ML18038A193
| ML18038A193 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 08/06/1986 |
| From: | Mangan C NIAGARA MOHAWK POWER CORP. |
| To: | Adensam E Office of Nuclear Reactor Regulation |
| References | |
| (NMP2L-0807), (NMP2L-807), NUDOCS 8608110101 | |
| Download: ML18038A193 (74) | |
Text
REQUL RY I NFORNATION DISTR IBUT SYBTEI"2 (R IDS)
ACCESSION NBR: 86080}01 DOC. DATE: 86/08/06 NOTARIZED: YEB DOCKET 0 FACIL: 50-4}0 Nine Nile Point Nuclear Station> Unit 2i Niagara Noha 050004}0 AUTH. NAI'IE AUTHOR AFFIL1 ATION NANQANi C. V. Niagara Nohatok Power Corp. RECIP. NANE RECIPIENT AFFILIATION ADENBAN> E. Q. BWR PY'9JBC t 'Dll Btttv'ELtB 3 Q
SUBJECT:
Fortuards changes to Final draft Tech Specs. transmitted bM NRC 860627 itr> categorized as necessary for certifi for operational flexibility or clarification. cation'eeded Expeditious resolution of items requested. DISTRIBUTION CODE: B00}D COPIES RECEIVED: LTR ENCL SI ZE: TITLE: Licensing Submittal: PBAR/FBAR Amdts 8< Related Correspon ence NOTES: REC IP IENT COPIES RECIPIENT COPIES ID CODE/NA2'2E LTTR ENCL ID CODE/NANE LTTR ENCL BNR EB 1 BWR EICSB 2 2 BNR FOB 1 1 BNR PD3 LA BNR PD3 PD 1 1 HAUQHEY,22 0} 2 2 BNR PSB 1 8NR RSB 1 INTERNAL: ACRS 4} 6 6 ADM/LFNB 1 0 ELD/HDB3 1 0 IE FILE 1 1 IE/DEPER/EPB 36 } 1 IE/DG*VT/GAB 21 1 1 NRR BMR ADTB } 0 NRR PNR-B ADTS 1 0 N 1 NRR/DHFT/NTB 1 1 I 04 1 RQN} 3 3 2I /2'21 8 0 EXTERNAL: BNL<ANDTS ONLY) 1 DMB/DBS (ANDTS) LPDR 03 1 1 NRC PDR 02 NSIC 05 1 PNL QRUELi R TOTAL NU2'2BER OF COPIES REQUIRED: LTTR 36 ENCL 31
l <<Cr j'-, ~ " 'X C ~ AA>> '3< k'>> 4 &4G '$M, 4'k P $ 3 k! C>>, <<k j,<<(C>>'<< ll j, . C f kk 4 jk 1 4 ~
!'(" "<'j 44 TC ' '<<UG <<.< <<t
>>jjkj j k
'k 1>
I 4 p .S<<1
'3 k~ C.. b9., "5'>> 'll
'Yt- >>Ck I' ~ ".
