47 1. 17 and Reg. Guide 1.22 Position D.3a and 3b; 1. 18
- 2. The design philosophy used in the selection of 1.19 equipment/systems to be monitored, inc uding auxiliary 1. 20 and support ayatama. 1.21
- 3. How the desi g n of the b YP ass and noperabla status 1.22 indication ayatama comply with Positions Bl through B6 1.23 of ICSB Branch Technical Position 21. 1.24
- 4. The list. of system automatic and manual bypasses aa pertains to the recommendations of Reg. Guide 1.47.
it 1.25 1.26
- 5. Discuss .hardware features- employed Co provide a 1.27 consolidated, human factored, display of the bypassed 1.28 and inoperable status of ESF equipment. 1.29 RESPONSE o 1.31
,p . gaa~hK
- 1. Refer to er for degrees of compliance<'+ ~S~ g f "f 1.33
- 2. Automatic bypassed and inoperable status indication ia 1.34 provided for all systems that affect, plant safety. ,This 1.36 indication system accompanies any operational procedure for all safety-related systems.
~r Any deliberate action vhich makes a support system 1.38 inoperable and also inhibits the proper function of a 1.39 dependant safety system will a'iso be automatically indicated.
~ 1. 40 r tha 1. 41~
is .annunciated (component evel). In this 1.42 ~
group, e equipment Chat when bypassed causes a system to be defeated, will contribute to its respective system 1. 43 level annunciator and indicate a bypassed
~
system condition.
it If a system is declared inoperable, vhether be a redundant portion of a system or a total system,
- 1. 44
- 1. 45 and also supports other systems, the bypassed and inoperable indication vill cascade into the dependent 1. 46
~p 4Jc~5 Amendment. 9 Q&R F421.4-1 March 1984 CIMl ~I chl21771 02/06/84 112
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Mnemonic Re uirements 2.36 4
ADS Automatic De ressurizati S stem VDC -'ÃI~P.38+
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. CCP Reactor Building Closed l.oop Cooling Mater 2.52 (Component Eevel'nly) 2.53 Containment Atmosphere Monitoring 2.55 CPS Primary Containment Purge (Component fevel 2.57 Only) 2.58 CSH High Pressure. Core Spray - Power Supply 3.2 CSK l,ow Pressure Core Spray 3.4 DER Reactor Building Equipment Drains (Component 3.6 f,evel Only) 3.7 EGA Standby Diesel Generator Air Startup 3.9 EGF Standby Diesel Generator Fuel 3.11 EGP Standby Diesel Generator Protection (Breaker) 3.13 EGS Standby Diesel Generator Protection 3.15 (Generator) 3.16 EJS Standby Station Service Substation 3.18 ENS Standby Station Service Supply Breakers 3.20 Amendment 9 QER F421. 4-3 March 1984 chl217718f qr9d 02/06/84 112
Nine Mile Point Unit 2 FSAR System S stems Containin B assed Ino erable Mnemonic Re uirements Fire Protection - Water (Component Level Only) 3.22 Feedwater System (Component Level Only)
GTS Standby Gas Treatment 3.26 HCS COL Hydrogen Recombiner 3.28 Control Building Air Conditioning 3.30 Control Building Chilled Water 3.32 Standby Diesel Generator Building Ventilation 3.34 i
Reactor Building Uentilation (CO Qg+l 4h .36 Yard struoture ventilation (tdenen+ Qrt[ onl4$ 3.38 IAS Instrument Air (Component Ievel Only) 3.40 MSS Main Steam (Component Level Only) -. . .3.42 Residual Heat Removal 3.44 eisa~
SFC Fuel Pool Cooling and 3.46 SWP n Service Water 3.48 WCS Reactor Water Cleanup (Component Level Only) 3.50
~l
- 5. It is not a requirement of Regulatory Guide 1.47 that the bypassed and inoperable status indication system be 3.57 3.58 Class 1E. However, the control circuit wiring is 4.1 associated with the Class lE components and is designed 4.2 in accordance vith Class 1E requirements. Optical '.3 isolation vill be employed to separate the annunciator, which is a non-Class lE circuit, from the bypassed and inoperable logic circuits.
The component level inoperability vill be displayed 4.6 using master specialty svitch-light units vithin their 4.7 respective operating area on the control panel.
The system=level indicators used are the standard plant 4.8 annunciator vindovs separated into divisions applicable. At the systcru level, the indicators will be if 4.9 4.10 accompanied by an audible alarm. All bypassed and 4.11 Amendment, 9 QEcR F421.4-4 March 1984 chl217718fqr9d 02/06/84 112
Nine Mile Point Unit 2 FSAR inoperable status indication has ga er illuminatio~
Amendment 9 @BR F421.4-5 March 1984 chl217718fqr9d 02/06/84 112
Nine Mile Point Unit 2 FSAR QUESTION F421.10 (7.1)
Section 1.10 of the FSAR provides a response to NUREG-0737.
The discussion on item II.D.3 does not mention alarms associated with the valve position indication. Confirm that alarms are provided zn conjunction with the position monitoring system. The discussions on items II '.3.13, II.K.3.21 and II.K.3.22 briefly address modifications that will be made to the RCIC and HPCS systems. Provide a detailed discussion on the design modifications proposed for these systems. Use one-line diagrams and other drawings as appropriate.
RESPONSE
See revised Section 1.10 for Item II.D.3.
TMI Item II.K.3.13 resulted in the-modification of the RCIC system to allow automatic restart after RCIC system shutdown due to high water level (Level 8) signal. Instead of tripping the RCIC turbine which required operator action to allow restart of the system, RCIC steam supply valve E51-F045 is closed, shutting down RCIC turbine/pump operation. Four separate transmitter/trip units energize individual relays, arranged in a one-out-of-two-twice logic configuration, to provide the closure signal for the valve F045. If the water level falls to Level 2, the system initiation logic will reopen the steam supply valve, restarting RCIC operation.
See revised Section 1.10 for Item II.K.3.21.
TMI Item II.K.3 '2 resulted in the modification of the RCIC system to allow automatic switchover of pump suction from the condensate storage tank to the suppression pool condensate storage tank falls to a preset low level.
if the Low level in the tank is monitored by two redundant level transmitters. If either transmitter senses low level, pump suction is automatically transferred to the suppression pool. These are different transmitters/trip units from those that activate switchover for the HPCS system. The condensate storage tank suction valve will be signaled to close upon opening of the suppression pool suction valve.
The RCIC elementary (Drawing No. 807E173TY, Rev. 14) shows the circuitry details.
The PSID and elementary diagram will be revised to reflect a relocation of the transmitter to the pump suction line.
Se ~<scil- Amendment 10 QEcR F421. 10-1 April 1984 e.Mg+ J.
Wc Mgc the length of time the operator will be required ocr hold the RCIC initiation button in a depressed condition to assure injection into the reactor. The concern is that if the manual initiation button is depressed only momentarily the opening of the RCIC injection valve will not be sealed in and reactor injection will not occur. The NRC has recently indicated tha they feel this design may not satisfy IEEE-279, Paragraph 4.16, which requires the s stem corn The logic for the RCIC injection valve E51-F013 is shown in Attachment
- 1. Contacts of relays K3, K20, and K40 must all be closed for F013 to open in response to a manual initiation signal or a low reactor water level 2 signal. Relay K3 is a momentary contact relay which is energized when the manual initiation button is depressed or when reactor water level is below Level 2. Relay K20 is energized when the turbine trip and throttle valve is partially or fully open. Since the trip and throttle valve is open during system standby the contacts of relay K20 will already be closed when RCIC is started. Relay K40 is energized when the steam admission valve E51-F045 is fully closed. Since F045 is closed when the system is on standby the contacts of K40 are open at that time.
Given this logic, to manually initiate RCIC and assure the injection valve opening is sealed in, the operator must maintain the initiation switch in a depressed condition until valve F045 comes off its seat causing closure of relay K40 contacts. A red-valve position indicating light ill inform the operator when F045 has started to open. At this t m he initiation switch can be released since the seal-in circuit in the MCC for valve F013 will now drive it to the full open position.
gJ Limit switch LS6 energizes relay K40 when valve F045 is fully closed.
gd Depending on the adjustment of this limit switch, it is not expected to take more than 1-2 seconds for relay K40 to be deenergized and its contacts closed when F045 starts to open. This is the time required for the operator to hold the initiation button down to assure vessel e injection.
- 0. ~
As explained in the above the contacts of relay K3 in the initiation logic have to be closed only 1 to 2 seconds before the injection valve opening logic is sealed in for automatic initiation. For an actual transient event requiring the RCIC system (i.e., loss of feedwater events) reactor ~ater level will be below the initiation level for well over this time required to seal-in the injection valve logic, since water level will not begin to recover until the RCIC and/or HPCS is initiated. It is GE's position that this meets the intent of IEEE-279 in that the RCIC system initiation will go to completion when it to perform its safety function. A momentary Level 2 lasting less required'or that 1 to 2 seconds is considered very unlikely and could only occur if feedwater flow is reestablished in time to reverse the water Mvel drop.
In this case it would be preferable not to initiate RCIC, thereby avoiding injection of cold water into the reactor.
0 0
In conclusion, GE considers the current RCIC design to be adequate and that it satisfies IEEE-279, Paragraph 4.16. Requiring the operator to hold the button for 1 to 2 seconds for a manual start does not impose a hardship on the operator. Normally, on a manual start the operator will stay with RCIC for at least 30 seconds or more to verify turbine speed, flow and valve positions. Operating procedures will include a precaution statement for the operator to ensure that he holds the manual initiation switch/button for RCIC until the valve position indicator shows the valve is opening.
~ 1
~ ~
Nine Mile Point Unit 2 FSAR QUESTION F421.13 (7.1, 7.2, 7.3, 7.4, 7.5, 7-6)
~ ~ ~ ~ ~ ~ 1.10 Various instrumentation and control system circuits in the l. 12 p?ant rely on certain devices to provide electrical l.13 isolation capability in order to maintain the independence 1.14 between redundant safety-related circuits and between 1.15 safety-related circuits and nonsafety-related circuits. 1.16 Provide the following information: 1.17 (1) Identify the types of isolation devices which are used 1.)9 as boundaries to isolate nonsafety-related circuits from 1.20 the safety-related circuits or to iso)ate redundant 1.21 sa foty-re 1atod ci rcui ts. 1.22 (2) Provide a summary of the performance characteristics 1.24 from the purchase specifications for each isolation 1.25 dovico idantified in rosponao to part (1) above. 1.26 (3) Describe the, type of testing that was conducted on the 1.28 isolation devices to ensure adequate protection against 1.29 the effects of electromagnetic interference, short- 1.30 circuit failures (line to lihe to ground), voltage 1.31 faults, and/or surges.
RESPONSE 1.33 Fc r. SOP
) The following list identifies the types of isolation devices that are used to isolate nonsafety-related circuits from the safety-related circuits or to isolate redundant safety-related circuits.
- 1. CE optical isolators r
- 2. Potter and Brumfield MDR relays
- 3. Valedyne multiplexers (MC37OAD-QE)
- 4. Kaman Industries isolation devices KESIMS (serial data line communication isolator)
~ b. KEI -D (di gi ta 1 i so l ati on module)
I C. KEI-A analog isolation module) 9-&&7 a
421.13 Insert Each type of device used to accomplish electrical isolation is tested to demonstrate isolation capability under maximum credible fault conditions.
These tests verify that the maximum voltage/current to which the device could be exposed within the panel/cabinet will not jeapordize the integ-rity of the class lE circuits. In addition, it will be shown that any destructive effects caused by application of the worst creditable fault will not jeapordize the function of any redundent divisional circuits or devices in physical proximity to the failed device. All isolation devices comply with environmental qualifications (10CFR50.49) and seismic qualifications requirements which are the bases for plant licensing.
0
~ NINE MILE POINT 2 FSAR UESTION 421. 13 Various instrumentation and control system circuits in the plant 7.1 rely on certain devices to provide electrical isolation capability 7.2 in order to maintain the independence between redundant safety-7.3 related circuits and between safety-related circuits and nonsafety-7.4 related circuits. Provide the following information:
7.5 7.6 '1)
Identify the types of isolation devices which are used as boundaries to isolate nonsafety-related circuits from the-safety-related circuits or to isolate redundant safety-related circuits.
(2) Provide a summary of the performance characteristics from the purchase specifications for each isolation device identified in response to part (1) above.
(3) Oescribe the type of testing that was. conducted on the isolation
.= "devices to ensure adequate protection against the effect of
.. electromagnetic interference, short-circuit failures (line to
. I.ine to grouhd), voltage faults;. and/or surges.
< ~
4sss RESPONSE The isolation-'devices used to electrically separate nonessential and essential circuits are designed to the guidelines of IEEE 384. Both relay and optical isolation devices are employed.
The optical isolators use a'fiber-optic light pipe'to electrically separate the input from the output. For example, an essential logic signal activates a light emitting diode; the light is transmitted through the light pipe to a photo switch; and the switch changes state upon receipt of the light signal and either blocks or transmits. These are the same types of optical >solators used in other GE plants.
The relay isolation devices provide a functionally equivalent degree. of separa-tion and are used typically for control voltage separation applications, i.e.,
120-Vac and 125-Vdc essential to nonessential and redundant essential circuits.
The relays are designed and- mounted so that a metal barrier separates the coil from the contacts with a mihimum distance of one inch between the coil and barrier and between the contact and barrier.
The designs of- isolation devices are responsive to the concerns regarding susceptibility to noise, shorts, surges, and faults. Adverse conditions affecting the coil or the semiconductor device cannot propagate through the isolation barrier (i.e., metal enclosure or fiber-optic light'ipe). Converse-ly, adverse conditions affecting the contacts or receivinq semiconductor cannot propagate through the isolating barrier and affect the coil or transmitting semiconductor. Therefore, essential systems or circuits are electrically isolated from nonessential and/or redundant systems or circuits.
BAN:rm/A08223" 8/30/84
0 NINE NILE POINT 2 FSAR Summary of Purchase Specification:
A. HDR RELAY
- 1. Design Specification
- a. HIL-R-19523
- b. Contract Specification
- c. Coil Specification
- d. Insulation Specification e., Design Life
- f. Reliability
- 2. Class 1E Safety Function
- a. Functional Specification
- b. Reli abi 1 ity h
- 3. gualification Testing Ambient and Desi g n Environments b; Normal Hountfng B. -'ISOLATOR'.
Application Data Specification
- 2. Performance Specification
- 3. gualification Testing
- a. Tested as a part of panel subassembly The documents listed above are available for review at the General Electric offices in San Jose, CA.
