ML18037A157
| ML18037A157 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 02/15/1979 |
| From: | Eric Thomas LEBOEUF, LAMB, LEIBY & MACRAE |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML17053A421 | List: |
| References | |
| NUDOCS 7902200001 | |
| Download: ML18037A157 (90) | |
Text
REGULATOR~ INFORMATION DISTRIBUTION 5 STEM (BI DS)
ACCESSION NBR<790220000l DOC.DATF-'9/Q2/15 NOTAI ED< NO DOCKET 0
~ Fr~CILAO-220 Nine Mile Point Nuclear Stations, Niagara Mohawk Powe 05000220
'UTH. NAME AUTH()B AFFILIATION LeBoeuf, Lamb, Leiby 8, MacBae
'HOMAS,E.B.
RFCIP.NAME DENTON,H.B.
RECIPIENT AFFILIATION Office of Nuclear Reactor Regulation 0 SUBJECTS Forwards application for amend to License DPR-63 8, copy of proposed changes to Tech Specs.I'I/encl license amend fee.
DISTRIBUTION CODE < AOOI S COPIES RECEIVED-LTR Q ENCL g~ SIZE~~ME+
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LLON A ALLEN>JR CAMKRON F. MACRAE RANDALI J. L5$ 0EUF> JR. I929 l97$
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JOHN B.CHASE HARVEY A. NAPIKR TFLKPHONC 2I2 ~ 2$ 9 II00 ROGER D.FCLDMAN 5 JAM CS 0 MAI.LEY,JR. 4 CABLE ADDRESS CUGENE R FIDCLL 4 J. MICHAEL PARISH JACOB FRIEDLANDCR WILLIAMW. ROSENBLATT LCSWIN>NCW YOAX GERARD GIORDANO JOHN A. RUDY TCLEXI 4234ld DONALD J.GREENE PATRICX J, SCOGNAMIGLIO JAMES A. GRCER>11 HAROLD M ~ SCIDKI.
JOHN L.GROSC HALCYON G ~ SKINNER 47 $ KRXCLCY SQUARC DOUGLAS W.HAWES JOSEPH S.STRAUSS CARL D.HOBELMAN SAMUCL M.SUGDCN LONDON WIX $ 08 ENGLAND MICHAEL IOVCNKO JAMES F JOHNSON, 47>> CUGENK 8. THOMAS, JR. >5 TCLCPHONC Ol 49I 409I RONALD D. JONES LEONARD M.TROSTEN 4 HARRY H ~ VOIGT 44 TCLKXI 25955 JAMCS A. LAPCNN LCX II~ LARSON > 4 H. RICHARD WACHTEL GRANT S. LEWIS GCRARD P. WATSON KIMBA W. I.OVEJOY THOMAS A. 2IERK RESIDENT PARTNERS WASHINGTON OFFICE
~ RESIDENT PARTNERS LONDON OFFICE 4 ADMITTED TO THE DISTRICT OF COLUMBIA BAR February 15, 1979 Mr. Harold R. Denton Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Re: Niagara Mohawk Power Corporation Nine Mile Point Nuclear Station, Unit No. 1 Docket No. 50-220
Dear Mr. Denton:
As counsel for Niagara Mohawk Power Corpora-tion, I enclose the following:
(1) Three (3) originals and nineteen (19) copies of an Application for Amendment to Operating License; and (2) Forty (40) copies each of three (3) documents designated Attachments A, B and C which set forth the requested change in the Technical Specifications along with its technical basis, and supporting information, which demonstrates that the proposed change does not in-volve a significant hazards considerarion, nor would authorize any change in the types or any increase in the amount of effluents or any change in the authorized power level of the facility.
IIOI gII V9022000 ~f
Mr. Harold R. Denton February 15, 1979 page two The proposed amendment to the Operating License has been evaluated and determined to fall within the de-finition of'lass III of 10 C.F.R. 5 170.22; therefore, a check in the amount of $ 4,000.00 is enclosed to cover the appropriate fee.
Very truly yours, LeBOEUF, LAMB, LEIBY a MacRAE By ugene B. Thomas, Jr.
Enclosures
ATTACHMENT A Niagara Mohawk Power Corporation License No. DPR-63 Docket No. 50-220 Proposed Changes to Facility Operating License (Appendix A)
Attached is a new Page 212a and revised Pages 6, 13, 188, 192-201, 203-215, 231, 232, 232a, 233 and 237a. The revised pages have been retyped in their entirety and the marginal markings indicate the specific changes to the text.
, rma."r mp. 50-220 ACCES: 7902200001
0 SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING
- c. The neutron flux shall not exceed its scram d.. The reactor water low level scram trip setting for longer than 1.5 seconds as indicated setting shall be no lower than -12 inches by the process computer. When the process (53 inches indicator scale) relative to computer is out of service, a safety limit the minimum normal water level (302'9").
if the violation shall be assumed neutron flux exceeds the scram setting and control rod e. The reactor water low-low level setting for scram does not occur. core spray initiation shall be no less than -5 feet (5 inches indicator scale)
To ensure that the Safety Limit established relative to the minimum normal water level in Specifications 2.l.la and 2.1.1b is not (Elevation 302'9").
exceeded, each required scram shall be initiated by,its expected scram signal. The Safety Limit f. The flow biased APRM rod block trip settings shall be assumed to be exceeded when scram is shall be less than or equal to that shown accomplished by a means other than the expected in Figure 2.1.1.
scram signal.
- d. Whenever the reactor is in the shutdown condition with irradiated fuel in the reactor vessel, the water level shall not be more'than 7 feet ll inches (3.88 inches indicator scale) below minimum normal water level (Elevation 302'9"), except as specified "e" below.