~ I 4 ,<< ll>>
f~ C>> Tl<1,p <
"C 4>>, Ck>> jS 'It",y
!I'1 JX f(>>t:,
i:C A.-k ': '1 <<'<<
$ 4 $ ,]C'+ I 4( 4 4' <<4*>>>> g ~ ~ kk<<',<<$ ',">>>> V ~
+4 kk>>
j <<>>) 4 4 4 I 1 4 RICE, ),j>>'4 k V'4, 4<<f pl;,'i,;," 4 k k.. kla C,k"' a gg l,k .44 4 kl 1 <<',J 4 4, It>>
"1!'> 'q . '4
'1 nfl, Cj C">>1 k,f>>' CCk . 'P 4 1<<4C-<<<<
.'"(. 3R
( C, j kkg('jk k 4 lk'4 Ck N 4
iM U MQIMWK NIAGARA MOHAWK POWER CORPORATION/300 ERIE BOULEVARD WEST, SYRACUSE, N.Y. 13202/TELEPHONE (315) 474.1511 August 6, 1986 (NMP2L 0807) Ms. Elinor. G. Adensam, Director BWR Project Directorate No. 3 U.S. Nuclear Regulatory Commission 7920 Norfolk Avenue Washington, DC 20555
Dear Ms. Adensam:
Re: Nine Mile Point Unit 2 Docket No. 50-410 We have substantially completed our review of the Final Draft Technical Specifications for Nine Mile Point Unit 2 which you provided by letter dated June 27, 1986. As a result, we have identified a number of necessary changes which are categorized as necessary for certification, editorial, needed for operational flexibility, or clarifying in nature. The specific changes to the Technical Specifications and their justification are provided in the enclosure to this letter. Where the requested changes and justifications have already been provided to you, reference to the transmittal letter is made. Where the requested changes to the Technical Specifications also affect the Final Safety Analysis Report or the Safety Evaluation, the changes to the appropriate pages are enclosed. A list of the changes to the Technical Specifications, the Final Safety Analysis Report, and the Safety Evaluation Report are included to aid your staff in the review of these changes. Since certification of the Technical Specifications now appears to be a critical step in the licensing of Nine Mile Point Unit 2, we would appreciate your expeditious resolution of these items. We are continuing our review of the recently received Supplement 3 of the Safety Evaluation Report and will inform you of our comments when it is complete. Very truly yours, C. V. Mang Senior Vice President LSL: ja 1889G Enclosures xc: W. A. Cook, NRC Resident Inspector Project File (2) (oh ~ 8608110101 PDR 860806
*DOCK'5000410 i'
PDR';,
C' 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. Man an , being duly sworn, states that he is Senior 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 swor to before me, a Notary Public in for th State of York and County of , this ~~ day of d New 1986. tary Public in and for County, New York My Commission expires:
'JANIS M. MACRO Notary Public ln thc Stato ol Ncw York
,Quallllcd In On nrta"a County No. 47845" 5
jest LIST OF TECHNICAL SPECIFICATIONS, FINAL SAFETY ANALYSIS REPORT, AND SAFETY EVALUATION REPORT PAGES CHANGED Descri tion Document(>) ~Pa e ~Cate or Primary Containment Isolation T.S. B3/4 6-5 Clarification Valves Radioactive Liquid Effluent T.S. 3/4 3-99 Certification Monitoring Instrumentation T.S. 3/4 3-101 FSAR 11.5-13 SER 11-10 Safety Relief Valves T.S. 3/4 4-10 Operational T.S. 3/4 5-3 T.S. B3/4 4-3 T.S. B3/4 5-2 SER 6-3 Accident Monitoring T.S. 3/4 3-85 Clarification Instrumentation T.S. 3/4 3-86 Administrative Controls<2> T.S. 6-11 Certification Source Check<2> T.S. 1-8 Certification Secondary Containment T.S. 1-7 Certification Integrity Hain Steam Isolation Valve T.S. 3/4 6-6 Operational Leak Rate Fire Protection Program<2> T.S. 3/4 7-25 Operational T.S. 3/4 7-30 Primary Containment T.S. 3/4 6-28 Certification Isolation Valves 860811010 e
Descri tion Document (1) ~Pa e ~Cate or Rod North Ninimizer T.S. 3/4 1-16 Certification 3/4 10-2 Other Items T.S. 2-1 Certification T.S. 2-7 T.S. 3/4 3-66 T.S. 3/4 3-82 T.S. 3/4 3-83 T.S. 3/4 3-85 T.S. 3/4 4-16 T.S. 3/4 5-5 T.S. 3/4 6-24 T.S. 3/4 7-3 T.S. 3/4 7-6 T.S. 3/4 10-7 Other Items T.S. 3/4 3-3 Editorial T.S. 3/4.3-45 T.S. 3/4 3-79 T.S. 3/4 6-28 T.S. 3/4 6-42 T.S. 3/4 7-26 T.S. 3/4 8-2
Descri tion Document (1) ~Pa e ~Cate or Other Items (cont.) T.S. 3/4 8-3 Editorial T.S. 3/4 8-21 T.S. 3/4 8-22 T.S. 3/4 8-23 T.S. 3/4 8-28 T.S. 83/4 4-7 Notes: (1) T.S. = Technical Specifications FSAR = Final Safety Analysis Report SER = Safety Evaluation Report (2) This item was previously submitted to the Nuclear Regulatory Commission.