The optical isolator comprises semiconductors, resistors, and capacitors mounted on a printed circuit board. As designed, this device satisfies elec-trical isolation requirements.
The NHP-2 NSSS uses two qenerations of optical isolators to provide isolation/
separation between two divisional or divisional and nondivisional circuits.
The PGCC uses one generation of isolator cards, and the Redundant Reactivity Control System uses a later generation. The basic difference is that the later generation has current-limitsng resistors on its input circuits to more fully protect the card from damage due to excessive input signals. Installation in the panels is the same for both generations. Each is mounted in panel racks designed to hold..the input and output cards separated by a 1" quartz rod through a ceramic barrier.
Specifications control the type of testing and qualification required on the isolators. The basic difference is that line to line voltage tests (140 VDC for two minutes and 400 V pulse for one msec.) were performed on the new generation isolators. Instead of this test, an input circuit 5KV line-to-ground BAH:rm/A08223" 8/30/84
NINE MILE POINT 2 FSAR test was performed on the older generation isolators. In either case, subse-quent .to the test, it was confirmed that there was no degradation of the card on the other side of the barrier.
Additionally, the RRCS used isolated lamp drivers (card mounted relays) to isolate class lE signals from certain non-class 1E loads (p. g., indicators);
As part of its qualification, a 200 VDC line-to-line test across output con-tacts was performed to determine no degradation will be propagated back to the input circuit on the card.
Since the same kind of panel enclosures is used for both generations of isola-tors, running the 5KV test on the old generation will be sufficient to confirm the barrier (dielectric) capability for both generations of isolator cards and their housing. In addition, since the 5KV test greatly exceeds the voltage to be applied during the line-to-line test of the new generation cards, it can be considered equivalent to the test on the new generation cards, with respect to causing detriment to the cards on the other ssde of the barrier.
The isolator enclosures are designed to hold either four or eight isolator cards; only cards representing csrcuits from the same division are contained in the same enclosure.' worse case failure would only cause loss of function to one division; because of built-in redundancies in other divisions, safety functions would not be lost.
Copies of test plans, procedures, and results are on file at GE.
A summary of the qualification test performed on the MDR relay and the optical isolators are given in Attachments 1 and 2.
An additional test of the optical isolators to verify that they can withstand the maximum credible voltage applied in the'ransverse mode is being scheduled.
This test will verify that the maximum credible voltage applied to the optical isolator s in the transverse mode will-not be propagated through the quartz barrier to the other side of the device.
COMMENT TO SMEC/NMPC SPEC should provide the portion of the response concerning BOP devices used in electrical isolation. This response should be incorporated into SPEC/NMPC FSAR revision.
BAM:rm/A08223" 8/30/84
k ATTACHMENT 1 TO QUESTION 421.13
SUMMARY
OF QUALIFICATION TEST PERFORMED ON HDR AUXILIARY RELAY
- a. Normal Operation:
~ ~ ~
plication of normal coil rating voltage to coil terminals and
~
~
o servance of relay contact status change. Repeat test with
~
~
gradually removing applied voltage.
- b. Contact *Current Rating Test:
Application of contact rated load and observance of contact status change while relay coil energfzation and deenergization.
c'. Dropout and Pickup Voltage Test:
Gradual decrease and increase of relay coil voltage application, observance of contact status change.
- d. Response Time Test:
Energization and deenergization of relay coil and recording of cycle time.
- e. Dielectric Strength Test:
Application of appropriate voltage based on Hil Spec R-19523A (1230V for 120 VAC nominal, 2375V for 125 VDC nominal, 1265V for 24 VDC nominal) for one minute between relay coil circuit and relay main frame.
Acceptance Criteria - Relay shall not short out between coil circuit and contacts or frame during one minute exposure to applied voltage.
- f. Typical Test Set-Up (see Figure 421.13-1)'II.
SEISMIC TEST Clutter and contact bounce monitoring in the. energized and* deenergized state at different times during seismic excitation.
BAM:pc:rm/L08225*-1 8/22/84
Rela State NC Contact NO Contact Oe-energized 9 6.7g 5 msec. max. No transfer of contact Energized 0 17g No transfer of contact 2 msec. max.
IV. ENVIRONMENTAL TEST Exposure to temperature and humidity environment of each extreme and various conditions in between and demonstration of relay operation before, during, and after such exposure.
Environmental Exposure
- a. 710F~ HC RH 55oF'(g b
4loFH RH
- d. 614F, 35K RH e 81oF'ly RH
- f. 101 P, 65X RH
- g. 102 F, 80K RH
- h. 119 F 90K'RH V. CONCLUSION Test samples successfully demonstrated that the relay wi.ll function before, during, and after the test exposure environment. The relay met all functional requirements as specified.
BAH:pc:rm/L08225"-2 8/22/84
~ ~
NUCLKAR KNfAQY USINESS GROUP SKI)KRAK O KI.KCTRIC 40-901-Ib
)IK; Qh s~ ~o. II TYPICAL RECORDER CO lsiCTIG'.iS
+- Vo1 tage + Voltage Contact Coil I Y 6 0)i."
1030 GHN Recorder Recorder TYPICAL TEST SET-UP' Voltage
?n$ tietfng Device ~ ~ ~0 RECC)"DER RECORDER Lo ic Rela;
ATTACHMENT 2 TO QUESTION 421.13
SUMMARY
OF QUALIFICATION TEST PERFORMED ON OPTICAL ISOLATORS OEVICE Field Contact 004 5V Logic Input 204B6190AAG 003 12V Logic Input 204B6190AAG004 5V Logic Input 204B6190AAG005 High Speed Input. 204B6198AAG002 Analog Input Analog Input TEST'04B6186AAG 204B6208AAG002 204B6208AAG003 Floating Low Level Output 198B6241AAG003 High Level Output 204B6188AAG002 5V Logic Output 204B6194AAG002 High Speed Output 204B6196AAG002 Analog Output . 204B6220AAG002 Isolator Power Supply 198B6203AAG004 Optical Isolator, 133D9947G003 Optical Isolator 133D9947G004 FUNCTIONAL The optical isolators were tested to verify that they met the requirements as specified in 272A8638, Isolator Application Data Information Document.
III. SEISMIC TEST.
The optical isolators were tested using 22A4320 Seismic qualification Procedure for Class 1E Electrical Equipment Test Specification.
IV. ENVIRONMENTAL EXPOSURE TEMPERATURE F RELATIVE HUMIDITY RH DURATION 137 80'0K 100 hrs 153 8 hrs 70&+ 5 (Ambient) 50+15% (Ambient) 12 hrs 40 80K 100 hrs V. HIGH VOLTAGE TEST A 5KV hi-pot test was; performed on the Isolators to assure that electrical isolation between the input or output will not impair the function of devices on the other side of the barrier.
BAM: rm/A08294"-1 8/29/84
~ ~
VI. DETERMINATION OF TEST VOLTAGE A generic review of the voltage sources present within the plants utiliz-ing optical isolators indicated that 4160 volts is the maximum voltage that could conceivably be present. Therefore, a test voltage source of 5OOO volts was chosen.
The actual voltages that could be present in a panel are determined by a specific plant analysis.
VII. CONCLUSION Test samples successfully demonstrated that the optical isolators will function before, during and after the test exposure environment and meet the qualification requirements of IEEE 323-1971 and IEEE 344-1975. It was also demonstrated that electrical isolation is maintained between input and output.
BAM:rm/AO8294*-2 8/29/84
TYPICAL TEST CONFIGURATION FOR OPTICAL ISOLATORS XSo<A,~oK HouS[mg
~~ops>oa Chub S t" Bra.~i<. ~4r r>e.r Figure 421.13-2 Figure 421.13-3
88r28~ 88 14 S8kJ CH3C 2739 2R M3. 881
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Nine Nile Point Unit 2 FSAR
( P QUESTSN F421.15 {7.1, 7 3, 7.7) ~
Tabl>> 7.1-2 of the FSAR provides a listing of th>> safety 1.12" related systems simi 1 arity to licensed reactors. Eleven 1.14
~ yscemo are shown to have no similarity. i'or these systems sufficient design details have not been provided to enable 1 '5 1.16 the NRC ataf f to v>>rf fy con fnnnance to the acceptance 1.17 criteria of the Standard R>>view Plan (NURRG 0800). For each le 19 of these oyst>>ms provid>> a detailed comparison of the d>>sign 1.24 to the app)icable raquirem>>nte and recommendations delineated f n Table 7-1 o f NUREG-0800. 8 pecif ical ly X.22 identify and )beatify deviations from thea>> provisions.
I
RESPONSE
S>> S>>ction 7. l, Tabl>> 7 I-S for applicability of standards; 1.27 S>>ction 1. 8, 'able 1.6 J for D>>gree of Compliance
~
to Recclscosy ccidssI sod ssctios 7.1.1.2 soy sppli'cable ~s 1.29 section JSs syscew dssccipciop doe'Beep~~.
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>) st~i ~ I eyyIy 1 o hmendment 9 QhR F421.15>>1 March 1984 ch1217718fqr9ai 02/06/84 I
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/P(f 421.1 8+provide a detailed discussion on the methodology used to es ab'lish (7.2)
(7. 3) the technical specification trip setpoints and allowable values for s the Reactor Protection System (including Reactor Trip and Engineered Safety Feature .channels) assumed to operate in the FSAR accident and transient analyses. Include the following information:
(1) The trip setpoint and allowable value for the technical speci.fications.
(2) The safety limits necessary o protect the integrity of the physical barriers which guard aoainst uncontrolled release of radioactivity. The safety limits should be the limits established for licensing purposes, for example the technical specification safety limi s on minimum critical power ratio (1.06), and reactor coolant system pressure (1325 psig).
The values assigned to each component, of the combined channel error allowance (e.g., modelino uncertainties, analytical uncertainties, transient overshoot, response time, trip unit setting accuracy, test equipment accuracy, primary element accuracy, sensor drift, nominal and harsh environmental allowances, trip unit driat), the basis for these values, and the me hod used to sum the individual errors. Mhere zero is assumed for an error a justification that the error is negligible should be provide .
(4) The margin (i.e, the difference be ween the s fety limii and the setpoint less the combined channel error allowanc ).
Identify any trip for which the setpoint and allowable value in the technical specifications will be assigned bes- estimate values and for which you do not have an analysis of errors and/
or uncertainties to confim that the trip function will occur before the actual value of the measured parameter exc ds that assumed in the plant safety analysis. Provide justification for this nonanalytical aporoach.
Res e~ st t Ql'.6 S J- cAy~ g~ C
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NINE IlILE POINT 2 FSAR 4FI4!
QUESTION 421. 20 Operating reactor experience indicates that a number of (7. 2) failures have occurred in BWR reactor vessel level sensing (7-3) lines and that in most cases the failures have resulted in (7.a) erroneously high reactor vessel level indication. For BWRs, cownon sensing lines are used for feedwater control and as the basis for establishing vessel level channel trips for one or more of the protective functions (reactor scram, HSIV closure, RCIC, LPCI, ADS or HPCS initiation). Failures in such sensing lines may cause a reduction in feedwater flow and consequential defect of a trip within the related protective channel.
If an additional failure, perhaps of electrical nature, is assumed in a protective channel not dependent on the failed sensing line, protective action may not occur or may be delayed long enough to result in unacceptable consequences. This depends on the logic for combining channel trips to achieve protective actions.
Identify each case where a reactor vessel water level tap or sensing line failure concurrent with an additional random single electrical failure induces a transient and precludes the automatic operation of reactor scram and/or engineered safety feature system. For 'each case identified provide an. evaluation which demonstrates how the redundancy or diversity of the plant design provides for reactor scram or safety system operation within acceptable limits. Where manual action is required by the operators discuss the instrumentation and time available for the operator to take such corrective action.
To reduce the consequences of sensing line failures in combination with a single failure in a protection channel not dependent on the failed sensing line, a modification of the protection system logic may be required. Logic configurations which may be considered for NRC approal on this plant are described in the BWR owners group study entitled "Review. of BWR Reactor Vessel Mater Level Measurement Systems", SLI-8211, prepared by.
S. Levy Inc.
RESPONSE
A postulated break in an instrument line plus an additional failure is beyond the design basis for this plant; however, an assessment of plant response to this event has been performed on the basis of the following methodolo'gy and assumptions.
Nethodolo Determine the logic for combining channel trips to achieve protective actions.
KC: pc/L01262"-1 4/27/84
NINE MILE POINT 2 FSAR 4&Sf
- 2. Identify each case where a reactor vessel water level tap or sensing line failure concurrent with an additional random single electrical failure induces a transient and precludes the automatic operation of a reactor protection and/or an engineered safety feature (ESF) system.
Assum tfons Instrument reference line failure (break)
Single electrical device failure (no power supply failure)
ARI operable No operator action.
A review of various failure combinations resulted fn the identification of the worst postulated failure path as the failure of division 1 instruments reference leg line (f.e., connected to condensing chamber B21-D004A) combined with a failure 4high" of B21-NOBOC.
The manual selection switch for feedwater. controller is assumed to be onthe failed instrument line, and the operator fs assumed not to switch control to the other instrument. line as would be expected. This causes the feedwater controller to respond to the high water level error signal by reducing the feedwater flow. Following the loss of feedwater, water level will decrease to level 4 initiating a low water level alarm. Water level vill further decrease to level 3 initiating a second low water level alarm, and reactor scram vill not occur due to the assumed failure.
When water level decreases to lev'el 2, a third low water level alarm wfll be initiated, reactor scram will occur due to Alternate Rod Insertion (ARI). RCIC system vill automatically start, and both recirculation pumps will trip.'PCS system fs unavailable (tripped) due to the assumed faf lure.
The core thermal hydraulic analysis using the REDY transient code shows that the water level inside the shroud drops to a minimum of 1.9 feet above the top of the active fuel at 1436 seconds and slowly rises thereafter.
Since the core remains covered throughout the transient, no core heatup fs expected.
Note: The justification of 'Assumption 2 fs as follows:
Section 4.4.3 of BWROG-8253, "BWR Owners Group Reactor Vessel Water Level Measurement System Report," from T. J. Dente (BWROG) to H. R. Denton (NRC), dated August 13, 1982, stated: "...the ATMS events...indicate that mechanical failures, not instrument failures, fn the system...are the largest contributor to core melt.. Events involving electrical failure, which included instrument failures, are less than O.lX of the total core melt frequency."
KC: pc/L01262*-2 4/27/84
NINE MILE POINT 2 FSAR UESTION 421. 23
~ f ~ $ ~ j 705)
Provide an evaluation of the effects of high temperatures on reference legs of water level measurinq instruments subsequent to high energy line breaks, including the potent>al for reference leg flashing)boil-o f, the indication/
annunciation available to alert the control room operator of erroneously high vessel level indications resulting from high temperatures, and the effects on safety systems actuation (e.g., delays).