- e. For the purpose of performing major maintenance (not to exceed 12 weeks in duration) on the reactor vessel; the reactor water level may be lowered 9'elow the minimum normal water level (Elevation 302'9"). Whenever the reactor water level is to be lowered below the low-low-low level set point redundant instrumentation will be provided to monitor the reactor water level.
Amendment No , , 15
BASES FOR 2.1.1 FUEL CLADDING SAFETY LIMIT During periods when the reactor is shut down, consideration must also be given to water level requirements, due to the effect of decay heat. If reactor water level should drop below the top of the active fuel during this time, the ability to cool the core is reduced. This reduction in core cooling capability could 1'ead to elevated cladding temperatures and clad perforation. The core will be cooled sufficiently to prevent clad melting should the water level be reduced to two-thirds of the core height.
The lowest point at which the water level can normally be monitored is approximately 4 feet 8 inches above the top of the active fuel. This is the low-low-low water level trip point, which is 7 feet 11 inches (3.88 inches indicator scale) below minimum normal water level (Elevation 302'9"). The safety limit has been established here to provide a point which can be monitored and also can provide adequate margin. However, for performing major maintenance as specified in Specification 2. 1. l.e, redundant instrumentation will be provided for monitoring reactor water level below the low-low-low water level set point. (For example, by installing temporary instrument lines and reference pots to redundant level transmitters, so that the reactor water level may be monitored over the required range.) In addition written procedures, which identify all the valves which have the potential of
. lowering the water level inadvertently, are established to prevent their operation during the major maintenance which requires the water level to be below the low-low level set point.
The thermal power transient resulting when a scram is accomplished other than by the expected scram signal (e.g., scram from neutron flux following closure of the main turbine stop valves) does not necessarily cause fuel damage. However, for this specification a safety limit violation will be assumed when a scram is only accomplished by means of a backup feature of the plant design. The concept of not approaching a safety limit provided scram signals are operable is supported by the extensive plant safety analysis.
Amendment Ho. , 14 13
0 LIHITIHG CONOITION FOR OPERATION SURVEILLANCE RE(UIREMEHT 3.6.2 PROTECTIVE INSTRUMENTATION 4.6.2 PROTECTIVE INSTRUMENTATION A~li bill Applies to the operability of the plant Applies to the surveillance of the in-instrumentation that performs a safety strumentation that performs a safety function.. function.
~0b ective: ~0b ective:
To assure the operability of the instrumentation To verify the operability of protective required for safe operation. instrumentation.
- a. The set points, minimum number of trip a. Sensors and instrument channels shall systems, and minimum number of instrument be checked, tested and calibrated channels that must be operable for each at least as frequently as listed in position of the reactor mode switch shall Tables 4.6.2a to 4.6.2k.
be as given in Tables 3.6.2a to 3.6.2k.
If the requirements of a table are not met, the actions listed below for the respective type of instrumentation shall be taken.
(1) Instrumentation that initiates scram - control rods shall be inserted, unless there is no fuel in the reactor vessel.
188
C l~
Table 3.6.2a (cont'd)
INSTRUMENTATION THAT INITIATES SCRAM Limitin Condition for 0 eration Minimum No. of Minimum No. Operable Instrument Reactor Mode Switch of Tripped or Channels per Position in Which Operable Operable Function Must Be Parameter ~T Tri S stem Set Point 0 erable
~ CL O CJ a
4 dJ S-t5 CY CA CA (6) Main-Steam-Line 4(h) <10 percent (c) (c) X Isolation Valve valve closure Position from full open (7) High Radiation <5 times normal X -X X Hain-Steam-Line background at rated power (8) Shutdown Position (k) X X of Reactor Mode Switch (9) Neutron Flux (a) IRM (i ) Upscal e 3(d) <96 percent of (9) (g) (g)
. full scale 192
X Table 3.6.2a (cont'd)
INSTRUMENTATION THAT INITIATES SCRAM Limiting Condition for Operation Minimum No. of Minimum No. Operable Instrument Reactor Mode Switch of Tripped or Channels per Position in Which Operable Operable Function Must Be Parameter ~T Tri S stem Set Point 0 erable O
C/)
(ii) Inoperative 3(d) X X (b) APRM (i) Upscale 2 3(e) Figure 2.1.1 X X X (ii) Inoperative 2 3(e) X X X (iii) Downscale 2 3(e) >5 percent of (g) (g) (g) full scale (10) Turbine Stop <10$ valve Valve Closure closule (ll) Generator Load Rejection 193
i 0 Table 4.6.2a INSTRUMENTATION THAT INITIATES SCRAM Surveillance Re uirement Instrument Instrument Channel Parameter Sensor. Check Channel Test Calibration (I) Manual Scram None Once per 3 None months (2) High Reactor None Once per month Once pqr 3 Pressure months(>)
(3) High Drywell None Once per month Once pyr 3 Pressure months(1)
(4) Low Reactor Water Once/day Once per month Once pqr 3 Level monthstl)
(5) High Water Level None Once per month Once per 3 Scram Discharge months Volume (6) Main-Steam-Line None Once per 3 None Isolation Valve months Position (7) High Radiation Once/shift Once per week Once per 3 months Hain-Steam Line 194
Table 4.6.2a (cont'd)
INSTRUMENTATION THAT INITIATES SCRAM Surveillance Re uirement Instrument Instrument Channel Parameter Sensor Check Channel Test Calibration t (8) Shutdown Position None Once during each None of Reactor Mode major refueling Switch outage (9) Neutron Flux f (a) IRM (i) Upscale (f) (f) (f)
(ii) Inoperative (f) (f) (f)
(b) APRM (i) Upscale Hone Once/week Once/week (ii) Inoperative None Once/week Once/week (iii) Downscale Hone Once/week Once/week (10) Turbine Stop None Once per 3 months None Valve Closure' (ll) Generator Load Hone Once per month Once per 3 Rejection months 195
NOTES FOR TABLES 3.6.2a AND 4.6.2a (a) May be bypassed when necessary for containment inerting.