Changes to Technical Specification Bases 3/4.6.3 "Primary Containment Isolation Valves"
Sub]ect: Justification for change to Technical Specification Bases 3/4.6.3, "Primary Containment Isolation Valves" The requested change is enclosed. This change will clarify the relationship between isolation system instrumentation response time and isolation valve closing time. CHANGE REQUESTED FOR CLARIFICATION
CONTAINMENT SYSTEMS BASES PRIMARY CONTAINMENT PRIMARY CONTAINMENT ISOLATION VALVES 3/4.6. 3 (Continued) GDC 54 through 57 of Appendix A to 10 CFR 50: Measurement of the closure time of automatic containment isolation valves is performed for the purpose of de-monstrating PRIMARY CONTAINMENT INTEGRITY and system OPERABILITY (Specifica-tion 3/4.6.1). The maximum isolation times for primary containment automatic isolation valves listed in this specification are either the analytical times used in the acci-dent analysis as described in the FSAR; or times derived by applying margins to the vendor test data obtained in accordance with industry codes and standards. For non-analytical automatic primary containment isolation valves, the maxi-mum isolation time is derived as follows:
- 1) Valves with full stroke times less than or equal to 10 seconds, maximum isolation time approximately equals the vendor tested closure time multi-plied by 2.0.
- 2) Valves with full stroke time greater than 10 seconds, maximum isolation time approximately equals the vendor tested closure time multiplied by 1.5.
3/4.6.4 SUPPRESSION CHAMBER -, DRYMELL VACUUM BREAKERS Vacuum relief breakers are provided to equalize the pressure between the suppres-sion chamber and drywell. This system will maintain the structural integrity of the primary containment under conditions of large differential pressures.'he vacuum breakers between the suppression chamber and the drywell must not be inoperable in the open position since this would allow 'bypassing of the suppression pool in case of an accident. Ther e are four pairs of valves to provide redundancy so that operation may continue for up to 72 hours with no more than one pair of vacuum breakers inoperable in the closed position. 3/4.6.5 SECONDARY CONTAINMENT Secondary containment is designed to minimize any ground level release of radioactive material which may result from an accident. The reactor building 'and associated structures provide secondary containment during normal opera-tion when the drywell is sealed and in service. At other times, the drywell may be open and, when required, secondary containment integrity is specified. Establishing and maintaining a subatmospheric condition in the reactor build-ing with the standby gas treatment system once per 18 months, along with the surveillance of the doors, hatches, dampers, and valves, is adequate to ensure that there are no violations of the integrity of the secondary containment. NINE MILE POINT - UNIT 2 B3/4 6-5
1 II r', I E.."
Changes to Technical Specifications Tables 3.3.7.10-1 and 4.3.7.10-1 in the Area of Radioactive Liquid Effluent Monitoring Instrumentation
~ ~
Subject:
Justification for changes to Technical Specification Tables 3.3.7.10-1 and 4.3.7.10-1 in the area of radioactive liquid effluent moni toring instrumentation The current Technical Specification Section 3.3.7.10 requires the Liquid Radwaste Monitor to be OPERABLE at all times, whether radwaste discharge is occurring or not. System design provides three valves to prevent inadvertent discharge. These valves must be specifically lined up in the course of making a discharge. Inherent in this design is the isolation of the small section of discharge line from and to which the Liquid Radwaste monitor's sample pump takes supply and return. When in continuous use, the sample pump produces more heat than can be dissipated in the small volume of water contained in this section of pipe. Therefore, it is requested to revise Technical Specification Tables 3.3.7.10-1 and 4.3.7.10-1 to provide:
- 1. The Liquid Waste Monitor must be OPERABLE at all times during discharge of liquid waste.
- 2. The CHANNEL CHECK and SOURCE CHECK are to be performed P (prior to discharge).