RESPONSE
High drywell temperature does not significantly affect measured reactor water level when reactor pressure is greater than the saturation pressure of water in the water level sensing lines because the vertical drop of the wide range, narrow range and fuel zone range reference and variable leg sensing lines in the drywell are approximately equal. The water level indication is not affect-ed because the comparable vertical drops of the reference and variable leg sensing lines in the drywell result in nearly equal changes in hydrostatic pressure in these lines due to reduced water density at increased drywell temperature.
If reactor pressure decreases to less than the saturation pressure of the water in the water level sensing lines, the water in the lines will flash and boil.
,The flashing and boiling may result in loss of some of the water in the sensing lines. Loss of water from the sensing lines results in reactor water -level measurement error until'perator action refills the sensing lines.
Analyses have demonstrated that water level activated safety trips will be initiated for high energy line breaks before reactor pressure decreases to less than the saturation pressure of the water in the sensing lines. Therefore, these safety trips will be initiated before high drywel1 temperature signifi-cantly affects water level measurement.
The NMP2 containment monitoring design consists of eighteen redundant Class 1E temperature elements distributed throughout the Primary Containment: The
'ontainment temperature monitoring system constantly scans and selects the highest containment temperatures fol control room indication and annunciation.
The control room indication of containment temperatures includes metered indication as well as temperature recorders. The control room annunication alerts the operator of high containment temperatures which could lead to possible erroneous level indication. r Long term (i.e. following RPV blowdown and ref looding) water level measurement errors due to flashing and 'boiling of water in the sensing lines are postulated to occur as a result of multiple failures by the operator to follow established 8AM: csc: rm/I08271""1 8/29/84
emergency procedures. The BWR Owners Group (BWROG) has established the posi-tion with the NRC that potentially large water level measurement errors resulting from high drywell temperature increase the probability of core melt and that these errors should be minimized and/or eliminated. This position was established with the NRC via the BWROG reports ¹SLI-8211, titled "Review of BWR Reactor Water Level Heasurement System" and ¹SLI-8218, titled "Inadequate Core Cooling Oetection in BWR's" prepared by S. Levy, Inc.
This response provides the results of aii evaluation of the NHP2 reactor water level sensing line arrangement in the drywell based upon the criterion accepted on the Shore4am docket. Specifically, t'he acceptance criterion is that:
"Following initial reactor water level stabilization after reactor depres-surization and assuming the operator fails to properly monitor reactor water level during long term (on the order of hours) post-LOCA conditions, the operator shall receive a low reactor water level alarm before the lower tap is uncovered."
It should be emphasized that the stated criteria is based on the assumption of multiple failures., Under the highly unlikely scenario postulated it is assumed that the operator:
- 1) fails to properly monitor reactor. water level (i.e., the most impor-tant post-LOCA parameter),
- 2) stops all systems providing reactor core inventory,
')
fails to properly monitor drywell temperature 'and vessel pressure and reflood t'e reactor in order to recover/restore water level indica-tion, as required by the emergency procedures, when drywell tempera-ture near the- instrument lines exceeds that saturation temperature of the reactor vessel, and
- 4) fails to initiate the drywell spray at the high drywell temperature specified in the emergency procedures.
The evaluation assumes loss of all water in the part of the reference. leg sensing line located in the drywell as a result of failure of the operators to follow established emergency procedures. The loss of water from the reference leg is assumed to occur due to high drywell temperature conditions (i.e.,
flash-off that occurs when the reactor is depressurized and long-term boil-off of water due to drywell temperatures higher than reactor temperatures). The error resulting from flash-off and boil-off is proportional to the evati on change of the re ference 1 eg in the drywel 1 . NMP2 has a maximum of vertical'l g // ~~ee4. vertical elevation of reference legs in the drywell. This evaluation is based on the nominal trip setpoint.
The results of this evaluation indicate that a low reactor- water level 2 alarm will be received before the lower instrument tap connected to the alarm is uncovered.
BAH: csc: rm/I08271"-2 8/29/84
Nine Mile Point Unit 2 FSAR QUESTION F421.25 (7.2, 7.3, 7.4, 7.5, 7.6, 7.7)
Reg. Guide 1.118, which provides guidance with respect to periodic testing of the reactor protection system engineered safety features and supporting systems, (reactor'rip, RCIC) excludes lifting of leads to perform surveillance tests and accepts opening of a breaker to perform surveillance tests only if opening of the breaker causes the trip of the associated channel. Confirm that the Nine Mile.
Point Unit 2 surveillance tests will conform to the above cited guidance.
RESPONSE
See Table 1.8-1, Regulatory Guide 1.118.
said, Crf $ hsn$ Mh Periodic surveillance testing procedures i for the reactor protectio system are currently under development. Lifting of leads is not normally utilized in procedure steps unless the procedure would require this action to 'ccomplish the =
purpose of she procedure and no other method, suoh as ov opening a breaker, is possible. Ahenever lifted leads are Su utilized, an analysis will be performed to Verify that this is the only practical method available to accomplish the kfto f surveillance. Within the body of a procedure speci ic instructions are given to identify, lift, and replace the f
leads Such actions are performed under strict admini trative control with independent verification for safet related systems or components.
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CE/Q NINE NILE POINT-2 FSAR QIIEom 421.27 Mode switch contact and mode switch operating mechanism have caused inadvertent protective actions.'alfunctions
.Similar malfunctions could have rendered redundant channels of protective functions inoperable. IE Information Notice 83-42 provided noti-fication of potentially significant events concerning mode switch malfunctions. Section 7.2..1 of the FSAR indicates that the reactor mode switch is used to bypass and enable protective functions, rod withdrawal interlocks and refueling equipment interlocks. Provide a detailed discussion on how the mode switch is.incorporated into the overall design, supplemented with detailed drawings and schematics.
Please include the following:
(1) Identification of the reactor protection system, rod block, refueling interlock and other, functions important-to-safety that are dependent on proper mode switch contact operation.
(2) Identification of the analyzed transients and accidents where
. credit is taken for the operation of any function identified in (1) above.
(3) The surveillance actions necessary to positively verify mode switch contact positions, detect mode switch contact failures and detect mode switch operating mechanism failures for each function identified in (1) above.
RESPONSE dn assessment of the system impact of postulated misopeeatioes for the presently installed mode switch is provided below. The assessment includes an 'valuation of the impact of postulated misoperations on the analyies described in Chapter 15. It identi-fies normal switch contact positions for each mode of operation (RUN, SHUTDOWN, REBEL, and STARTUP), and summarizes the consequnce should one or more pairs of contacts misoperate. '11 of these misoperations are detectable by annunciation, instrumentation checks, a S~urveillance testing.
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1-0273 1
Assessment of Effects of Mode Switch Miso eration C
NOTE: Each pair of svitch contacts is . identified by identical digits vith the letter C as a suffix on one digit <(e.g., 1-1C, 2-2C, etc.). For brevity, only one digit vill be used, thus:
contact 1, contact 3, etc.
I. Contacts Normall Closed in RUN Position
).1 IRM B ass Contacts-3 5 19 21 35 37 51 53 1.1.1 If any of the above contacts are open in the RUN position the IRM scram function would be enabled and half-scrams or scrams could result if IRM's vere upscale'or inoperative.
1.1.2. If any of the above contacts are closed in STARTUP,.
REFUEL, or SHUTDOWN svitch positiions, the IRM scram function vould be bypassed. This would not be detected immediately. but vould be evident 'during weekly channel functional tests because half scrams due to the IRM function could not be induced in the affected channel(s).
1.2 Shutdown Scram Interlock Contacts-9 25 41 57 (Note: These contacts are also normally closed in STARTUP and.
REFUEL.)
1.2.1 If any of the above contacts are open in the RUN position a scram or half scram vill result.
1.2.2 If any of the above contacts are closed in SHUTDOWN, the shutdown trip function of the affected logic circuit vould be disabled.
1-0273 2
1.3 APRM Interlock Contacts-ll 27 28 43 44 59 1.3.1 If any of the above contacts are open in the RUN position, the APRM setdown scram trip function 'would be t
enabled and the APRM's would provide half acrams or full scrams at reactor power levels of 15'r greater.
1.3.2 If any of the above contacts are closed in any position but RUN, the APRM setdown scram trip function would be disabled and the trips setpoints would be raised to their high setpoint level of about 113'f reactor rated power.
1.4 Rod Block Interlock Contacts-30 62 1.4.1 If either of the above'contacts are open in the.RUN position, an annunciated rod block signal would be sent to the reactor manual control system to prevent removal of more than one control rod.
1.4.2 If either of the'bove contacts are closed in STARTUP, REFUEL, or SHUTDOWN, there would be an unannunciated permissive for the reactor manual control system to move more than one control rod. In the STARTUP or REFUEL mode, the permissive would be redundant.
1.5 Conclusions For multiple failures of mode switch contacts which are normally closed in the RUN mode, the principal concerns are:
- b. Failure to cause scram when moving the mode switch to SHUTDOWN-1-0273 3
- c. The unannunciated bypass of the APRN setdown scram function in switch positions other than RUN.
- d. The unannunciated permissive to move more than one control rod when in the SHUTDOWN mode.
II. Contacts Normall Closed in STARTUP Positions Z.l KSIV Closure Scram B ass Contacts-7 23 39 55 (Note: These contacts are also normally closed in REFUEL and SHUTDOWN.)
2.1.1 If any of the above contacts are open in the STARTUP position, the MSIV closure scram trip function would be enabled vithout immediate operator knovledge, unless one of the tvo bypass'nnunciati'ons vere to cease. Thi:s vould require at least tvo of the four sets of contacts to open.
2.1.2 If any of the above contacts are closed in RUN, an annunciated bypass of the MSIV closure scram trip function would occur.
2.2 Shutdown Scram'Interlock Contacts-9 25 41 57 (Note: These'contacts are also normally closed in RUN and REFUEL.)
2.2.1 If any of the above contacts are open in the STARTUP position, a scram or half scram vill result.
2.2.2 If any of the above contacts are closed in the SHUTDOWN position, the shutdown scram trip function of the affected logic circuit would be disabled.
2.3 Steamline Lov Pressure Isolation Tri B ass Contacts-10 26 42 58 1-0273 4
(Note: These contacts are also normally closed in REFUEL and SHUTDOWN. )
4 2.3.1 If any of the above contacts are open in the STARTUP S
position, the MSIV isolation-on-low-steamline-pressure function would be enabled; A MSIV isolation trip or half trip would occur. The isolation trip could be followed by a scram or half scram on HSIV closure.
2.3.2 If any of the above contacts are closed in the RUN position, HSIV isolation on low steamline pressure would be bypassed.
2.4 Rod Block Interlock contacts-31 63 2.4.1 If ei;ther of the above contacts are open in the STARTUP position, an annunciated rod block signal would be sent to the 'eactor 'anual control system to prevent removal of more than one control rod.
2.4.2 If either of the above contacts are cloyed when not in STARTUP, there would be an unannunciated permissive for the reactor manual control system to move more than one control rod. In the RUN and REFUEL modes the permissive would be redundant.
2.5 Conclusions For multiple failures of mode switch contacts which are normally closed in the STARTUP mode, the principal concerns are:
- a. Qe annunciated bypass of the kSIV closure scram trip function in the RUN mode.
- b. The unannunciated bypass of the %IV-isolation-on-low-steam-line pressure function in the RUN mode.
1-0273 5'
- c. The failure to initiate scram whea the mode switch is moved to the SHUTDOWN position.
- d. The uaannunciated permissive to move more thaa onc, coatrol rod in the SHUTDOWN mode.
III Contacts Normall Closed ia REFUEL Position 3.1 HSIV Closure Scram B ass Contacts-7 23 39 55 (Note: Thyrse coatacts arc also normally closed in STARTUP and mmOWN.)
3.1.1 If any of the above contacts are open in the REFUEL position, the MSIV closure, scram trip functioa vould be enabled vithout immediate operator knowledge, unless one of the tvo bypass. annunciatioas vere to cease. This would require at least tvo of the four sets of contacts to be open.
3.1.2 If any of the above .contacts are closed in the RUN position, aa annunciated bypass of the MSIV closure scram trip 'function vould occur.
3.2 SDV Hi h Vater Level Scram ass Contacts-8 24 40 56 (Note: These contacts are also normally closed in SHUTDOWN.)
3-2.1 If any of the above contacts are open in the REFUEL position, the SDV high vatcr level scram trip function vould be enabled for the affected logic channel.vithout iamediate operator knowledge, ualess one of the tvo bypass aanunciatioas vere to cease. This would require at least two of the four sets of contacts to be open.
3.2.2 If any of the above contacts are closed in RUN or STARTUP, 1-0273 6
the SDV high-water-level scram trip bypass would be enabled. (h separate bypass switch for each channel must also be closed to effect the bypass.)
3.3 Shutdown Scram Interlock Contacts-9 25 41 57 0 ~
(Note: These contacts are also normally closed in RUN and STARTUP) 3 3.1 If any of the above contacts are open in the REFUEL position, a scram or half scram will result.
332 If any of the above contacts are closed in the SHUTDOWN position, the shutdown scram trip function of the
~ ffected logic channel would be disabled.
t 3.4 Steailine Low Pressure Iso)ation Tri B ass Contacts-10 26 42 58
'(Note: These contacts are also normally closed in STARTUP and Simoom.)
3.4.1 If any of the above contacts are open in the REFUEL
'position, the MSIV isolation-on-low-steamline-pressure function would be enabled. h ESIV isolation trip or half trip would occur.
3.4.2 If any of the above contacts are closed"in the RUN position, MSIV isolation on low steamline pressure would be bypassed.
3.5 Rod Block Interlock Contacts-29 61 3.5-1 If either of the above contacts are open in the REFUEL position, an annunciated rod block signal would be sent to the reactor manual control system to prevent removal of more than one control rod.
1-0273 7
3.5.2 If either of the above contacts are closed when not in the REFUEL mode, there would be an unannunciated pezmissive for the reactor manual control system to move more than one control rod. In the RUN and SThRTUP modes, the permissive would be redundant.
3.6 Conclusions For multiple failures of mode switch contacts which are normally closed in the REFUEL mode, the principal concerns are:
ao The annunciated bypass of the HSIV closure scram trip function in the RUN mode.
- b. The unannunciated bypass of the %IV-isolation-on-low-steamline pressure function in the RUN mode.'.
The failure to cause a scram when the mode switch is moved to the SHUTDOWN position.