(b) May be bypassed in the refuel and shutdown positions of the reactor mode switch with a keylock switch.
(c) May be bypassed in the refuel and startup positions of the reactor mode switch when reactor pressure is less than 600 psi.
(d) Ho more than one of the four IRM inputs to each trip system shall be bypassed.
(e) Ho more than two C or D level LPRN inputs to an APRM shall be bypassed and only four LPRN inputs to an APRN shall be bypassed in order for the APRM to be considered operable. No more than one of the four APRN inputs to each trip system shall be bypassed provided that the APRN in the other instrument channel in the same core quadrant is not bypassed. A Travelling In-Core Probe (TIP) chamber may be used as a substitute APRN input if the TIP is positioned in close proximity to the failed LPRN it is replacing.
(f) Calibrate prior to starting and normal shutdown and thereafter check once per shift and test once per week until no longer required.
(g) IRM's are bypassed when APRM's are onscale. APRM downscale is bypassed when IRM's are onscale.
(h) Each of the four isolation valves has two limit switches. Each limit switch provides input to one of two instrument channels in a single trip system.
(i) May be bypassed when reactor power level is below 45K.
(j) Trip upon loss of oil pressure to the acceleration relay.
(k) May be bypassed when placing the reactor mode switch in the SHUTDOWN position and all control rods are fully inserted.
(1) Only the trip circuit will be calibrated and tested at the frequencies specified in Table 4.6.2a, the primary sensor will be calibrated and tested once per operating cycle.
196
1 V Table 3.6.2b INSTRUMENTATION THAT INITIATES PRIMARY COOLANT SYSTEM OR CONTAINMENT ISOLATION Limitin Condition for 0 eration Minimum No. of Minimum No. Operable Instrument Reactor Mode Switch of Tripped or Channels per Position in Which Operable Operable Function Must Be Parameter Set Point 0 erable a CL O Q S-dJ a$ LL CC t/l CA PRIMARY COOLANT ISOLATION
~Main Steam, Cleanup, and Shutdown)
(1) Low-Low Reactor Water Level >5 inches X X (Indicator Scale)
(2) Manual X X X X MAIN-STEAN-LINE ISOLATION (3) High Steam Flow Hain-Steam Line <105 psid X X Amendment No. , 14 197
I Table 3.6.2b (cont'd)
INSTRUMENTATION THAT INITIATES PRIMARY COOLANT SYSTEM OR CONTAINMENT ISOLATION Limitin Condi ti on for 0 erati on Minimum No. of Minimum No. Operable Instrument Reactor Mode Switch of Tripped or Channels per Position in Which Operable Operable Function Must Be Parameter Tri S stems Tri S stem Set Point 0 erable e
QJ CL O
QJ S-OC n5 CY l/) CA (4) High Radiation Hain-Steam Line <5 times X X normal back-ground at rated power (5) Low Reactor >850 psig Pressure (6) Low-Low-Low >7 in. mercury (a) X Condenser Vacuum vacuum (7) High Temperature <200F X X Main-Steam-Line Tunnel 198
Table 3.6.2b (cont'd)
INSTRUMENTATION THAT INITIATES PRIMARY C OL NT SYSTEM R CONTAINMENT ISOLATION Limitin Condition for 0 eration Minimum No. of Minimum No. Operable Instrument Reactor Mode Switch of Tripped or Channels per Position in Which Operable Operable Function Must Be Parameter Tri S stems Tri S stem Set Point 0 erable CL oo C S-n$ OC V) C/l CLEANUP SYSTEM ISOLATION (8) High Area Temperature <190 X X X X SHUTDOWN COOLING SYSTEM ISOLATION (9) High Area Temperature <170 X X X X CONTAINMENT ISOL T N (10) Low-Low >5 inches (c) X X Reactor Water (Indicator Scale) 199 Amendment No. P, 14
Table 3.6.2b (cont'd)
IHSTRUMENTATIOH THAT INITIATES PRIMARY COOLANT SYSTEM OR CONTAIHMEHT ISOLATION Limitin Condition for 0 eration Minimum No. of Minimum Ho. Operable Instrument Reactor Mode Switch of Tripped or Channels per Position in Which Operable Operable function Must Be Parameter Set Point 0 erable (11) High Drywell (3.5 psig (c) (b) (b)
Pressure (12) Manual X X X X 200
Table 4.6.2b INSTRUMENTATION THAT INITIATES PRIMARY COOLANT SYSTEM OR CONTAINMENT ISOLATION Surveillance Re uirement Instrument Instrument Channel Parameter Sensor Check Channel Test Calibration PRIMARY COOLANT ISOLATION
~Hain Steam, Cleanup and Shutdown)
(1) Low-Low Reactor Once/day Once per month . Once per 3 months (d) plater Level (2) Manual- Once during each major refueling outage MAIN-STEAM-LINE ISOLATION (d)
(3) High Steam Flow Once/day Once per month (d) Once per 3 months Main-Steam Line (4) High Radiation Once/shift Once/week Once per 3 months Main-Steam Line (5) Low Reactor Once/day Once per month Once per 3 months (d)
Pressure 201
Table 4.6.2b (cont'd)
INSTRUMENTATION THAT INITIATES PRIMARY COOLANT SYSTEM OR CONTAINMENT ISOLATION Surveillance Re uirement Instr ument Instrument Channel Parameter Sensor Check Channel Test Calibration CONTAINMENT ISOLATION (10) Low-Low Reactor Once/day Once per month Once per 3 months Hater Level (ll) High Drywell Once/day Once per month -Once per 3 months (d)
Pressure (12) Manual Once during each operating cycle 203
I NOTES FOR TABLES 3.6.2b AND 4.6.2b (a) Hay be bypassed in the refuel and startup positions of the reactor mode switch when reactor pressure is less than 600 psi.