The requested changes to Technical Specification Tables 3.3.7.10-1 and 4.3.7.10-1 are enclosed. These changes also affect the Final Safety Analysis Report and the Safety Evaluation Report. Changes to the appropriate pages of these reports are also enclosed. CHANGE REQUESTED FOR CERTIFICATION
~ ~ INSTRUMENTATION MONITORING INSTRUMENTATION RADIOACTIVE LI UID EFFLUENT MONITORING INSTRUMENTATION LIMITING CONDITIONS FOR OPERATION 3.3.7.10 The radioactive liquid effluent monitoring instrumentation channels shown in Table 3.3.7. 10-1 shall be OPERABLE with their Alarm/Trip Setpoints set to ensure that the limits of Specification 3. 11. 1. 1 are not exceeded, The Alarm/Trip Setpoints of these channels shall be determined and adjusted in accordance with the methodology and parameters in the OFFSITE DOSE CALCULATION MANUAL (ODCM). APPLICABILITY: At all times ACTION: With a radioactive liquid effluent monitoring instrumentation channel Alarm/Trip Setpoint less conservative than required by the above specifi-cation, immediately suspend the release of radioactive liquid effluents monitored by the affected channel, or declare the channel inoperable, or change the setpoint so it is acceptably conservative.
- b. With the number of channels OPERABLE less than the Minimum Channels OPER-ABLE requirement, take the ACTION shown in Table 3.3.7.10-1. Restore the instruments to OPERABLE status within 30 days and, if unsuccessful, explain in the next Semiannual Ra'dioactive Effluent Release Report why the inoperability was not corrected in a timely manner.
C. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. SURVEILLANCE RE UIREMENTS 4.3.7.10 Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST. at the frequencies shown in Table 4.3.7.10-1. NINE MILE POINT - UNIT 2 3/4 3-98
e TABLE 3. 3.7.10-1 RADIOACTIVE LI UID EFFLUENT MONITORING INSTRUMENTATION MINIMUM CHANNELS INSTRUMENT OPERABLE ACTION
- 1. Radioactivity Honitors Providing Alarm and Automatic Termination of Release Liquid Radwaste Effluent Line 128
- 2. Radioactivity Monitors Providing Alarm But Not Providing Automatic Termination of Release
- a. Service Water Effluent Line A 130
- b. Service Water Effluent Line B 130
- c. Cooling Tower Blowdown Line 130
- 3. Flow Rate Measurement Devices
- a. Liquid Radwaste Effluent Line 131
- b. Service Water Effluent Line A 131
- c. Service Water Effluent Line 8 131
- d. Cooling Tower Blowdown Line 131
- 4. Tank Level Indicating Devices* 132
- Tanks included in this specification are those outdoor tanks that are not surrounded by liners, dikes, or walls capable of holding the tank contents and do not have tank overflows and surrounding area drains connected to the liquid radwaste treatment system, such as temporary tanks.
NINE NILE POINT - UNIT 2 3/4 3-99
0 TABLE 3.3.7. 10-1 (Continue6) RAOIOACTIVE LI UID EFFLUENT MONITORING INSTRUMENTATION TABLE NOTATIONS ACTION 128- With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases may continue provided that before initiating a release:
- a. At least two independent samples are analyzed in accordance with Specification 4. 11. 1.1.1, and
- b. At least two technically qualified members of the facility staff independently verify the release rate calculations and discharge line valving; Otherwise, suspend release of radioactive effluents via this pathway.
ACTION 129 - Not used. ACTION 130 " With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided that, at least once per 12 hours, grab samples are collected and analyzed for radioac-tivity at a limit of detection of at least 5 x 10-~ microcuries/ml. ACTION 131- With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue, provided the flow rate is estimated at least once per 4 hours during actual releases. Pump per-formance curves generated in place may be used to estimate flow. ACTION 132- With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, liquid additions to this tank may continue provided the tank liquid level is estimated during all liquid additions to the tank. NINE MILE POINT - UNIT 2 3/4 3-100
TABLE 4.3.7.10-1 foal RADIOACTIVE LI UID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE RE UIREMENTS CHANNEL CHANNEL SOURCE CHANNEL FUNCTIONAL INSTRUMENT CHECK CHECK CALIBRATION TEST
- 1. Radioactivity Monitors Providing Alarm and Automatic Termination of Release Liquid Radwaste Effluent Line R(c) M(a)(b)
- 2. Radioactivity Monitors Providing Alarm But Not Providing Automatic Termination of Release
- a. Service Water Effluent Line A R(c) SA(b)
- b. Service Water Effluent Line B R(c) SA(b)
- c. Cooling Tower Blowdown Line R(c) SA(b)
- 3. Flow Rate Measurement Devices
- a. Liquid Radwaste Effluent Line D(d)
- b. Service Water Effluent Line A D(d)
- c. Service Water Effluent Line B D(d)
- d. Cooling Tower Blowdown Line D(d)
- 4. Tank Level Indicating Devices"
- Tanks included in this specification are those outdoor tanks that are not surrounded by liners, dikes, or w capable of holding the tank contents and do not have tank overflows and surrounding area drains connected the liquid radwaste treatment system, such as temporary tanks.