- d. The unannunciated permissive to move more than one control rod in the SHUTDOWN mode.
IV. Contacts Normall Closed in SHUTDOWN Position 4.1 Shutdown Scram Reset Controls-1 2 17 18 33 34 49 50 4.1. 1" If any of the above contacts are open in the SHUTDOWN position, the shutdown scram/manual scram logic for the affected logic channel will not be configured to pezmit the logic channel to be reset after a scram trip.
4.1.2 If any of the above contacts are closed in any position 1-0273 8
45/Q except SHUTDOWN, there vould be no immediate effect. If both sets of contacts in any logic channel (1 and 2 for logic hl, 17 and 18 for logic Bl, etc.) vere.. closed vhen not in SHUTDOWN, the shutdovn scram function vould be in the "reset" configuration; and a scram trip vould not occur for that logic channel vhen the mode svitch is moved to the SAG'DOWN position.
4.2 NSIV Closure Scram ass Contacts-7 23 39 55 (Note: These contacts are also normally closed in STARTUP and REFIT..)
4.2. 1'f any of the above contacts arc open in the SHUTDOWN position, the NSIV closure scram trip function would be enabled vithout immediate operator knovledge, unless one of the tvo bypass annunciations vould cease. This vould require at least two of the four sets of. contact to be opene 4.2.2 If any of the above contacts are closed in the RUN position, an annunciated bypass of the MSIV closure scram trip'unction vould occur.
4.3 SDV Hi h Water Level Scram B ass Contacts-8 24 40 56 (Note: These contacts are also normally closed in REFUEL.)
4.3.1 If any of the above contacts are open in the SHUTDOWN position, the SDV high vater level scram trip function vould bc enabled for the affected logic channel vithout immediate operator hnovledge, unlesi one of the tvo bypass annunciations vere to cease. This vould require at least tvo of the four sets of contacts to be open.
4.3.2 If any of the above contacts arc closed in RUN or I 0273 9
STARTUP, the SDV high vater level scram trip bypass vould be enabled. (Closure of a separate bypass svitch for each
'channel vould be .required to complete the bypass.
4.4 Steamline Low Pressure Isolation Tri B ass Contacts-IO .
26 42 58 (Note: These contacts are also normally closed in STARTUP and REZOEL. )
4.4.1 If any of the above contacts are open in the K03TDOWN position, the MSIV isolation-on-lov-steamline-pressure function vould be 'enabled. h MSIV isolation trip or half trip vould occur.
4.4.2 If any of the above contacts are closed in RUN, MSIV isolation on lov steamline pressure vould be bypassed.
4.5 Conclusions For multiple failures of mode switch contacts vhich are normally closed in the SHUTDOWN mode, the principal concerns are:
- b. The unannunciated bypass of the MSIV-isolation-on-low-steamline pressure function in the RUN mode.
V. Summa and Conclusions A. All failure modes for the mode switch contacts where contacts open that should be closed vould result in scrams or half scrams
, depending on the number of contacts that are open. At the same 1-0273 10
time, for conditions of operation vhere steamline pressure is low, isolation of the main steamlines vould occur.
P'.
In the "STARTUP", REFUEL", and "SHUTDOWN" positions of the mode svitch, closures of contacts that should be open (3,5,19,21, wpv/d 35,37,51,53)>result in a bypass of the IRM scram function in one or more of the RPS channels. Closure of contacts 11, 27, 28, 43, 44, 59, would result in raising, the setpoint of the normally setdown APRM high flux scram function from 15~ to 113$ in one or more of the RPS/NMS channels.
C. In. item B above, although the mode svitch failure,- (i.e., contacts closing), would not be immediately apparent to the;.plant operator, the failure would be detected during the veekly IRM"and APRM channel functional tests. If these'ests veie performed prior to the power increase and after transferring the mode switch,.to the "STARTUP"'
position, then the IRM channel functional tests vould detect the IRM .
failures because no half scram vould result. The proposed technical specification requirement vill be that the IRM channel functional test and the APRM channel functional test be performed vithin 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to startup, if it has not been perfozmed in the previous seven days. Meekly surveillance vould be required for the case vhereby the "hot standby" (STARTUP position) condition is maintained for long periods of time.
D. In the 'RUN" position of the mode svitch, closures of contacts 7, 23, 39, 55 vould result in the bypass of one or more RPS trip channels related to the MSIV closure scram functions. Closure of contacts 10 '6, 42, 58 would result in the bypass of one or more NSSSS trip channels related to the steamline low pressure isolation function. Concurrent vith the incorrect'ode switch contact closures, there would be annunciations that one or more of the RPS MSIV closure scram trip channels have been bypassed.
1-0273 11
ALSO'NCE A REFUELING CYCLE AFTER THE N3DE SWITCH IS PLACED IN THE STARTUP POSITION A CHANNEL FUNCTION TEST OF THE IN S HIGH FLUX TRIP WILL BE PERFORNED
Closure of contacts 9, 25, 41, 57 can bypass the "SHUTDOWN¹'ode scram function. If the contacts remain closed during and after transfer of the mode svitch to the "SHUTDON" position,'such closed contacts would not allow a scram to occur from positioning'f the svitch. That is, only a half scram or no scram vould re'suit. This fact would be immediately apparent to the operator. Mannual scraming of the plant is accomplished by depressing the manually scram switch. The ability to scram the plant from the mode iwitch is only a secondary effect, and one of several backup alternates to the scram pushbutton.
In the "SHUTDOWN" mode, closure of mode svitch contacts 29, 30, 31, 61, 62, 63 vould remove the normal rod vithdrawal block restriction associated vith this mode. This fact vould be apparent'o the operator because the window for the normal rod vithdrawal block annunciator vould be extinguished, and its change of state vould alert the operator. The manual positioning of rods is under strict procedural controls. The rod block positioning restrictions are only backups to those controls. hdditionally, the operator vould become avare of the situation via standard technical specification driection by verifying this rod block by attempting to vithdrav a second rod after the first one is vithdravn.
1-0273 12
NINE MILE POINT 2 FSAR VI. Evaluation of the Effects of Mode Switch Miso eration on Cha ter 15
~na ses.
The potential impacts of the effects of mode switch misoperation on the analyses of transients and accidents presented in .Chapter 15 were evaluated. The focus was on certain specific events because of previously expressed NRC concerns with those events or because the events might impact the limiting transients. These specific events were classified into two groups according to the consequences of mode switch misoperation.
- 1. ~Grou 1 The events in Group I include:
.a.. The abnormal startup of an idle recirculation loop..
- b. The failure of the recirculation flow controller with increasing flow.
- c. A rod drop accident.
~ \
These are events for which the con'cern is related to the bypass of the scram function of the intermediate range monitor (IRM) while the mode switch is in the "STARTUP," "REFUEL," or "SHUTDOWN" positions. This would also raise the scram setpoint of the average power-range monitor (APRM). from the 15K "startup" value to the 118K "run" value, which corresponds to the analytical limit of 121% used for the analyses of Chapter 15 transients and accidents.
None of the Chapter 15 analyses of the events in Group I takes credit. for either the IRM scram function or the APRM scram function with the setpoint setdown to the 15-to-25K level.
Events a-and b of Group I were analyzed from a "RUN"-mode power condition since the Chapter 15 analyses are initiated from about 56K power and 40K core flow. In the "RUN" mode, the IRM trips are bypassed and the APRM flux scram-setpoint is approximately 118K (121% analytical limit). The rod drop accident analysis was initiated from OX power, (50K rod density); consequently, the mode switch would be in the "STARTUP" position.
No impact would result from the misoperation of the mode switch in the "REFUEL" or "SHUTDOWN" modes.
- a. For the analysis of the abnormal recirculation-loop startup transient, no credit was taken for the flow reference in the scram for high neutron flux.. The high neutron flux setpoint of 121% was used. The Analysis of this event was initiated from a power level significantly in excess of where recirculation-loop startups would PCY: pes/1 lOBB-I 4/27/84
NINE MILE POINT 2 FSAR normally originate and corresponding to the mode switch in the "RUN" mode. At lower power levels, tge consequences of the event would be less severe; consequently, the impact of the mode switch misoperation on the analysis of this event is of no significant consequence.
The initiation of an abnormal recirculation-loop startup transient when the mode switch is in the "STARTUP" position would also be of no consequence since operating procedures would require the initial power level to be less than IS%. The resulting power increase probably would not cause a scram. If the resulting power level were in excess of technical specification requirements related to power, pressure, and core flow, the operator .
would take corrective action in accordance with those requirements.
- b. The Chapter 15 analyses of the recirculation flow-controller failure with increasing flow were initiated from a 55% power and 35.7X core flow conditions, with a 12ll flux:scram terminating the power excursion. Similar events originating from the startup power range of O.to 15K power would be of lesser consequence. Also, at this low power level, normal operating procedures would infer minimum pump speed with individual'oop operation. These operating conditions would lessen the effect of a single-loop .flow increase and would preclude the event of flow control failing with increasing flow on both loops.
ce The analysis in Chapter 15 of the rod drop event only takes credit for the 121% APRM trip and takes no credit for the IRM scram function. The event, as analyzed from the OX power level, is terminated by the Doppler effect arid is of significance only below about 2 to 3% power.
At high power levels, the rod drop would be less of a problem because of the influence of the resulting steam voids in the core on the local high reactivity.
- 2. ~Grou 2 The events in Group 2 include:
- a. The inadvertent closure of the main steam isolation valve.
- b. The loss of an auxiliary power transformer.
- c. The break of a main steam line outside the containment.
- d. The failure in the open position of the steam pressure controller.
PCY: pes/IIOBB-2 4/27/84
NINE HILE POINT 2 FSAR These are events for which the concern is either the bypass of the main steamline isolation function due to low steamlfne pressure by the nuclear steam supply shutoff system (NS ) in the "RUN" mode or the loss of the position scram function of the HSIVs in the "RUN" mode. Only the isolation fbnction that should result whenever the turbine-inlet steamline Pressure drops below the (analysis) setpoint level of approximately )20 psig is of concern. Ko other isolation functions of the NS are impacted by the potential mode switch misoperations.
- a. The analysis of the HSIV closure event in Chapter 15 does take credit for the scram initiated from limit switches of the HSIV while the mode switch is in the "RUN" mode.
Potential mode switch misoperation could cause this scram function to be bypassed while the mode. switch is in the position. However, this bypass would be 'RUN" annunciated in the control room. The operating procedures would require corrective action since the technical specification requirement that all four-channels for the HSIV-closure trip function be operable in the "RUN: mode would be violated. Depending upon the number of inoperable channels, the affected channels and at least one trip system of'the reactor protection system (RPS) would have to be placed in the tripped condition within one hour. If both RPS trip systems were affected, the plant would have to. be placed in the "STARTUP" condition'ithin 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
- b. The consequences of the auxiliary power, as analyzed in Chapter 15, are also not affected by any mode switch misoperation. The scram and isolation that occur at about 2 seconds (or later) are a direct result of the
'oss of power to the RPS motor-generator sets and the subsequent disconnection of all power to the loads on the RPS bus.
Ce The analysis of the main steamline break outside of the containment does not take credit either for the low steamline isolation signal that would probably result from low steamline pressure or for the scram from HSIV closure. In this analysis, the event is initiated at the Level 3 scram to start out with a minimum inventory. At about.0.5 seconds into the event the isolation is assumed to be initiated because of high steamline flow. Although this is not addressed in the analysis, a level-8 high-water turbine trip would be expected due to sudden depressurization.
PCY:pes/llOBB-3 4/27/84
NINE NILE POINT 2 FSAR
- d. Failure of the steam pressure controller in the open position would result in a level-8 high-water. level turbine trip, which would initiate a scram and a.
recirculation pump trip. Further depressurization would be limited to the capacity of the turbine bypa'ss. Since an annunciation in the control room would have alerted to the bypass of the isolation function, the operator would be prepared to actuate HSIV closure manually should this event occur.
Conclusions from these evaluations are that all misoperations of the mode switch are detectable. by one or more of the following means.
- a. The operator would be imnediately aware of a problem
'because of the annunciation of bypasses that should not
. exist for the given postion of the mode switch. All mode switch misoperations that, might impact the severity of consequences of transients and accidents analyzed in Chapter 15 are in this category; Hence, the probability.
of,a transient occuring be'fore the operator 'takes .
corrective action would be extremely low.
- b. The operator would be iranediately aware of a problem in the RPS because of scrams or half scrams, which are also annunciated.
- c. The remaining modes of mode switch misoperation would be detected during the weekly channel functional tests of the NMS channel inputs to the RPS. If these tests were performed prior to the power increase and after the transfer of the mode switch to the "STARTUP" position, the IRM channel functional tests would detect the failures because no half scram would result.
PCY:pes/110BB-4 4/27/84
Nine Mile Point Unit 2 FSAR QUESTION F421.28 (7.3)
Provide a detailed responseonse too the concerns addressed by IE Controls) issued to operating reactors MMare h 13 1980 Fo all safety-related-equ-e ip ment, which does not remain n emergency mode follow owi n g an ESF rese,t provide p adequate e chan g e of state o f eac h piece of equipment or p ro p osed corrective ac mons changes (e.g., equipment return in ng too its normal operational status).
RESPONSE
See revised Section 7.3.2.1.2.2 1.
A review of ESF system (E CCS RHR, NSSS) documents current M984 for Nine Mile Point - Unit an ESF system reset control action NMP2'utomati'c s
idance o f IE B depressurization system (
ll)'he i 8o-06 trois return eich ADS safety/relief valve to .its close d position. ,Th is deliberate e operator action clos'ing e reliefalves B22-F013 C, H, K, M, N R, and U to
- t. inadvertent reactor depressurization xs e tion to IE 80-06 compliance.
Besides this exception, no deviation from thee gu idance in-dicated in IE Bulletin 80-06 was found.
In addition, RCIC (not no considered an ESF) was reviewe iewed and found to be in conformance wiith the guidance of IE Bulletin 80-06.
Amendment 10 QEcR F421.28-1 April 1984
Nine Mile Point Unit 2 FSAR QUESTION F421.34 (7.4)
Section 7.4.1.4 of the FSAR provides information on the Remote Shutdown System (RSS). Attachment 1 provides the instrumentation and Control Systems Branch (ICSB) guidance for remote shutdown capability. The attachment provides guidance for meeting the requirements of GDC 19. Provide supplemental information to identify the extent that the design of the RSS at Nine Mile Point - Unit 2 conforms to the guidance provided in Attachment 1.