(b) Hay be bypassed when necessary for containment inerting.
(c) Hay be bypassed in the shutdown mode whenever the reactor coolant system temperature is less than 215 F.
(d) Only the trip circuit will be calibrated and tested at the frequencies specified in Table 4.6.2b, the primary sensor will be calibrated and tested once per operating cycle.
204
Table 3.6.2c INSTRUMENTATION THAT INITIATES OR ISOLATES EMERGENCY COOLING Limitin Condition for 0 eration Minimum No. of Minimum No. Operable Instrument Reactor Mode Switch of Tripped or Channels per Position in Which Operable Operable Function Must Be Parameter Tri S stems Tri S stem Set-Point 0 erable CL O QJ 4 S-QJ A5 CY CY CA Cfl EMERGENCY COOLING INITIATION (1) High-Reactor Pressure 2. <1080 psig (b) X X (2) Low-Low Reactor Water Level > 5 inches (Indicator Scale) (b) X X EMERGENCY COOLING ISOLATION
~for each of two systems)
(3) High Steam Flow Emergency Cooling System 2 (a) <19 psid (b) X X (4) High Radiation Emergency Cooling System Vent <25 mr/hr (b) x x Amendment No. 205
, 14
Table 4.6.2c INSTRUMENTATION THAT INITIATES OR ISOLATES EMERGENCY COOLING Survei1 lance Re uirement Instrument Instrument Channel Parameter Sensor Check Channel. Test Calibration EMERGENCY COOLING INITIATION
~1 High Reactor None Once per month Once per 3 months Pressure (2) Low-Low Reactor llater Level Once/day Once pet month Once per 3 months EMERGENCY COOLING ISOLATION
~for each of two systems)
(3) High Steam Flow None Once per 3 months Once per 3 months Emergency Cooling System (4) High Radiation Once/shift Once during each Once during each Emergency Cooling major refueling major r efuel i ng System Vent outage outage 206
NOTES FOR TABLES 3.6.2c AND 4.6.2c (a) Each of two differential pressure switches provide inputs to one instrument channel in each trip system.
(b) May be bypassed in the cold shutdown condition.
(c) Only the trip circuit will be calibrated and tested at the frequencies specified in Table 4.6.2c, the primary sensor will be calibrated and tested once per operating cycle.
207 Amendment No. 14
I
- Table 3.6.2d INSTRUMENTATION THAT INITIATES CORE SPRAY Limitin Condition for 0 eration Minimum No. of Minimum No. Operable Instrument Reactor Mode Switch of Tripped or Channels per Position in Which Operable Operable Function Must Be Parameter Set-Point 0 erable oO C/l START CORE SPRAY PUMPS (I) High Drywel1 Pressure <3.5 psig (d) x, (a) (a)
(2) Low-Low Reactor Water Level >5 inches (b) X X X
-(Indicator Scale)
OPEN CORE SPRAY DISCHARGE VALVES (3) Reactor Pressure >365 psig X X X X and ei ther ( I) or (2) above.
208
J /
Table 4.6.2d INSTRUMENTATION THAT INITIATES CORE SPRAY Surveillance Re uirement Instrument Instrument Channel Parameter Sensor Check Channel Test Calibration START CORE SPRAY PUMPS (I) High Drywel 1 Once/day Once per month Once per 3 months Pressure (2) Low-Low Reactor Once/day Once per month Once per 3 months Water Level OPEN CORE SPRAY DISCHARGE VALVES (3) Reactor Pressure None Once per month Once per 3 months and either (I) or (2) above 209
I C' NOTES FOR TABLES 3.6.2d AND 4.6.2d (a) May be bypassed when necessary for containment inerting.
(b) Nay be bypassed when necessary for performing major maintenance as specified in Specification 2. 1.l.e.
(c) Only the trip circuit will be calibrated and tested at the frequencies specified in Table 4.6.2d, the primary sensor will be calibrated and tested once per operating cycle.
(d) Hay be bypassed when necessary for integrated leak rate testing.
(e) The instrumentation that initiates the Core Spray System is not required to there is no fuel in the reactor vessel.
be operable, if 210
F Table 3.6.2e INSTRUMENTATION THAT INITIATES CONTAINMENT SPRAY Limitin Condi tion for 0 erati on Minimum No. of Minimum No. Operabl e Instrument Reactor Mode Swi tch of Tripped or Channels per Position in Which Operable Operable Function Must Be Parameter 'Tri S stems Tri S stem Set-Point 0 erable a ~tU n. c D K M CY U W 5-crt C/)
Cfl (1) a. Hi gh Drywell < 3.5 psig (a) x Pressure (
and
- b. Low-Low Reactor Water Level > 5 inches (Indicator Scale) (a) X X 211
Table. 4.6.2e INSTRUMENTATION THAT INITIATES CONTAINMENT SPRAY Surveillance Re uirement Instrument Instrument Channel Parameter Sensor Check Channel Test Calibration (1)a. High Drywell Pressure Once/day Once 'nce per month (b) per 3 months
- b. Low-Low Reactor Once/day Once per month Once per 3 months Water Level 212
NOTES FOR TABLES 3.6.2e AND 4.6.2e (a) May be bypassed in the shutdown mode whenever the reactor coolant temperature is less than 215 F.