"* During liquid additions to the tank.
TABLE 4.3.7.10-1 (Continued) RADIOACTIVE LI UID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE RE UIREMENTS TABLE NOTATIONS (a) The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occurs the instrument indicates measured levels above the Alarm/Trip Setpoint. if (b) The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exists: (1) Instrument indicates measured levels above the Alarm Setpoint, or (2) Circuit failure, ot (3) Instrument indicates a downscale failure, or (4) Instrument controls not set in operate mode. (c) The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Bureau of Standards, standards that are traceable to the National Bureau of Standards, or using actual samples of liquid effluents that have been analyzed on a system that has been calibrated with National Bureau of Standards trace-able sources. These standards shall permit calibrating the system over its intended range of energy and measurement. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration may be used. (d) CHANNEL CHECK shall consist of verifying indication of flow during periods of release. CHANNEL CHECK shall be made at least once per 24 hours on days on which continuous, periodic, or batch releases are made. NINE MILE POINT - UNIT 2 3/4 3" 102
Nine Nile Point Unit 2 FSAR radioactivity alarm in the main control room. Tritium in the plant areas is determined on the basis of representative grab samples collected from the effluent points or ventilation exhaust ducts. Grab samples are obtained from locations indicated in Table 11.5-2. Samples are analyzed in the health physics laboratory, or by contracted laboratories. 11.5.3 Effluent Monitoring and Sampling All potentially radioactive gaseous and liquid effluent discharge paths are either continuously monitored or routinely. sampled for radiation level during discharge (Section 11.5.2). Solid waste shipping containers are monitored with gamma sensitive portable survey instruments. The following gaseous effluent paths are sampled and monitored:
- 1. Plant main stack exhaust.
- 2. Combined radwaste/reactor building ventilation exhaust.
The following liquid effluent paths are sampled and monitored:
- 1. Liquid radwaste system effluent.
- 2. Circulating water system cooling tower blowdown line.
- 3. Service water system discharge.
All monitor ranges are listed in Table 11.5-1. An isotopic analysis is performed periodically on samples obtained from each liquid effluent release path to verify the adequacy of effluent processing to meet the discharge limits to unrestricted areas. This effluent monitoring and sampling program is comprehensive and provides the information for the effluent measuring and reporting programs required by 10CFR50 Section 36a, Appendix A, General Design Criterion 64, and Appendix I and Regulatory Guide 1.21 in semiannual reports to the NRC. The frequency of the periodic sampling and ana'lysis described in the technical specifications is a minimum and is increased if effluent levels approach technical speci-fication limits. Isotopic content of gaseous'ffluents is continuously monitored by offline monitors. All
equipment performance, and (4) monitor and control radioactivity levels in plant discharges to the environs. Table 11.5 provides the proposed locations of continuous monitors. Monitors on certain effluent release lines will automatically terminate discharges in the event that radiation levels exceed a predetermined value. Systems that are not amenable to continuous monitoring, or for which detailed isotopic analyses are required, will be periodically sampled and the samples will be analyzed in the plant laboratory. The potential airborne radioactivity releases to the environs from NMP-2 from two monitored points: (1) plant main stack, and (2) the combined radwaste and reactor building vent. The plant main stack receives inputs from standby gas treatment, turbine building ventilation, turbine gland seal exhaust, condenser offgas, and mechan-ical vacuum pump exhaust. The combined reacto~ and radwaste building vent receives inputs from radwaste building ventilation and normal reactor building ventilation. Each release point is monitored by an online isotopic gaseous radiation monitoring system that is designed (1) to obtain continuous isokinetic and representative samples; (2) to continuously determine gaseous effluent isotopic content; (3) to store the information on releases of radioactive particulates, iodine, and noble gas; (4) to retrieve the data on command; and (5) to alarm in the main control room in the event that specified rates of. release of radioactive material are exceeded. Each online isotopic gaseous radiation monitor consists of three detectors (one each for iodine, particulate, and noble gas channels) with associated valving, electronics, multichannel analyzer, and computer. The monitor satis-fies all requirements of NUREG-0737 with a range that satisfies RG 1. 