. Include the following information in your discussion using drawings as appropriate:
a) Design criteria for the -remote control station equipment including the 'transfer switches and separation requirements for redundant functions.
b) Discuss the separation arrangement between safety related and nonsafety-related instrumentation and controls on the auxiliary shutdown panel.
c) Location of transfer switches and the remote control stations.
. d) Description of isolation, 'eparation and transfer/override provisions. This should include the design basis for, preventing electrical interaction between the control room and remote shutdown equipment.
e) Description of the administrative and procedural control features to both restrict and to assure access, when necessary, to the displa'ys and controls located outside the control room.
f ) Description of any communication systems required to coordinate operator actions, including redundancy and separation.
g) Means for ensuring . that cold shutdown can be accomplished.
h) Description of control room annunciation of remote control or override status of devices under local control.
Discuss the proposed start-up test program to demonstrate remote shutdown capability in accordance with the guidance provided in R.G. 1.68.2.
Amendment, 10 QER F421-34-1 April 1984
Nine Mile Point Unit 2 FSAR j) Discuss the testing to be performed during plant operation to verify the capability of maintaining the plant in a safe shutdown condition from outside the control room.
k) Di scus s the equipment c1 assi fication using the guidelines contained in FSAR Table 3.2-1.
RESPONSE
The response to Items a, b, c, and d is as follows:
Unit, 2 compliance with the guidance contained in Attach-ment 1 is escribed below:
- 1. The remote shutdown panel (RSP) is designed to achieve and maintain hot shutdown in the event the control room is inaccessible. This is achieved by the use of redundant, safety grade instrumentation~
identical .(in. most all cases) to nonsafety-'related that used: in the ~~gj room. Additionally, some A'ontrol indicators and recorders are provided for operation use.
- 2. See, Item 1.
- 3. No jumping, rewiring, circuit disconnection, or manual action (in locations. other than the RSP) is used to achieve the desired shutdown condition.
- 4. The design of the RSP is such that cold shutdown is achieved using safety grade, redundant instrumentation.
- 5. Loss of offsite power will not negate shutdown capability since power is supplied by reliable safety grade power sources.
GU1C a 7 r
- 7. The acces the rem e shutdown room is controlled at all times and the transfer switches response to Item e are the lockable type. See below.
8 Design of the RSP is under evaluation for conformance to the requirements of Appendix R to
~
10CFRSO.
'I Amendment 10 Q&R F421.34-2 April 1984
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Nine Mile Point Unit 2 FSAR
- e. Access to the remote shutdown system controls will be controlled by the plant security system. Access will be restricted as with other vital areas. In the event of the failure of the card reader, access will be possible by use of metal keys. Additionally, key lock switches requiring the use of metal keys are utilized for panel transfer/override functions. These keys will be administratively controlled according to key control procedures and readily available to both the Control Room Operator and the Shift Supervisor in the event of a control room evacuation.
Use of communications to coordinate operator actions is not required as all functions to safely shut down the plant can be performed independently from either the control room or the remote shutdown system controls.
- g. A description of how cold shutdown is achieved utilizing the -remote shutdown panel is contained in Section 7.4.1.4.
- h. A description of control room annunciation when transfer switches are changed 'o contained in Section 7.4.1.4.
the emergency position is 1
A separate startup test procedure, as dyscribed in test abstracts, Table 14.2, will be performed to demonstrate the capability of the remote shutdown system to safely shut down the plant within guidelines provided by Reg.
Guide 1.68.2.
Instrumentation and controls associated with the remote shutdown system shall be calibrated and functionally verified by surveillance testing as required in technical specifications. Remote shutdown system design is such that functional. testing of systems verifies the operational capability to provide remote shutdown.
Additionally, periodic testing of the transfer switches will be performed to verify control functions.
- k. See revised Table 3.2-1.
Amendment 10 Q&R F421.34-3 April 1984
Nine Mile Point Unit 2 FSAR 421.34 ATTACHMENT 1 ICSB GUIDANCE FOR THE INTERPRETATION OF GENERAI DESIGN CRITERIA 19 CONCERNING RE UIREMENTS FOR REMOTE SHUTDOWN STATIONS A. BACKGROUND GDC 19 requires that equipment at appropriate locations outside the control room be provided to achieve a safe shutdown of the reactor. Recent reviews of remote shutdown station designs have demonstrated that, some designs cannot accommodate ~
a single failure in accordance with the guidance of SRP Section 7.4
{Interpregation of GDC-19). The following provides supplemental guidance for the implementation. of the requirements of GDC-19 concerning remote shutdown stations. Requirements for remote .shutdown capability following a fire are detailed in Appendix R to 10CFRSO:
It should be noted that although GDC 19 and Appendix R requirements are complementary, the potential'xists that modifications to bring a design into conformance with GDC 19 will violate Appendix R criteria and vice versa. For example, remote manual devices for a second division of instrumentation and controls added to satisfy single failure requirements would not be acceptable if the added devices were located in the same fire area as existing transfer switches in the redundant division. In addition, transfer switches added to isolate the remote shutdown equipment from the control room fire area would not be acceptable ESF actuation; unless this is done if they disable in accordance with Item B6 below. The acceptability of remote shutdown station designs given a fire is determined by the Auxiliary Systems Branch (ASB) as outlined in Section 9.S.1 of the SRP.
B. ICSB GUIDANCE To Meet GDC-19 As Inte reted In SRP Section 7.4
- 1) The design should provide redundant safety grade capability to achieve and maintain hot shutdown from a location or locations remote from the control room, assuming ,no fire damage to any
~
required systems and equipment and assuming no accident has occurred. The remote shutdown station equipment should be . capable of maintaining functional operability under all service conditions to (including
'GR postulated occur abnormal environments such as loss of ventilation), but need Amendment 10 F421.34-4 ~ April 1984
Nine Mile Point Unit 2 FSAR not be environmentally qualified for accident li conditions unless envi ronmental qua fication is required for reasons other than remote shutdown.
The remote shutdown station equipment, including indicators, should be seismically qualified.
- 2) Redundant instrumentation (indicators) should be provided to display to the operator(s) at the remote shutdown location(s) those parameters which are relied upon to achieve and verify that a safe shutdown condition has been attained.
- 3) Credit may be taken for manual actions (exclusive of continuous control) of systems from locations that are reasonably accessible from the Remote Shutdown Stations. Credit may not be taken for manual actions involving 5umpering, rewiring, or disconnecting circuits.
The design should provide redundant safety grade capability for. attaining subsequent cold ,shutdown through the use of suitable procedures.
- 5) Loss of offsite power should not negate shutdown capability from the remote shutdown stations. The design and procedures should be such that of control from the remote;shutdownfollowing'ctivation location, a loss of offsite power will not result in subsequent overloading of essential buses or the diesel generator.= Manual restoration of power to shutdown loads i~ acceptable provided that sufficient information is available such that can be performed in a safe manner.
it
- 6) .. The . design should be such that if manual transfer of control to the remote location(s) disables any automatic actuation of ESF equipment, this equipment can be manually placed in service from the remote shutdown station(s). Transfer to the remote location(s) should not change the operating status of equipment.
Where either access to the remote shutdown station( s) or the operation of equipment at the station(s) is dependen't upon the use of keys (e.g.,
key lock switches) access to these keys shall be administratively controlled and shall not be precluded by the event necessitating evacuation of the contxol room.
Amendment 10 QEcR F421.34-5 April 1984
Nine Mile Point Unit 2 FSAR
- 8) The design should comply with the requirements of Appendix R to 10 CFR 50.
Amendment 10 Q&R F421. 34-6 April 1984
Nine Mile Point Unit 2 FSAR QUESTION F421.36 (7.5)
The NRC staff has recently issued Revision 2 to Regulatory 1.12 Guide 1.97, "Instrumentation for Light-Mater-Cooled Nuclear 1.13 Power Plants to Assess Plant and Environs Conditions During 1.14 and Following an Accident" via Supplement 1 to NUREG-0737. 1.15 This Reg. Guide revi sion re flects a number of ma j or changes 1.16 in post-accident instrumentation. Supplement 1 to 1.18 NUREG-0737 includes specific Reg. Guide 1.97 implementation 1.19 requirements for plants in the operating license review 1.20 stage.
Provide a description of how the Nine Mile Point Unit 2 1.21 design conforms to the provisions of Reg. Guide 1.97, 1. 22 Revision 2. This description should be in the form of a 1.24 table that includes the following information for each Type 1.25 A, B, C, D, E variable shown in Regulatory Guide 1.97: 1.26 (1) instrument range 1.28 (2) environmental qualification (as stipulated in guide or 1.29 state criteria) 1. 30 (3) seismic qualification (as stipulated in guide or state l. 31 criteria) 1.32 (4) quality assurance (as stipulated in guide or state 1.33 criteria)
(5) redundancy and sensor(s) location(s) 1.34 (6) power supply (e.g., Class 1E, non-Class lE, battery 1.35 backed)
(7) location of display (e.g., control room board, SPDS, 1.36 chemical laboratory 1.37 Deviations from the guidance in Reg.'uide 1.97 should be 1. 39 explicitly shown, and supporting justification or 1. 40 alternatives should be presented. 1.41 RESPON 1.42 J<<Ar <<~d r<<<<, Ii LC f V ~ iJ<< ~P ~ e Q i/r
> Ia'f Amendment 9 QEcR F421.36-1 March 1984 chl217718fqr9ae 02/01/84 112
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LDCN GE124 Attachment I Page I of 2 NINE MILE POINT 2 FSAR QUESTION 421.37 If reactor controls and vital instruments derive power from (7.S) common electrical distribution. systems, the failure of such electrical distribution systems may result in an event requiring operator action concurrent with failure of important instrumentation upon which these operator actions should be based. IE Bulletin 79-27 addresses several concerns related to the above subject. You are requested to provide information and a discussion based on each IE Bulletin 79-27 concern. Also, you are to:
- 1) Confirm that all a.c. and d.c. instrument buses that could affect the ability to achieve a cold shutdown condition were reviewed. Identify these bus'es.
- 2) Confirm that all instrumentation and controls required by emergency shutdown procedures were considered in the review. Identify these instruments and controls at the system level of detail.
- 3) Confirm that clear, simple, unambiguous annunciation of loss of power is provided in the control room for each bus addressed in item I above. Identify any exceptions.
- 4) Confirm that the effect of loss of power to each load on each bus identified in item I above, including ability to reach cold shutdown, was considered in the review.
- 5) Confirm that the re-review of IE Circular No. 79-02 which
-is required by Action item 3 of Bulletin 79-27 was extended to include both Class lE and non-class lE inverter supplied instrument or control buses. Identify these buses or conform that they are included in the listing required by Item I above.
RESPONSE
A methodology has been developed for addressing concerns raised in IE Bul.letin 79-27. This methodology, applied on WNP2, has been reviewed and appr'ov'ed by the NRC. The same methodology will be used in performing the required study for the NMP2 project. The NMP2 study is currently scheduled for completion in the second quarter of 1985. The methodology provides for a systematic and comprehensive analysis to ensure that, in the event of a single power bus failure, sufficient control room indicators, instruments, and controls exist for 'the operators to achieve reactor cold shutdown. The following paragraphs outline the methodology to be used in addressing the concerns identified in IE Bulletin 79-27.
KC: pes/118B-6 I/24/84
LDCN GE124 Attachment 1 Page 2 of 2
- 1) Review the Class 1E and non-Class 1E buses, including inverters, supplying power to instrumentation and controls in systems used in attaining the cold shutdown condition. All buses that could affect the ability to achieve cold shutdown are identified. Existing plant operating procedures and procedures already developed for the event of certain 'power bus failures are used to ensure the identification of all potential power buses.
- 2) Identify the instrumentation and control devices connected to each identified power bus. Evaluate the effects of a power loss to each load, including the limiting effects on the ability to achieve cold shutdown.
- 3) Create "bus trees" denoting the bus hierarchy and cascading bus configuration of all buses that power instrumentation and controls used by the operator to achieve cold shutdown.
- 4) Determine the annunciators and alarms that would alert the operators to a failure of any of the identified buses.
- 5) Determine the effect of any single power bus loss on the ability to continue in the particular shutdown path being used at the time the bus loss occurs. This analysis includes the cascading effects of any bus los's and considers alternative indications and controls powered by unaffected buses that may aid the operator in the event of a bus loss. Identify alternative shutdown paths available in the event of a bus loss and existing procedures for restoration of the affected bus.
- 6) Document the results of the analysis with recomendations of hardware or procedural changes.
KC:pes/118B-6.1 1/24/84
NINE MILE POINT 2 FSAR UESTION. 421. 42 The transient and accident analyses included in the FSAR are intended to demonstrate the adequacy of safety systems in mitigating anticipated operational occurrences and accidents.
Based on the conservative assumptions made in defining these "design bases" events and the detailed review of the analyses by the staff, it is likely that they adequately bound the consequences of single control system failures. To provide assurance that the design basis event analy-sis for Nine Mile Point 2 adequately bounds other more fundamental credi-ble failures, provide the following:
(1) Identify those control systems whose failure of malfunction could seriously impact plant safety..
(2) Indicat'e which, if any, of the control systems identified in (1) common power sources. The power sources whose receive power from failure or malfunction could lead to failure or malfunction of more than one control system and should extend to the effects of cascading power losses due to the failure of higher level distribution panels and load centers.
(3) Indicate which, if any, of the control systems idehtified in (1) receive input signals from common sensors. The sensors considered should include common taps, hydraulic headers and impulse lines feeding pressure, temperature, level or other signals to two or more control systems.
(4) Provide justification that any malfunctions of the control systems identified in (2) and (3) resulting from failures or malfunctions of the applicable common power source or sensor including hydraulic components are bounded by the analyses in Chapter 15 and would not require action or response beyond the capability of operators or safety systems.
RESPONSE
Two system interaction studies, common power failure analysis and common sensor failure or sensing line analysis, are required to address the issues of this question. The methodology to be applied for these analyses for NHP-2 has already been approved by the NRC for Grand Gulf, Shoreham and WNP-2 analyses.
The studies will evaluate the consequences of single power source or sensing line failures on control grade systems and determine whether the limiting case events are bounded by Chapter 15 analyses. These studies are scheduled for completion second quarter of 1985.
BAM:rf:csc/G08077"-1 8/14/84
A. Common Power Source Failure The following paragraphs outline the methodology to be used in the common power source failure analysis.
- 1) Identify all non-safety control grade systems that could affect the critical reactor parameters, i.e., water level, pressure, and power.
- 2) Review these control systems at the component level. Identify the effect of the loss of power to each system component and subsequent interactions with other components and systems.
- 3) Generate "bus trees" which represent the bus hierarchy and cascading configuration of all power buses that supply components of control systems under .study.