(b) Only the trip circuit will be calibrated and tested at the frequencies specified in Table 4.6.2e, the primary sensor will be calibrated and tested once per operating cycle.
212a
Table 3.6.2f INSTRUMENTATION THAT INITIATES AUTO DEPRESSURIZATION Limitin Condition for 0 eration Minimum No. of Minimum No. Operable Instrument- Reactor Mode Switch of Tripped or Channels per Position in Which Operable Operable Function Must Be Parameter Tri S stems Tri S stem Set-Point 0 erable o
CA INITIATION (1) a .. Low-Low-Low Reactor Water Level 2 (a) 2 (a) > 3 88 inches (b) (b) x (Indicator scale) and
- b. High Drywell 2 (a) 2 (a) < 3.5 psig (b) (b) x Pressure 213
Table 4.6.2f INSTRUMENTATION THAT INITIATES AUTO DEPRESSURIZATION Surveillance Re uirement Instrument Instrument Channel Parameter Sensor Check Channel Test Calibration INITIATION Low-Low-Low None Once per month Once per 3 months (c)
Reactor Water and
- b. High Drywell Once/day Once per month Once per 3 months Pressure 214
NOTES FOR TABLES 3.6.2f AND 4.6.2f (a) Both instrument channels in either trip system are required to be energized to initiate auto depressurization.
One trip system is powered from power board 102 and the other trip system from power board 103.
(b) May be bypassed when the reactor pressure is less than 110 psig and the reactor coolant temperature is less than the corresponding saturation temperature.
(c) Only the trip circuit will be calibrated and tested at the frequencies specified in Table 4.6.2f, the primary sensor will be calibrated and tested once per operating cycle.
215
Table 3.6.2k HIGH PRESSURE COOLANT INJECTION Limitin Condition for 0 eration Minimum No. of Minimum No. Operable Instrument Reactor Mode Switch of Tripped or Channels per Position in Which Operable Operable Function Must Be Parameter ~T Set-Point 0 erable CL O QJ O D +>
S CJ (ted OC C/) ll)
(1) Low Reactor Water Level > 53 inches (a) (a) X (Indicator scale)
(2) Automatic Turbine Trip (a) (a) X 231 Amendment No. , 14
Table 4.6.2k HIGH PRESSURE COOLANT INJECTION Surveillance Re uirement Instrument
'Instrument Channel Parameter Sensor Check Channel Test Calibration (1) Low Reactor Once per day Once per month Once per 3 months Water Level (2) Automatic None Once during each None Turbine Trip operating cycle 232
NOTES FOR TABLES 3.6.2k AND 4.6.2k (a) May be bypassed when the reactor pressure is less than 110 psig and the reactor coolant temperature is less than the corresponding saturation temperature.
(b) Only the trip circuit will be calibrated and tested at the frequencies specified in Table 4.6.2k, the primary sensor will be calibrated and tested once per operating cycle.
232a
BASES FOR 3.6.2 AND 4.6.2 PROTECTIVE INSTRUMENTATION The reactor protection system automatically initiates a reactor scram to prevent exceeding established limits.
In addition, other protective instrumentation is provided to initiate action which mitigates the consequences of accidents or terminates, operator error.
The reactor protection system is a dual channel type (Table 3.6.2.a). Each trip system except the manual scram has two independent instrument channels. Operation of either channel will trip the trip system, i.e.,
the trip logic of the channel is one-out-of-two. A simultaneous trip of both trip systems will cause a reactor scram, i.e., the tripping logic of the trip systems is two-out-of-two. The tripping logic of the total system is referred to as one-out-of-two taken twice. This system will accommodate any single failure and still perform its intended function and in addition, provide protection against spurious scrams. The it reliability of the dual channel system or probability that will perform its intended function is less than that of a one-out-of-two system and somewhat greater than that of a two-out-of-three system (Section VIII-A.1.0 of the FSAR).
The instrumentation used to initiate action other than scram is gener'ally similar to the reactor protection system. There are usually two trip systems required or available for each function. There are usually two instrument channels for each trip system. Either channel can trip the trip system but both trip systems are required to initiate the respective action. Where only one trip system is provided only one instrument channel is required to trip the trip system. All instrument channels except those for automatic depressurization are normally energized. De-energizing causes a trip. Power to the trip systems for each function is from reactor protection system buses ll and 12.
The signals for initiating automatic blowdown and rod block differ from other initiating signals in that only one of the two trip systems is required to start blowdown=or initiate rod block. Both instrument channels in the trip system must trip to initiate automatic blowdown. This difference is due to the requirement that automatic depressurization be prevented unless A.C. power is available to the emergency core cooling systems.
The instrument channels in the trip system for automatic depressurization are normally de-energized. In order to cause a trip both instrument channels must be energized. Power to energize the instrument channels is from power boards 102 and 103. If A.C. power is lost to one power board, one trip system becomes inoperable 233
BASES FOR 3.6.3 AND 4.6.2 PROTECTIVE INSTRUMENTATION
- b. The control rod block functions are provided to prevent excessive control rod withdrawal so that HCPR is maintained greater than 1.06. The. trip logic for this function is 1 out of n; e.g., any trip on one of the eight APRH's, eight IRH's or four SRM's will result in a rod block. The minimum instrument channel requirements provide sufficient instrumentation to assure the single failure criteria is met. The minimum instr'ument channel requirements for the rod block may be reduced by one for a short period of time to allow maintenance, testing, or calibration. This time period is only M% of the operating time in a month and does not significantly increase the risk of preventing an inadvertent control rod withdrawal.
The APRH rod block trip is flow biased and prevents a significant reduction in MCPR especially during operation at reduced flow. The APRM provides gross core protection; i.e., limits the gross core power increase from withdrawal of control rods in the normal withdrawal sequence.