97. The monitor is capable of functioning both during and after an accident, and provides continuous monitoring of high-level postaccident releases of radioactive noble gases from the plant. Gr rOtakiN,<(y~e /Cd The potential radioactive liquid ffluent release points to the environment are (1) processed 1 i qui d radwas te discharge, (2) circulating water system cooling tower bl owdown ine, (3) esidual heat removal system service water 1 effluent,~yg,(4) service water s stem discharge to l.ake Ontario. All re'lease points are'~continuously monitored for radioactivity discharge. The liquid radwaste discharge is automatically terminated in the event that radio-ihs activity concentration reaches a preset value. All other monitors will alarm in the main control room upon detection of radioactivity in the discharge. The staff's review included the locations and types of effluent and process monitoring provided for NMP-2. On the basis of the plant's design, continuous monitoring locations, and intermittent sampling locations, the staff concluded that all normal and potential release pathways are monitored. The staff also determined that the sampling and monitoring provisions are adequate for detecting radioactive material leakage to normally uncontaminated systems and for moni-toring plant processes that could affect radioactivity releases. On these bases, the staff considers that the monitoring and sampling provisions satisfy the requi rements of GOC 60, 63, and 64 and conform to the guidelines of RG
- l. 21.
NMP"2 SER 11-10
Changes to Technical Specifications in the Area of Safety/Relief Valves
g li t,
Subject:
Justification for changes to Technical Specifications in the area of safety relief valves Requested changes to the Technical Specifications are enclosed. These changes support the power ascension test program. Justification for these changes is in Appendix 15C, "Two Safety/Relief Valves Out of Service," of the Final Safety Analysis Report. Proposed change to Supplement 2 of the Safety Evaluation Report is also enclosed. This change is consistent with Appendix 15C. CHANGE RE(VESTED FOR OPERATIONAL FLEXIBILITY
1 S ( REACTOR COOLANT SYSTEM 3/4.4.2
~ ~ SAFETY/RELIEF VALVES
'I LIMITING CONDITIONS FOR OPERATION I&
3.4.2 The safety valve function of at 'least ~of the following reactor coolant system safety/relief valves shall be OPERABLE with the specified code safety valve function valve shall be OPERABLE: lift settings"; the acoustic monitor for each OPERABLE 2 safety/relief valves 8 1148 psig 21K 4 safety/relief valves. 9 1175 psig kiX 4 safety/relief valves 9 1185 psig tlX 4 safety/relief safety/relief valves valves I 1195 psig t1X 4 8 1205 psig tlat APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3 l;-I re,~w a. Valve. & oPaahsLE. ACTION: s4zhlg, wi+irt 72 holltL of' With the safety valve function of one ~~ of the above required /~
~
~
safety/relief valves inoperable, be in at least HOT SHUTDOWN within +ha sex't 12 hours and in COLD SHUTDOWN within the 24 hours. gollou)i~
- c. P. With one or more safety/relief valves stuck open, provided that the average water temperature in the suppression pool is less than 110'F, close the. stuck-open safety/relief valve(s);
valve(s) within 5 minutes or if if unable to close the open the average water temperature in the suppression pool is 110'F or more, place the reactor mode switch in the Shutdown position. With one or more safety/relief valve acoustic monitors inoperable, restore the inoperable monitor(s) to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours. I n e f y gable PnoIi'on ino~nalnle fan Nom, fkan o'ne of- 44e alnoVe oamto foe I ia CafJLQ /relief palnee, he i>> cia leant HOT <H<~>++ I2 hours and ie CcLb 5H.lL73>otdN ld94io
" The lift setting pressure shall correspond to ambient conditions of the valves at nominal operating temperatures and pressures.