Perform a combined effects analysis. Evaluate the failure of each power bus, i.e., load center, motor control center, etc. starting with the lowest level source common to multiple control systems and.
working up the "bus trees" to the highest common power level. At each level, examine the effects of the single bus failure and conse-quential cascading bus failures on all control'ystems'omponents affected.
- 5) Postulate 'the limiting transient events as a result of the combined effect analysis. Compare these events to those analyzed in Chapter 15.
- 7) Identify any hardware or procedural change recommendations that result from this study.
B. Common Sensor or. Sensing Line Failure The following paragraphs outline the methodology to be used in the common sensor or sensing line failure analysis.
Identify all non-safety control grade systems that could affect the critical reactor parameters, i.e., water level, pressure, and power.
- 2) Identify all instrument sensing lines and sensors common to two or more of these control systems.
- 3) Analyze the effects of a failure of a common sensors, a complete plug, or a guillotine break in each of these common instrument lines.
Examine the effects of the erroneous signals on each affected instru-ment and on each function, i.e., scrams, trips, permissives, etc.,
actuated or rendered inoperative.
BAH:rf:csc/G08077"-2 8/14/84
- 4) Examine the interactive effects among.all systems affected by the common sensing line failure and the consequential combined effect on the critical reactor parameters.
J
- 5) Compare the consequences of these postulated events with the Chapter 15 analyses to ensure that Chapter 15 bounds the effects and the events would not require action or responses beyond the capability of operators or safety systems. Perform any additional transient calculations or analyses necessary to determine whether the postulat-ed limiting event: are bounded" by those events analyzed in Chapter 15, assuming there is a single active failure in a safety system required to mitigate effects of the event.
- 6) Identify any hardware or procedural change recommendations that result from this study.
- The term "bounded" means within the consequence limits for abnormal operational transients given in Section 15.0.3.1.2 of the FSAR or, if the combined proba-of occurrence of both the initiating event annJ the single active failure 'ility is similar to the occurrence probabilities of limiting faults (see Section 15.0.3.1), "bounded" means within the consequence limits for limiting the faults given in Section 15,0.3.1.3.
BAM:rf:csc/G08077"-3 8/14/84
NINE NILE POINT 2 FSAR QUESTION 421.43 If control systems are exposed to the environment resulting from the rupture of reactor coolant lines, steam lines, or feedwater lines, the control systems may malfunction in a manner which would cause consequences to be more severe than assumed in safety analyses. IEE Information Notice 79-22 discusses certain non-safety grade or control equipment, which if subjected to the adverse environment of a high energy line break, could impact the safety analyses and the adequacy of the protection functions performed by the safety-related, systems.
The staff is concerned that a similar potential may exist at light water facilities now under construction. You are, therefore, requested to perform a review per the IKE Information Notice 79-22 concern to determine what,
'esign changes or operator actions would be necessary to if any, assure that high energy line breaks will not cause control system failures to complicate the 'event beyond the FSAR analyses. Provide the results of your review including all identified problems and the manner in which you have resolved them.
The specific "scenarios" discussed in the above referenced Information Notice are to be considered as examples of the kinds of'nteractions which might occur. Your review should consider analogous interactions as relevant to the BWR design.
RESPONSE
KC: pes/118B-12
.I/24/84
HIGH. ENERGY LINE BREAK AND CONTROL SYSTEM FAILURE EVALUATION INTRODUCTION IE Information Notice 79-22 identifies the concern that the performance of nonsafety grade equipment subjected to an adverse environme'nt could impact the protective functions performed by safety grade equipment. The purpose of this analysis is to determine if a malfunction of a nonsafety control system, associated with a high energy line break, might result in a severe event not bounded by FSAR Chapter 15.
METHODOLOGY The HELB/control system failure evaluation will be analyzed as follows:
- 1. Identify all nonsafety control systems and components within these systems which may impact critical reactor parameters (water level, pressure, power).
- 2. Establish the criteria for energy lines, break postulation, and consequence evaluation.
- 3. Identify critical nonsafety grade components located in areas of high energy piping..
- 4. Postulate breaks in these areas and determine the resultant effects on the components.
Evaluate the events to determine if the event is bounded by FSAR Chapter 15. If not bounded, additional analysis or a cor-rective action will be taken.
NONSAFETY CONTROL SYSTEMS All plant nonsafety'. control systems are included in the initial evalua-tion for HELB. The following criteria is used for the elimination of systems from the initial list prior to performing a detailed HELB analysis.
- 1. Dedicated inputs into the process computer, as well as the com-puter itself.
- 2. Control systems which have no direct or indirect interaction with reactor operating parameters. Examples are communica-tions, lighting, ventilation for exterior buildings, machine shop systems, refueling or'maintenance systems, etc.
- 3. Control systems that do interact or interface with reactor operating systems, but which cannot affect the reactor param-eters either directly or indirectly.
C3/l2177/367/5YH
- 4. Electrical systems, the loss of which will result in a condi-tion similar to- total or partial loss of offsite power. Ex-amples include the station transformers, ac instrument power, and dc instrument power.
- 5. Systems which are not used during normal power operation. For example, refueling systems, turning gear, and turbine bearing lift pumps.
- 6. Safety systems or safety portions of control systems.
- 7. Mechanical and structural type systems. Examples include structural steel, turbines, cranes, etc.
All control components, including power sources, within systems not eliminated by the above criteria are evaluated for component elimination by the following criteria prior to the final HELB analysis.
Instruments which provide only indication or position status information are excluded from the detailed analysis.
- 2. Components which provide passive inputs into the control logic, examples of which are arming-type permissives which require additional manual action to command equipment to operate, are excluded from the detailed analysis.
- 3. Instruments and other dedicated inputs to the process computer are excluded from the detailed analysis.
- 4. Position switches on air- and motor-operated valves which are not interlocked with other equipment but rather provide posi-tion indication or position status to the process computer are excluded from the detailed analysis.
- 5. Mechanical type components, such as structural steel, tanks, and pipes are not considered "components" which can fail. How-ever, associated instruments, taps, tubing, and control compo-nents not eliminated by Items 1 through 4 and physically located on the above mechanical components, are evaluated.
PIPE BREAK CRITERIA The pipe break criteria is taken directly from FSAR Section 3.6.
Pi e Criteria High energy piping is defined as including those systems or portions of systems in which the maximum operating temperature exceeds 200 F or the maximum operating pressure exceeds 275 psig during normal full power operation. Those lines that operate above these limits for only a relatively short period of time (less than 2 percent) to perform their intended func-tion, are classified as moderate energy and excluded from con-sideration.
C3/12177/367/5YH
- 2. Break Postulation High energy pipes are assumed to break only at terminal ends and at each intermediate pipe fitting or weld attachment. Each longitudinal or circumferential break in high energy fluid svs-tem piping is considered separately as a single postulated initial event occurring during normal plant conditions.
- 3. Conse uence Evaluation Pipe breaks are evaluated for the effects of pipe whip, jet impingement, and environmental effects.
- a. ~Pi e Whi Pipe whip is assumed to occur in the plane defined by the piping geometry and to cause movement in the direction of the jet reaction.
- b. Jet Im in ement Jet impingement loads are determined by taking the jet force as being constant at all effective distances from, and normal to, the break area and by assuming that the jet stream diverges conically at a solid angle of 20 degrees.
ANALYSIS
- l. Utilizing current plant drawings, the nonsafety control components and high energy systems will be located in particular zones'.
In small zones it will be assumed that any HELB would incapacitate all nonsafety control components in the zone.
- 3. In large zones the effect of a high energy line break on each compo-nent will be evaluated based upon the pipe criteria.
- 4. Postulate breaks and evaluate the effects on the controls equipment.
- 5. Compare postulated effects with events as reported in CESAR Chapter 15 to determine if they are bounded.
- 6. If not bounded, determine if protection or relocation of the con-trols equipment is appropriate. me. Ol '0 If requiredr ) additional analysi/te determine if the effect is significant and then a corrective action will be taken.
- 8. Draft final report.
C3/12177/367/5YH
NINE MILE POINT 2 FSAR QUESTION 421. 44 A. Table 7.1-1 of the FSAR lists the safety-related (7. 7) instrumentation and control systems. Nonsafety-related systems are identified in Table 7.7-1. From a review of Chapter 15 of the FSAR the staff has deter-mined that the analysis of certain anticipated opera-tional occurrences (i.e., the feedwater controller failure-maximum demand) and design basis accidents (i.e., recirculation pump seizure) take credit for the operation of nonsafety-related instrumentation and control systems. It is the staff's position that for events classified as anticipated operational occurrences, credit can be taken for nonsafety-related systems to mitigate the event provided only high availability nonsafety-related systems are being relied upon.
Therefore, identify each instrumentation and control system/component which is not classified as safety-related but assumed in the FSAR analyses to mitigate the consequences of transients. Provide a justifica-tion for the assumption of operability of'his equip-
~
ment based upon'system design, equipment quality, and proposed technical specifications. In addition,
=
provide a discussion on the interfaces with the safety-related portions of the Nine Nile Point-Unit 2 design.
B. It is the staff's position that no credit may be taken for nonsafety-.related instrumentation and control systems/components in mitigating the consequences of design bases accidents. Therefore, identify each instrumentation and control system/component which is classified as nonsafety-related but assumed in the FSAR analyses to mitigate the consequences of accidents.
Either redo the analysis assuming no credit for the operation of this equipment, or propose modifications to upgrade the equipment to safety-related status.
RESPONSE
A. The following non-safety grade systems/components may be actuated during the course of anticipated operational occurrences (transients) shown in Chapter 15:
- a. Level 8 turbine trip,
- b. Level 8 feedwater trip,
- c. turbine bypass,
- d. recirculation runback,
- e. rod sequence control system,
- f. rod block monitor, and
- g. relief function of the safety relief valves.
PCY: pc: r f/L01242"-1 8/14/84
NINE NILE POINT 2 FSAR None of these systems are required to mitigate the events analyzed in Chapter 15.
The assumed performance of .the nonsafety grade system listed in Table 440.43-1, in response to guestion 440.43-1, is based on extensive. failure rate data for equipment of similar design and quality requirements. Among the nonsafety grade systems listed in Table 440.43-1, the failures of the L8 trip and turbine bypass would have the most adverse effects on hCPR. For the most limiting transient, a feedwater controller failure at maximum demand, the estimated increase in hCPR is about 0.02 for L8 trip failure and 0.11 for turbine bypass failure. Since this postulated event is 2 to 3 seconds in duration, no fuel failure is expected.
In summary, certain transient events assume the operation of specific nonsafety-grade equipment to provide a realistic transient signature; however, failures of such equipment would still yield events bounded by the safety limits of transient and limiting fault events analyzed in Chapters 6 and 15. The peak vessel pressure is bounded by the over pressure protection analysis described in Chapter 5.
See FSAR Section 7.7.1.5 and 10.4.4 for turbine bypass and Section 7.7.1.3 for L8 IKC design information.
B. FHEA's take no credit for nonsafety-related instrumentation and control system/components in mitigating'he consequences of design basis accidents.
PCY: pc: r f/L01242"-2 8/14/84
NINE HILE POINT UNIT 2 FSAR UESTION
.4 From its review of the Nine Mile Unit 2 FSAR, the NRC staff has been unable to conclude that the separation of Class lE components and interconnecting cir-cuits is acceptable. Regulatory Guide 1.75, "Physical Independence of Electri-cal Systems" which endorses IEEE Standard 384, "IEEE Trial-Use Standard Criteria for Separation of Class 1E Equipment and Circuits" provides guidance with regard to separation. To provide the level of detail necessary to com-plete our review, we'request that you submit a comparison of the Nine Hile Point Unit 2 design to the criteria contained in R.G. 1.75 and IEEE 384. This comparison should focus on the instrumentation and control systems ~ithin both the Power Generation. Control Complex and the balance of plant. The information provided should supplement FSAR Table 1.8-1 and FSAR Sections 7.1.2, 7.2.6 and 8.3. 1 such that each regulatory position of R.G. 1.75 and each separation criterion of .IEEE 384 is addressed. Alternate methods of providing separation
~
to those contained in R.G., 1.75 and IEEE 384 should be identified and justi-fied. Where barriers (e.g., flexible conduit, sheet metal enclosures, fire retardant tape) are used to provide separation, the details of the testing used to qualify the barriers should be provided. Where analyses have been used to justify lesser separation than that recommended in- R.G. 1.75 and IEEE 384, a detailed discussion of the analyses including the assumptions,, methods, sup-porting tests and conclusions should be provided.
RESPONSE
A comparison of the Nine Nile Point Unit 2 design to the criteria contained in R.G. 1.75 and IEEE 384 is shown in the following tables for instrumentation and control systems within the PGCC and balance of plant.
Note: Balance of plant response provided by SWEC.
BAM:rm/407274"-1 7/27/84
NIHE NILE POINT UNIT 2 FSAR SEPARATION EVALUATION QUESTION 421.47 REGULATORY GUIOE 1.75, REV. 1 DES IGH COHFORHAHCE IEEE 384-74 CRITERIA REGULATORY POSITION PGCC OEFIHITIOHS Isolation device. A device in a circuit C.1 Since interrupting devices (fuses . Interrupting devices actuated which prevents aalfunctions in one 3uppleaent IEEE 384 definition .and/or circuit breakers) actuated only by fault current are not section of a circuit froa causing as follows "interrupting devices only by fault current are not used as isolation devices for unacceptable influences in other actuated only by fault current considered as isolation devices isolating non-class IE circuits sections of the circuit or other are not considered to be isola- a coablnatlon of two interrupting froa class-IE circuits. In the circuits. tion devices within the context devices or an EPA in conJunction case of control/instr~nt of this docusent. with. an interrupting device is used. associated circuits, fuse/
breakers actuated by fault current have been used to isolate non-class IE and devices.
Raceway. Any channel that is designed C.2 Interlocked armor cable is not used Neets this requlreaent.
and used expressly for supporting wires, Mn erlocked araor enclosing cable as a raceway.
cable, or busbars. Raceways consist should not be, construed as a prlaarlly of, but are not restricted "raceway".
to, cable trays, conduits, and inter locked araor enclosing cable.
CRITERIA 4.l Re uired Se aration. Separation No coment. Separation is provided to aaintaln the Heats this requireaent.
shall e prov e o aa ntain .the indepen- )Odependence of sufficient nuaber of dence of sufficient nmaber of circuits circuits and equipient r'equired for and equipaent so that the protective Independence is'iotective'function.
functions required during and following achieved through equipaent arrangeaent, any design basis event can be accoa- aaterlals, wiring practices and isola-
- plished. The degree of separation tion devices and/or space or by required varies with the potential analysis.
hazards in a particular area.