The trips are set so that HCPR is maintained greater than 1.06.
The APRM rod block also provides local protection of the core; i.e., the prevention of critical heat flux in a local region of the core, for a single rod withdrawal error from a limiting control rod pattern. The trip point is flow biased. The worst case single control rod withdrawal error has been analyzed and the results show that with the specified trip settings rod with-drawal is blocked before the HCPR reaches 1.06, thus allowing adequate margin. Below ~604 power the worst case withdrawal of a single control rod results in a HCPR > 1.06 without rod block action, thus below this level it is not required.
The IRH rod block function provides local as well as gross core protection. The scaling arrange-ment is such that trip setting is less than a factor of 10 above the indicated level.
Analysis of the worst case accident results in rod block action before MCPR approaches 1.06.
A downscale indication on an APRH or IRM is an indication the instrument has failed or the instrument is not sensitive enough. In either case the instrument will not respond to changes in control rod motion and the control rod motion is prevented. The downscale rod blocks are set at 5 percent of full scale for IRH and 2 percent of full scale for APRM (APRH signal is generated by averaging the output signals from eight LPRM flux monitors).
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ATTACHMENT B Niagara Mohawk Power Corporation License No. DPR-63 Docket No. 50-220 Su ortin Information The purpose of these changes is to make corrections and changes in the Protective Instrumentation Tables. The reason for each change is provided below:
Pa e 6 & 13 The low-low-low reactor vessel water level set point has been changed from 127.1 inches to 3.88 inches (Indicator Scale). The reason for this change is that new instruments are being installed in accordance with recommended manufacturer's instructions with the high pressure side connected to the core spray inlet header and the low side connected to the condensing pot. In this way the indicator scale will be decreasing as the water level decreases. Although the indicator scale set point will be different, the actual water level set point elevation (7'l" below minimum normal water level) will be the same. Therefore, the same margin of safety will be maintained.
~Pe e 188 Currently, Specification 3.6.2a(1) requires the control rods to be inserted, when the instrumentation that initiates scram is inoperable. Since this instrumentation is not required to be operable, when there is no fuel in the reactor vessel; Specification 3.6.2a(1) is being revised to delete the requirement for inserting control rods.
~Pe e 192 The Condenser Low Vacuum scram was deleted, since this trip is not considered part of the Reactor Protection System. As indicated on Page 9 of the Additional Information section of the Sixth Supplement to the Final Safety Analysis Report, this trip is provided for the protection of equipment other than the reactor.
The remaining parameters on Page 192 have been renumbered.
The wording of the setpoint for the High Radiation Main-Steam-Line has been clarified to accurately describe the setpoint as < 5 times normal background at rated power.
I The footnote (k) has been placed in the REFUEL column for the Shutdown Position of the Reactor Mode Switch.
The footnote is explained with the changes on Page 196.
~Pa e 193 Renumbered the remaining parameters on Page 193 due to the deletion of the Condenser Low Vacuum scram.
Footnote (g) has been added to the "RUN" column of the APRM downscale scram. This footnote which appears on Page 196, indicates that the APRM downscale scram is bypassed in the "RUN" mode when the IRM's are onscale.
This footnote addition allows the APRM's downscale scram to be bypassed in any operating condition whenever the IRM's are onscale.
~Pa e 196 Added footnote (1) to the Instrument Channel Calibration column for parameters (2), (3) and (4).
This footnote is explained with the changes on Page 196.
The surveillance requirement for the Condenser Low Vacuum change has been deleted for the same reasons'as indicated on Page 192.
The remaining parameters on Page '194 have been renumbered due to the deletion of the Condenser Low Vacuum scram.
~Pa e 195 The parameters on Page 195 have been renumbered due to the deletion of the Condenser Low Vacuum scram.
Pacae 196 Footnote (h) has been revised to reflect the actual design of the main steam line isolation valve scram logic (i.e. each isolation valve has two limit switches and each limit switch provides input to one of two instrument channels in a single trip system).
Isolation valves 01-01 and 01-03 provide inputs into trip system 11 and isolation valves 01-02 and 01-04 provide inputs into trip system 12. This provides an increased margin of safety because each valve has two limit switches providing input to separate instrument channels in the same trip system. The original design consisted of one limit switch for each isolation valves, with each limit switch providing input to each of two instrument channels in a single trip system.
Footnote (k) has been added to allow bypassing the shutdown position of Reactor Mode Switch scram in the "REFUEL" mode when placing the switch in the "SHUTDOWN" position and it has been verified that all control rods are fully inserted. This change is requested to prevent unnecessary hydraulic shock to the Control Rod Drive System when switching to
4 4
SHUTDOWN from the REFUEL mode with all rods fully inserted.
Footnote (I) has been added to indicate that the trip
'ircuits for the reactor pressure indicators, drywell pressure indicators and reactor water level indicators will be calibrated once per three months, and tested once per month, while the primary sensors will be calibrated and tested once per operating cycle. This change is requested because the existing primary sensors for this instrumentation are being replaced with electronic sensors which are reliable and maintain their accuracy for longer periods of time than mechanical switches, as discussed in NEDO-21617-A. The new sensors will be located outside pr imary containment and have been environmentally qualified in accordance with the IEEE-323-1974. The environmental conditions for which this equipment has been qualified are given in Table 4-2.of NEDO-21617-A and meets the requirements for the maximum abnormal environmental conditions io )he Reactor Building Outside Primary Containment.<~
~Pa e 197 The requirement that primary coolant isolation on low-low reactor water level signal be operable in the SHUTDOWN and REFUEL modes has been deleted. This is consistent with Specification 3.2.7 which requires the reactor coolant system isolation capability to be operable only during power operating conditions, whenever the reactor head is on.
The requirement that main steam line isolation on high steam flow in the main steam line signal be operable in the SHUTDOWN and REFUEL mode has been deleted.
This is consistent with the requirements of Specification 3.2.7.
~Pa e 198 The wording of the setpoint for initiation of main steam line isolation on high radiation has been clarified to accurately describe the setpoint as < 5 times normal background at rated power. The requirement that main steam line isolation on high radiation-main steam line, low-Iow-low condenser vacuum, and high temperature main steam line tunnel signal be operable in the SHUTDOWN and REFUEL modes has been deleted. This is consistent with the requirements of Specification 3.2.7.
~Pa e 199 The requirement that containment isolation on low-low reactor water level signaI be operable in the REFUEL mode has been deleted. A new footnote (c) has been added to the SHUTDOWN column. This footnote which appears on. Page 204 allows bypassing the containment (I)Nine Mile Point Unit 1 Pipe Whip Analysis Transmitted June 29, 1973, P. D. Raymond to A. Giambusso.
e p isolation initiation in the SHUTDOWN mode whenever the reactor coolant system temperature is less than 215 F.
These changes are consistent with the requirements of Specification 3.3.4.
Pacae 200 The requirement that containment isolation on high drywell pressure signal be operable in the REFUEL mode has been deleted. Footnote (c) has been deleted from the SHUTDOWN and REFUEL columns for containment isolation on high drywell pressure. New footnote (c) has been added to the SHUTDOWN column. This footnote which appears on Page 204 allows bypassing the containment isolation initiation in the SHUTDOWN mode, whenever the reactor coolant system temperature is less than 215 F. These changes are consistent with the requirements of Specification 3.3.4.
~Pa e 201 Footnote (d) has been added to the Instrument Channel Calibration and Instrument Channel Test columns for parameters (1) and (5). The footnote is explained with the changes on Page 204.
Pacae 203 Footnote (d) has been added to the Instrument Channel Calibration and Instrument Channel Test columns for parameters (10) and (11). The footnote is explained with the changes on Page 204.
Pa<ac 204 Footnote (c) has been changed to allow bypassing containment isolation initiation in the SHUTDOWN mode, whenever the reactor coolant system temperature is less than 215 F.
,Footnote (d) has been added which indicates that only the trip circuit will be calibrated once per three months, and tested once per month, while the primary sensor will be calibrated and tested once per operating cycle. The reason for this change is the same as for footnote (1) on Page 196, that the electronic sensors replacing the existing primary sensors are more reliable and maintain their accuracy fot longer periods of time than mechanical switches.
~Pa e 205 Emergency Cooling initiation has been changed from High-High Reactor Pressure to High Reactor Pressure.
This is the correct initiating signal at the same setpoint of 1080 psig. Nine Mile Point Unit 1 presently does not have a high-high reactor pressure set point.
The old footnote (b) has been replaced by a new footnote (b) in the SHUTDOWN column for Emergency Cooling Initiation and Isolation. The requirement for the Emergency Cooling System Initiation and Isolation
4 to be operable in the REFUEL mode has been deleted.
The reason for these changes are explained with the changes on Page 207.
~Pa e 206 The Emergency Cooling initiation parameter has been changed to High. Reactor Pressure=for the same reason as indicated on Page 205.
Footnote (c) has been added to the Instrument Channel Calibration and Instrument Channel Test columns for parameters (1) and (2). The footnote is explained on Page 207.
~Pa e 207 The old footnote (b) has been replaced by a footnote which indicates that the -Emergency Cooling initiation can be bypassed in the cold shutdown condition. This is consistent with Specification 3.1.3 on Page 47 which indicates that the Emergency Cooling System is only required to be operable whenever tIIe reactor coolant temperature is greater than 212 F. Therefore, initiation of the Emergency Cooling System is not required in the COLD SHUTDOWN or REFUEL modes.
Footnote (c) has been added which indicates that only the trip circuit will be calibrated once per three months, and tested once per month, while the primary sensor will be calibrated and tested once per operating cycle. The reason for this change is the same as footnote (1) on Page 196, that the electronic sensors replacing the existing primary sensors are more, reliable and maintain their accuracy for longer periods of time than mechanical switches.
~Pa e 208 Footnote (d) has been added to the SHUTDOWN column for the start of the core spray pumps on High Drywell Pressure. The footnote is explained with the changes on Page 210.
Footnote (e) which appears on Page 210 has been added.
This footnote allows the instruments that initiate core spray to be inoperable, if there is no fuel in the reactor vessel. This change is consistent with Specification 3. 1.4 which requires the core spray system to be operable whenever there is irradiated fuel in the reactor vessel.
~Pa e 209 Footnote (c) has been added to the Instrument Channel Calibration and Instrument Channel Test columns for parameters (1), (2) and (3). The footnote is explained with the changes on Page 210.
~Pa e 210 Footnote (c) has been added to indicate that only the trip circuit will be calibrated once per three months,
1b and tested once per month, while the primary sensor will be calibrated and tested once per operating cycle. The reason for this change is the same as for footnote (1) on Page 196, that the electronic sensors replacing the existing primary sensors are more reliable and maintain their accuracy for longer periods of time than mechanical switches.
Footnote (d) has been added to indicate that the start of core spray pumps on High Drywell Pressure may be bypassed in the SHUTDOWN mode when necessary for integrated leak rate testing. The reason for this change is that integrated leak rate testing requires pressurizing the primary containment to 22 psig, well above the high drywell pressure scram set point.
Footnote (e) has been added as indicated in the changes for Page 208.