BAH: cal: na/A08305"-I 8/30/84
. NINE HILE UNIT 2 FSAR SEPARATIOH EVALUATION QUESTION 421.47 (Continued)
REGULATORY GUIDE 1.75, REV. 1 DESIGN CONFORHANCE IEEE 384-74 CRITERIA REGULATORY POSITION PGCC BOP 4.2 E ui nt and Circuits Re uirin No couaent. Equipaent hand circuits requiring Meets this requireeent.
Se ara on. qu peen an c rcu s separation are delineated in the plant requ r ng separation shall be deterained design documents and identified in a and delineated early in the plant design distinctive Ianner.
and shall be identified on doclssants and drawings in a distinctive Ianner.
4.3 Hethods of Se aration. The separa- C.3 The separation of circuits and equipaent Meets this requireaent.
tion o c rcu s an equ paent shall be whenever practicable and where its is achieved by locating'hea in achieved by safety class structures, use does not conflict with other. separate safety class structures, distance, or barriers, or any coebina- safety ob)ectives, locate redundant distance or barriers, or any tion .thereof. circuits and equipaent in separate 'oabinatfon thereof or by analysis.
safety class structures.
4.4 C atibilit with Hechanical No coint. Class 1E circuits are routed and/or Meets this requireaent.
~S steas; e separa on o ass lE protected such that failure of related
+c rcuuEs and equipaent shall be such aechanical equipaent of one Class 1E that the required independence vill not systeo vill not disable Class 1E be coeproalsed by the failure of circuits or equipaent essential to the aechanical systeas served by. the operation of its redundant systea(s).
Class lE systees. For exaeple, Class lE circuits shall be routed or protected such that failure of related aechanicai
. equipaent of one redundant systee cannot disable Class lE circuits or equlpeent essential to the operation of the other redundant systea(s).
BAH:cal:rN/A08305*-2 8/30/84
HIKE MILE UNIT 2 FSAR SEPARATIOH EVALUATIOH (UESTIOH 421.47 (Continued)
REGULATORY GUIOE 1.75, REV. 1 DESIGN COHFORMAHCE IEEE 384-74 CRITERIA REGULATORY POSITION PGCC BOP 4.5 Associated Circuits. Associated . C.4 and C.6 Associated circuits are either sub)ect Associated circuits are treated circu s s a coap y w th one of the Ksisoc s e clecults should be to all requlreaents placed on Class 1E as class lE circuits.
following: subject to all requireaents placed circuits or are analyzed to desonstrate (1) They shall, be uniquely identified on Class 1E circuits such as cable that the associated circuits will not as such and shall reiain with, or be der ating, environaental qualifica- degrade the Class 1E circuits below an separated the same as, those Class lE tion, flaae retardance, splicing acceptable level. Such an analysis, circuits with which they are associated. restrictions and raceway fill when perforsedb is aainta<ned as part (2) They shall be in accordance with unless it can be deaonstrated that of the design record./k'ote 5.
(1) above froa the Class 1E equipeent the absence of such requireaents to and including an isolation device. could not significantly reduce the Beyond the isolation device a circuit availability of the Class 1E is not subject to the requireients circuits.
of this docuaent provided it does not again becoae associated with a Class lE Analysis should be subaitted as.
systea. part of Safety Analysis Report, (3) They shall be analyzed or, tested and should identify those circuits to demonstrate that Class 1E circuits installed in accordance with this are not degraded below an acceptable section.
level.
4.6 HOH-CLASS lE CIRCUITS 4.6.1 Se aration frog Class lE Circuits. Ho coasent. Kon-Class 1E circuits coaply with the Meets this requireaent except in Hon-Class c rcu s s a separa e requireaents of IEEE 384 Section 5.1.3, the case of non-class 1E conduits frow Class lE circuits by the ainlaua 5.1.4 or 5.6 or they are treated as proxiaate to class lE trays. The separation requireaents specified in associated circuits. alnlaea distance between a non-Sections 5.1.3, 5.1.4, or 5.6 or they class lE conduit and a class lE becoie associated circuits. cable tray (open) shall be 1 inch.
See FSAR Section 1.8 RG 1.75 h
position for justification.
BAM: cal: ra/A08305e-3 8/30/84 se e
HINE HILE UHIT 2 FSAR SEPARATIOH EVALUATION QUESTION 421.47 (Continued)
REGULATORY GUIOE 1.75d REV. 1 DESIGN CONFORMANCE IEEE 384-74 CRITERIA REGULATORY POSITION PGCC BOP 4.6.2 Se aration from Associated C.6 Hon-Class lE circuits are separated Meets this requirement.
Circuits. on- ass c rcu s shall analysis performed in accordance from'Class 1E and associated circuits tse separated iron associated circuits with this section should be in accordance with the requirements by the Iiniaua separation requirements submitted as part of Safety of IEEE-384, Sections 5.1.3, 5.1.4, specified in Sections 5.1,3, 5. 1.4, Analysis Report and should identify or 5.6.2 or effects of lesser separa-or 5.6.2 or (1) the effects of lesser those circuits installed in accor tion are analyzed to demonstrate that separation between the Hon-Class lE dance with this section. Class lE circuits are not degraded circuits and the associated circuits below an acceptable level. Such shall be analyzed to demonstrate that C.7 ana]ysis, when performed, is a part .
Class 1E ~c rcu s are not degraded . Rona-Class 1E instrumentation and of the design record. Hon-Class lE below an acceptable level or (2) they control circuits should not be instrumentation and control circuits become associated circuits. Non-Class exempted from the provisions of are not exempted from the provisions ld instrtssentatton and contro~c rcu ~ Section 4.6.2. of Section 4.6.2.
are no re u re o e separa e rom assoc a e c rcuits.
Figure 1 shows examples of acceptable
'ircuit arrangements.
- 5. SPECIFIC SEPARATION CRITERIA
>~
bfu/SID 5.1 Cables and Racewa s C.B Separation of Class lE circuits and Generally, different quip-5.1. enera Section 5.1.1.1 should not be equipment makes effective use of such aent are located in d ffe 5.1.1~ie routing of Class lE construed to imply that adequate features as different safety structures rooms; different a les circuits and location of equipment sewed separation of redundant circuits , and separated areas for redundant are routed through different by these Class lE circuits shall be can be achieved within a confined circuits and equipment. A degree of areas; separate tunnels are used reviewed for exposure to potential space such as a cable tunnel that separation cosxsensurate with the damage for routing cables of different hazards such as high pressure piping, is effectively unventilated. potential of the hazard is provided missiles, flasoable material flooding, and wiring that is not flaae retardant.
such that the independence of the redundant Class 1E systems is main-y I p/gro&S A degree of separation coamensurate tained at an acceptable level.
with the damage potential of the hazard shall be provided such that the indepen-dence of redundant Class lE systems are BAH:cal:rm/A08305*-4 8/30/84
. HIHE NILE P NIT 2 FSAR SEPARATIOH EVALUATION (UESTIOH 421. 47 .(Continued)
REGULATORY GUIDE 1.75, REV. 1 DESIGN COHFORHANCE IEEE 384-74 CRITERIA REGULATORY POSITION PGCC BOP maintained at an acceptable level. The (See also FSAR, Section 8.3.1.4.2 for (See FSAR Section 8.3.1.4.2 separation of Class 1E circuits and exception.) for details.)
equipment shall make effective use of features inherent in the plant design such as using different rooms or opposite sides of rooms or areas.
5.1.1.2 In those areas where the C.6 Cable installations within PGCC have ~
Analysis has been used in the
. damage potential is limited to failures Analysis performed in accordance been analyzed for separation case noted in Section 4.6.1.
or faults internal to the electrical with this section should be The analysis is included in equipment or circuits, the minimum submitted as part of the Safety Section 1.8, RG 1,75 position.
separation distance can be established Analysis Report, and should adequacy.'pen by analysis of the proposed cable identify circuits installed in installation. This analysis shall be accordance with these section.
based on tests performed to determine the flame retardant characteristics of the proposed cable installation consider-ing features such as cable insulation and jacket materials, cable tray fill, and cable tray arrangement.
5.1.1.3 The minimum separation C.9 ventilated cable trays and Heets the criteria 1, 2 and 3 of distances specified in Sections 5.1.3 TRh section should be. supplemented cable splices are not used. Section 5.1.1.3 where the minimum and 5.1.4 are based on open ventiliated with Item 5.$ .1.3(4) as follows: separation distances of Sections cable trays of either the ladder or "Cable splices in raceways should 5.1.3 and 5.1.4 are used; cables trough type as defined in NEHA VE 1-1971, be prohibited". splices in raceways are prohibited; Cable Tray Systems. 10ere these splicing in electrical penetra-distances are used to provide adequate NOTE: Cable splices are not, by tions is considered to be exempt physical separation: themselves, unacceptable. If they from this requirement.
(1) Cables and raceways involved shall exist,-the resulting design should be flame retardant be justified by analysis. The (2) The design basis shall be that the analysis should be submitted as cable trays will not be filled above part of the Safety Analysis Report.
the side rails BAH:cal:rm/A08305*-5 8/30/84
HIHE HILE UHIT 2 FSAR SEPARATI EVALUATIOH QUESTION 421.47 (Continued)
REGULATORY GUIOE 1.75, REV. 1 OESIGN CONFORWNCE IEEE 384-74 CRITERIA REGULATORY POSITION PGCC
,(3) Hazards shall be li<<ited to failures or faults internal to the electric equip-
<<ent or cables.
If less separation distances are Used they shall be established as in Section 5.1.1.2.
5.1.2 exceed 15 Identification.
per<<anent <<armer at intervals not to and exiting lE raceways ft and
'f Exposed Class lE C.10 at points of entry to fro<< enclosed areas. Class points" should be understood to
<<ean at intervals not to exceed 5 ft throughout thy entire cable shall be <<arked prior to the length. Also the preferred <<ethod Heets the requirement. Heets the require<<ants except that the cables are <<arked at intervals of 15 ft. instead of 5 ft.; see FSAR Section 1.8, RG 1.75 position for the explanations. See FSAR Sec.
installation of their cables. of <<arking cable is color coding. 8.3.1.3 for the details of Cables installed in these raceways shall be <<arked in a <<armer of sufficient durability and at a sufficient nu<<ber
'.ll TQs section should be supple<<anted the <<ethods of used.
identification of points to fac a e n a ver i- as follows: "The <<ethod of identi-cation that the installation is in con- : fication used should be si<<pie and for<<ance with the separation criteria. should preclude the need tp consult any reference <<aterial to distin-These cable <<arkings shall be applied prior to or during installation. 'uish between Class 1E and Hon-Class lE cables shall be identified by Class lE circuits, between Hon-e per<<anent <<arker at each end in accor Class lE circuits associated with dance with tKeeeslgn drawings or cable different redundant Class IE schedule. syste<<s, and between redundant The <<ethod of identification used to Class lE syste<<s.
<<eet the above require<<ents shall readily distinguish between redundant Class lE syste<<s and between Class 1E and non-Class 1E syste<<s.
. BAN:cal:r<</A08305*-6 8/30/84
NIHE NILE HIT 2 FSAR SEPARATIO VALUATIOH.
l2l.l7 (Continued) 'UESTION REGULATORY GUIOE 1.75, REV. 1 DESIGH COKFORNAKCE IEEE 384-7I CRITERIA REGULATORY POSITIOH PGCC 5.1.3 Cable S readin Area The cable C.12 Pending issuance of other Cables feeding power to control Neets the requireaents; separate spreading area s e space s) adjacent acceptance criteria, those portions and instrueentatlon circuits are not cable riser areas are used for to the aain control rooa where instru- of Section 5.1.3 (exclusive of the required to run in conduit within the redundant circuits. Power aantation and control cables converge NOTE following the second para- PGCC. These cables are treated cables in the riser areas are prior to entering the control, terrain. graph) that paralt the routing of as control and instruwentation cables routed in enclosed raceways.
ation, or instruaent panels. The cable power cables through the cable and are run in PGCC ducts along with See FSAR Sec. 8.3. l.i.2.
spreading area shall not contain high spreading area(s) and by iepli- other cables in the saaa division.
energy equipeent such as switch-gear, cation, the control rooa, should
'ransforwers, rotating equipment, or not be construed as acceptable.
potential sources of aissiles or pipe Also, Section 5.1.3 should be whip and shall not be used for storing suppleaented as follows: "1&ere flaaeable aaterials. Circuits in the feasible, redundant cable spread-,
cable spreading area should be lie(ted ing areas should be utilized."
to control and instrmentation functions and those power supply circuits and facilities serving the control room and instruaent systems. Power supply feeders to instruwent and control rooa distri-bution panels shall be installed in enclosed raceways that qualify as barriers.
Other power circuits that are required to traverse this area Shall be assigned to a ainiam nuaber of routes consistent with their separation requireaents and allocated solely for these power circuits.
Such traversing power circuits shall be separated froa other circuits in this area by 4 alnimum distance of 3 ft. and barriers.
BAH:cal;rw/A08305"-7 8/30/Bi
NINE MILE UNIT 2 FSAR..
EVALUATION .'EPARATION QUESTION 421.47 (Continued)
'I REGULATORY GUIOE 1.75, REV. 1 OESIGN CONFORMANCE IEEE 384-74 CRITERIA REGULATORY POSITION PGCC BOP NOTE: An acceptable alternative rout1ng
.for such traversing power circuits would be to route thea in iabedded conduit or in a separate enclosure designed as safety class structure (for exaaple, a concrete duct bank or other suitable enclosure) which in effect reaoves thea froa the area defined as the cable spreading area.
The ainiaua separation distance between redundant Class 1E cable trays shall be deterained by Section 5.1.1.2 or, where the conditions of Section 5.1.1.3 a'e aet, shall be 1 ft. between trays separated horizontally and 3 ft. between trays separated vertically.
NOTE: Horizontal separation is aeasured froa the side rail of one tray to the side rail of the ad)acent tray. Vertical separation 1s aeasured froa the bottoa of .
the top tray to the top of the side rail of the bottoa tray. (See also Section 5.1.4).
@here tera1nation arrangeaents preclude C.13 aaintaining the ainiaua separation dis- Ro significance should be See notes 2, 3, 4, 6, 7, 9 and 10. Mhere the ainiaua separation tance, the redundant circuits shall be attached to the different tray requireaents cannot be aet, run in enclosed .raceways that qualify widths illustrated in Figure 2. Sil-teap tape aay be used as a as barrters or other barriers shall be separation barrier.
provided between redundant circuits.
The ainiaua distance between these
.BAM: cal: ra/A08305e-8 8/30/84
NINE NILE UNIT 2, FSAR 5EPARATIO EVALUATIOH (UESTIOH 421.47 (Continued)
REGULATORY GUIDE 1.75'EV. 1 DESIGN CONFORHANCE IEEE 384-74 CRITERIA REGULATORY POSITION PGCC redundant enclosed raceways and between barriers and raceways shall be in 1 in.
Figs. 2, 3, 4 and 5 illustrate examples acceptable'rrangeaants of of barriers and enclosed raceways where the ainimea separation distance cannot be <<aintalned.
5.1.4 General Plant Areas. In plant Hot applicable. Heats the requireeents.
as siss)les, external fires, and pipe whip are excluded, the ainiam separation distance between redundant cable trays shall be determined by Section 5.1.1.2 or, where the conditions of Section 5.1.1.3 are set, shall be 3 ft. between trays separated horizontally and 5 ft. between trays separated verticially. If, in addition, high energy electric equipment such as switchgear, transforwers, and rotating equipment is excluded and power cables are installed in enclosed raceways that qualify as barriers, or there are no power cables, the ainimua separation distance aay be as specified in Section 5.1.3.
libera plant arrangewents preclude aain-taining the ainiaua separation distance, the redundant circuits shall be run in enclosed raceways that qualify as barriers or other barriers shall be provided between redundant circuits.
The ainiaw distance between these
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HIHE NIL UNIT '2- FSAR SEPARATION EVALUATION'- '".
QUESTION 421,47 (Continued)
REGULATORY GUIDE 1.75p REV. 1 DESIGN COHFORNANCE IEEE 384-74 CRITERIA REGULATORY POSITION PGCC redundant enclosed raceways and between barr iers and raceways shall be in 1 in.
Figs.'2, 3, 4 and 5 illustrate exaeples of acceptable arrangeaents of barriers and enclosed raceways where the Iiniam separation distance cannot be aaintained.
'.2 STAHOBY POWER SUPPLY 5.2.1 Standb Generatin Units. Redun- C.14 Hot Applicable Neets this requireaent.
dant C ass s nd y genera ng units Section 5.2.1 should be supple- See Section 8.3.1.1.2.
shall be placed in separate safety class mented as follows: And should structures. have independent air supplies."
5.2.2 Auxiliaries and Local Controls. Heats this requireeent.
The aux ar as an oca contro s or See Section 8.3.1.1.2.
redundant standby generating units shall be located in the saae safety class structure as the unit they serve or be physically separated in accordance with the requlreaents of Section 4.
5.3 DC SYSTEN 5.3.1 Batteries. Redundant Class lE C. 15 Hot Applicable Heets this requireaent. See battert'ee~ s a be placed ie separate Qfiere ventilation is required, the Section 8.3.1.4.2.
safety 'class structur es. separate safety class structures required by Section 5.3.1 should be served by independent ventilation systeas.
5.3.2 Batte Char er s. Battery No coswent Not Applicable Heets this requireaent. See chargers or re un ant Class 1E Section 8.3.1.4.2.
batteries shall be physically separated in accordance with the requireients of Section 4.
BAN: cal: ra/A08305"-10 8/30/84
NIKE HIL T UNIT 2 FSAR SEPARATIOH EVALUATION QUESTION 421.47 (Continued)
REGULATORY GUIOE 1.75, REV. 1. OESIGH CONFORHANCE IEEE 384-74 CRITERIA REGULATORY POSITION PGCC BOP 5.4 DISTRIBUTION SYSl'EH Section'8.3.1.4.2.
5.4.1 Switch ear. Redundant Class 1E Heats this requiresent. See distribut on sw chgrear groups shall Section 8.3.1.4.2.
be physically separated in accordance with the requireeents of Section 4.
5.4.2 Hotor Control Centers. Redun- Neets this requireeent. See dant Class no or contro centers Section 8.3.1.4.2.
shall be physically separated in accordance with the requlreaents of Section 4.
5.4.3 Oistribution Panels. Redundant Heets this requirenent.
Class 1 s r ut on panels shall be See Section 8.3.1.4.2.
physically separated in accordance with the requirements of Section 4.
5.5 COHTAIHHEHT ELECTRICAL PEHETRA- No coeaent Hot Applicable Heats this requireaent. See TIOHS. e un an ass con a naent Section 8.3.1.4.2.
eiecErical penetrations shall be physically separated in accordance with the require" nants of Section 4. Coapliance vith Section 4 vill generally require that redundant penetrations be widely dis-persed around the circuaference of the containaent. The ainiaua physical sep-aration for redundant penetrations shall neet the requireaents for cables and raceways given in Section 5.1.4.
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HINE NIL UHIT 2 FSAR 5EPARAT EVALUATION QUESTION 421.47 (Continued)
REGULATORY GUIOE 1.75, REV. 1 OESIGN CONFORHAHCE IEEE 384-74 CRITERIA REGULATORY POSITION PGCC BOP Hon-Class lE circuits routed in pene-trations containing Class lE circuits shall be treated as associated circuits in accordance with the requlreaents of I Section 4.5. II P
BAN: cal: ns/A08305e-12 8/30/84
NINE HILE PO FSAR SEPARATION EVALUATION QUESTION 421.47 REGULATORY GUIOE 1.75 REV. 1 OESIGN CONFORNANCE IEEE 384-74 CRITERIA REGULATORY POSITIOH PGCC BOP 5.6 Control Switchboards Ho coszaent e See BOP ~ Heets the requ)reaents. The oca on an Arran eaent. ~a)n control switchboards (PGCC)
Ka)n contro sw c oar s s a e are located in the control located in a control rooa within a building which is a se)sa)c safety class structure. The control Cat. I structure. The aa)n rooa shall protect frea and 'shall not control rooa does not contain contain high energy switchgear, trans- any high energy equ)paent.
foraers, rotating equipuent, or poten- See Section 3.8.
tial sources of aissiles or pipe whip.
Local control switchboards shall be ~ Hot applicable. ~ Heets this requ)resent.
located so that hazards such as fires, a)ssiles, vibration, pipe wh)p, and water sprays shall not cause failures comon to redundant Class IE functions.
Separation of redundant Class 1E ~ Controls for redundant Class 1E equip- ~ Heets this requ)resent.
equ)paent and circuits aay be achieved aent are located on separate control by locating thea on separate control panels. However, due to operational switchboards physically separated in considerations, some of the redundant accordance with the requirements of Class IE controls, are located on the Section 4. '&ere operational consider saae control panel. These iteas are ations dictate that redundant Class IE provided with adequate separation to equipaent be located on a single con- aeet the single failure criteria.
trol switchboard the requ)resents of Sections 5.6.2, 5.6.3, 5.6.4, and 5.6.6 shall apply.
BAN;N/A08302*-1 8/30/84
NINE HILE P FSAR SEPARATION ATION QUESTION 421. 47 REGULATORY GUIDE 1.75 REV. 1 DESIGN CONFORHANCE IEEE 384"74 CRITERIA REGULATORY POSITION PGCC BOP 5.6.2 Internal Se aration. The ~ No comment. ~ The minimum separation distance between ~ Heets this requirement.
'inimum separa on s ance etween redundant Class 1E equipment and wiring redundant Class lE equipment and wiring inside the control panels is main-internal to the control switchboards tained at 6". Due to the circuit con-can be established by analysis of the proposed installation. Th1s analysis figuration, if 6" is not achievable, alternate scans are used to justify shall be based on tests performed to lesser degree of separation, such as determine the flame retardant charac- metallic barriers, enclosures, conduits, teristics of the wiring, wiring mate- isolation devices and/or analysis.
rials, equipment, and other'materials See notes 2, 3, 4, 5, 7, 9, and 10.
internal to the. control switchboard.
lkere the control switchboard mate-rials are flam retardant and analysis is not performed, the minimum separa-tion distance shall be 6in. In the event the above separation distances are not maintained, barriers shall be installed between redundant Class. 1E equipment and wiring.
5.6.3 Internal Mirin Identifica- ~ No coaaent. ~ Class IE wires and cables internal to ~ Heets this requirement.
tion. Class w>re un es or ane} are identified to distinguish
'cCa Tes internal to the control boards etween redundant Class lE and shall be identified in a distinct non-Class IE wiring. See notes 5 and permanent manner at a sufficient and 8.
number of points to readily distinguish between redundant Class lE wiring and between Class 1E and non-Class lE wiring.
5.6.4 Comon Terminations. lthere ~ No coaaent. ~ Cemon terminations within the control ~ Heets this requirement.
.'anels meet the provisions of IEEE-384 on coaaon device, the provisions of ..=,para 5,6.2. 'See not'e 1, Section 5.6.2 shall be met.
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NINE HILE 2 FSAR SEPARATION EVALUATION QUESTION 421.47 REGULATORY GUIOE 1.75 REV. 1 OESIGN CONFORHANCE IEEE 384-74 CRITERIA REGULATORY POSITION ~ PGCC BOP 5.6.5 Non-Class 1E Wirin . ~ No coaaent. ~ Non-Class lE wiring within the control ~ Heets this requireaent.
Non-Class w r)ng no separated froa panels is separated froa the redundant Class 1E wiring by the ainiaua separa- Class 1E wiring. Wherever tion distance (deterained in the non-Class 1E wiring cannot be Section 5.6.2) or by a barrier shall separated froa the Class 1E wiring, be treated as associated circuits in (1) it is treated as associated wiring, accordance with the requirements of or (2) an analysis is per fonaed to deaon-Section 4.5. strate the adequacy of lesser separa-tion, or (3) proper isolation (barrier or comon device) is provided to achieve the required separation.
5.6.6. Cable Entrance. Redundant ~ No coaaent. ~ Redundant Class lE cables entering the ~ Heets this requirement control .board enclosure are (a) separated board enclosure shall eeet the require- by a siniaum distance of six inches or a aent of Section 5.1.3. barrier or (b) enclosed in a raceway.
See notes 2 and 9.
5.7 Instreaentation Cabinets. C. 16 The first paragraph of ~ Redundant Class lE instruments are ~ Heets this requireaent.
Redun an ass ns roaents shall Section 5.7 should be auyaented located in separate cabinets or separate be located in separate cabinets or as follows: "The separation re- coipartaents of a cabinet. If redun-coapartaents of a cabinet. Where quireaents of 5.6 a~ply to instru- dant instrqments are reguired to be redundant Class 1E instruaents are aentation cabinets. located on a single cab>net or single located in separate coepartaents of single coapartjsent, barriers are pro-a single cabinet, attention aust be vided. Cables entering such cabinets iven to routing of external cables are separated by a ainiaua distance; or, o the instruaents to assure that barriers are provided between redundant cable separation is retained. coaponents and wiring.
In locating Class 1E instruaent ~ See BOP response. ~ Heets this requireaent.
cabinets attention aust be given to the effects of all pertinent design basis events.
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NINE 'NILE 2 FSAR SEPARATIO LUATTON QUESTION 421.47 REGULATORY 4UIOE 1.75 REV. 1 DESIGN COHFORNAHCE IEEE 384-74 CRITERIA REGULATORY POSITION =
PGCC BOP 5.8 Sensors and Sensor to Process ~ Ko comment. ~ Sufficient number of redundant sensors ~ Neets this requirement.
Connec sons. e un an ass sen- are provided to perform system level sors anraietr connections to the safety function, Adequate separation is process system shall be sufficiently maintained between required number of re-separated that functional capability dundant sensors to maintain the func-of the protection system will be main- tional capability of the protection sys-tained despite any single design basis tem. Neutron monitoring sensor cables event or result therefrom.. Considera- under the vessel are exempt from this tion shall be given to secondary criterion because of the space limita-effects of design basis events such as tions.
pipe whip, steam release, radiation, missiles, or. flooding.
Large components such as the reactor vessel can be considered a suitable barrier if the sensor to process con-necting lines are brought out at widely divergent points and routed so as to keep component between redundant lines.
Redundant pressure taps located on opposite'sides of a large pipe may be considered to be separated by the pipe, but the lines leaving the taps must be protected against damage from a credible coloon cause unless other redundant or diverse instresentation is provided.
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n .s, BAH: rm/A08302~-4 8/30/84
HIHE NILE FSAR SEPARATIOH E ALUATIOH QUESTION 421.47 REGULATORY GUIDE 1.75 REV. 1 DESIGH COHFORHAHCE IEEE 384-74 CRITERIA REGULATORY POS IT IOH PGCC BOP 5.9 Actuated E ui ent. Locations ~ No cosssenti ~ Redundant Class lE actuated equipments ~ Heets this requireaent.
of C ass ac ua e equipaent, such are separated to eeet the single failure as puap drive aotors and valve criteria and assure sufficient safety operating aotors are noraally dic- function to aitigate a OBE; tated by the location of the driven equipeent. The resultant locations of this equipment aust be reviewed to ensure that separation of redundant Class 1E actuated equipment is acceptable.
Hon-Class lE circuits not separated by 6 inches from Class 1E or associated circuits have been analyzed to deaonstrate the adequacy of lesser separation. The iteas analyzed are:
- 1. Cemon devices such as relays and con actors for Cl ss IE/Clasp 1E nd lass 1E/non'-Class 1E interfaces. Test report and analyses are availab e in GE 0 ign Record Files. ~
5'F'4
- 2. Sil-Teay tape as a separation barr er. he tes report and analysis is available in GE ORFs.
- 3. Use of flexible or rigid conduigs a separation barrier. The test report and analysis is available in GE DRFs.
Justification of separation less than six inches between saoke detector, its wiring and Class 1E wiring is available in GE DRFs.
Coeaon devices are also covered under the response to guestion 421.13. Sil-Teep.tape and flexible conduit are covered under the response to guestion 430.23.
- 5. NS panels P606, P608 and P633 are exceptions to RG 1.75.
- 6. Justification of running bare cable along with a conduit is available in GE ORF's.
- 7. Justification of separation of less than 1" between redundant enclosed raceways and between barriers u
sa Wraceways are available in GE DRF's.
- 8. Pre~ired vendor equipaent that does not sect color coding is identified in GE specification.
- 9. Use of cable connector housing as an acceptable separation barrier is available in GE ORF's.
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NINE NILE P FSAR SEPARATION E LUATION QUESTION 421. 47 REGULATORY GUIDE 1.75 REV. 1 DESIGN CONFORHANCE IEEE 384"74 CRITERIA REGULATORY POSITION PGCC BOP
- 10. Justification of separation of less than six inches between utility devices and its wiring and Class 1E siring is available in GE DRF's.
- 11. All analyses/)ustification for any exceptions are documented in GE DRF's and are available on request.
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