~pa e 211 Footnote (a) has been added to the SHUTDOWN column for parameters (l)a and b. The requirement for the Containment Spray System initiation to be operable in the REFUEL mode has been deleted. The reason for these changes are explained with the changes on Page 212a.
~pa e 212 Footnote (b) has been added to the Instrument Channel Calibration and Instrument Channel Test columns for parameters (1)a and b. The footnote is explained with the changes on Page 212a.
~pa e 212a Footnote (a) has been added which indicates that initiation of the Containment Spray System on high drywell pressure and low-low reactor water level may be bypassed in the SHUTDOWN mode wheIIever the reactor coolant temperature is less than 215 F. This change-=
is consistent with Specification 3.3.7 on Page 158 which indicates that the Containment Spray System is only required to be operable whenever t)e reactor coo'lant temperature is greater than 215 P. Therefore, initiation of Containment Spray System is not required in the COLD SHUTDOWN, or REFUEL mode.
Footnote (b) has been added which indicates that only the trip circuit will be calibrated once per three months, and tested once per month, while the primary sensors will be calibrated once per operating cycle.
The reason for this change is the same as for footnote (1) on Page 196, that the electronic sensors replacing the existing primary sensors are more reliable and maintain their accuracy for longer periods of time than mechanical switches.
ts 0
~Pa e 213 The set point for parameter (1)a has been changed to
> 3.88 inches (Indicator scale). The reason for this change is that these instruments will be installed so that the indicator scale will be decreasing as the water level decreases. Although the indicator scale set point will be different, the actual water level set point elevation (7'll" below minimum normal water level) will be the same. Therefore, the same. margin of safety will be maintained.
Footnote (b) has been added to the STARTUP and SHUTDOWN column for parameters (1)a and (b). The requirement that Auto Depressurization initiation be operable in the COLD SHUTDOWN and REFUEL conditions has been deleted. The reasons for these changes are discussed on Page 215 where footnote (b) is explained.
~Pa e 214 Footnote (c) has been added to the Instrument Channel Calibration and Instrument Channel Test columns for parameters (1)a and b. The footnote is explained with the changes on Page 215.
~Pa e 215 Footnote (b) has been added which indicates that Auto Depressurization initiation may be bypassed when the reactor pressure is less than 110 psig and the reactor coolant temperature is less than the corresponding saturation temperature. This is consistent with Specification 3.1.5 on Page 57, which indicates that the Auto Depressurization system is required to be operable whenever the reactor's coolant pressure is greater than 110 psig and the reactor coolant temperature is greater than'he corresponding saturation temperature.
Footnote (c) has been added which indicates that only the trip circuit will be calibrated once per three months, and tested once per month, while the primary sensors will be calibrated once per operating cycle.
The reason for this change is the same as for footnote (1) on Page 196, that the electronic sensors replacing the existing primary sensors are more reliable and maintain their accuracy for longer periods of time than mechanical switches.
~Pa e 231 The set point for the High Pressure Coolant Injection (HPCI) system initiation on low Reactor Water Level has been changed to > 53 inches (Indicator scale).
This set point is the same as the existing setpoint of 1 foot below minimum normal water level at Elevation 302'". This change is being made to be consistent with the other water level set points which are given in indicator scale inches.
The requirement -for (HPCI) System initiation on Low Reactor Water Level to be operable in the REFUEL mode has been deleted. Added a new footnote (a) in the STARTUP and SHUTDOWN column for HPCI initiation on Low Reactor Water. Level and Automatic Turbine Trip. These changes are explained with the changes on Page 232a.
~Pa e 232 Added footnote (b) to the Instrument Channel Calibration and Instrument Channel Test columns of parameter (1). This footnote is explained with the changes on Page 232a.
~Pa e 232a A new footnote (a) has been added which indicates that HPCI initiation on Low Reactor Water Level and Automatic Turbine Trip may be bypassed when the reactor pressure is less than 110 psig and the reactor coolant temperature is less than the corresponding saturation. temperature. This is consistent with Specification 3.1.8 on Page 71. Therefore, HPCI initiation is also not required in the COLD SHUTDOWN or REFUEL modes.
The previous footnote (a) has been deleted, because it is no longer necessary to bypass HPCI initiation in the COLD SHUTDOWN mode to perform maintenance as specified in Specification 2.1;le.
Footnote (b) has been added which indicates that only the trip circuit will be calibrated once per three months, and tested once per month, while the primary sensors will be calibrated and tested once per operating cycle. The reason for this change is the same as for footnote (1) on Page 196, that the electronic sensors, replacing the existing primary sensors are more reliable and maintain their accuracy for longer'eriods of time than mechanical switches.
~Pa e 233 A typographical error was corrected in the fourth sentence of the second paragraph. The wording "...in two-out-of-two" was changed to "...is two-out-of-two".
The reference in the second paragraph to the low condenser vacuum trip was deleted. This trip is not a safety related function and is only provided for protection of equipment as indicated for the changes on Page 192.
~Pa e 237a The last sentence on the page has been revised to indicate that the IRM and APRM downscale rod blocks are set at 5 and 2 percent of full scale respectively.
Thi's is consistent with Table 3.1.2g.
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~ fjc e
ATTACHMENT C NIAGARA MOHAWK POWER CORPORATION License No. DPR-63 Docket No. 50-220 AMENDMENT CLASSIFICATION The proposed amendment to the Operating License has been evaluated and determined to fall within the definition of Class III of 10CFR170.22, thereby four thousand dollars ($ 4,000.00).
requiring a fee of The proposed amendment includes some changes to the Nine Mile Point Unit 1 Technical Specifications which are administrative in nature. Other changes have clearly been identified by the Commission as acceptable in its review and acceptance of General Electric Topical Report NED0-21617.
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