NINE MILE POINT - UNIT 2 3/4 4-10
REACTOR COOLANT SYSTEM , 3/4.4.2 SAFETY/RELIEF VALVES SURVEILLANCE RE UIREHENTS 4.4.2.1 The acoustic monitor for each safety/relief valve shall be demonstrated OPERABLE with the setpoint verified to be 0.25 of the full-open noise level* by performance of a:
- a. CHANNEL FUNCTIONAL TEST at least once per 31 days, and a
- b. CHANNEL CALIBRATION at least once per 18 months.""
"Initial setting shall be in accordance with the manufacturers recommendation.
Adjustment to the valve full-open noise level shall be accomplished during the startup test program.
*"The provisions of Specification 4.0.4 are not applicable provided the surveillance is performed within 12 hours after reactor steam pressure is adequate to perform the test.
NINE MILE POINT " UNIT 2 3/4 4-11
EMERGENCY CORE COOLING SYSTEMS ECCS - OPERATING LIMITING CONDITIONS FOR OPERATION 3.5.1 (Continued) ACTION: elva.'~
- d. For ECCS Divisions I and II, provided that ECCS Division III is OPERABLE:
- 1. With LPCI subsystem "A" and either LPCI subsystem "B" or "C" inoperable, restore at least the inoperable LPCI subsystem "A" or the inoperable LPCI subsystem "B" or "C" to OPERABLE status within 72 hours.
- 2. With the LPCS system inoperable and either LPCI subsystems "B" or "C" inoperable, restore at least the inoperable LPCS system or the inoperable LPCI subsystem "B" or "C" to OPERABLE status within 72 hours.
- 3. Otherwise, be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours".
- e. For ECCS Divisions I and II, provided that ECCS Division III is OPERABLE and Divisions I and II are otherwise OPERABLE:
- 1. Nith 41&of ttte above required AOS valves inoperable, restore the 4aeperatAe-4BS valv o OPERABLE status within 14 days or be in at least HOT SHUTDOWN within the next 12 hours and reduce reactor steam dome pressure to less than or equal to 100 psig within the next 24 hours.
2.
~t ee With ~o- or more of:the above required ADS valves inoperable, be in at least HOT SHUTDOWN within 12 hours and reduce reactor steam dome pressure .to less than or equal to 100 psig within the next 24 hours.
- f. In the event an ECCS is a'ctuated and injects water into the reactor coolant system, a Special'eport shall be prepared and submitted to the Commission pursuant to Sp'ecification 6.9.2 within 90 days, describing the circumstances of the actuation and the total accumulated actuation cycles to date. The current value of the usage factor for each affected safety injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70.
" Whenever'two or more RHR subsystems are inoperable, if unable to attain COLD SHUTDOWN as required by this ACTION, maintain reactor coolant temperature as low as practical by use of alternate heat removal methods. I NINE MILE POINT " UNIT 2 3/4 5-3
REACTOR COOLANT SYSTEM BASES RECIRCULATION SYSTEM 3/4.4. 1 (Continued) recirculation pump and recirculation nozzles. Sudden equalization of a tem-perature difference > 145'F between the reactor vessel bottom head coolant and the coolant in the upper region of the reactor vessel by increasing core flow rate would cause undue stress in the reactor vessel bottom head. 3/4.4. 2 SAFETY/RELIEF VALVES The safety/relief valves operate during a postulated ATMS event to prevent the reactor coolant system from being pressurized above a design allowable value of 1375 psig in accordance with the ASME Code. A total of+ P BLE safety/relief valves is required to limit local pressure at active components to within ASME III allowable design values (Service Level A). All other appropriate ASME III limits are also bounded by this requirement. The safety-relief valve lift settings will be demonstrated only during shutdown in accordance with the provisions of Specification 4.0.5. 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.3. 1 LEAKAGE DETECTION SYSTEMS The RCS leakage detection systems required by this specification are provided to monitor and detect leakage from the reactor coolant pressure boundary. These detection systems are consistent with the recommendations of RG 1.45, "Reactor Coolant Pressure Boundary Leakage Detection Systems," May 1973. 3/4.4. 3. 2 OP ERAT IONAI LEAKAGE The allowable leakage rates from the reactor coolant system have been based on the predicted and experimentally observed behavior of cracks in pipes. The background leakage normally expected to result from equipment design and the detection capability of the instrumentation for determining system leakage were also considered. The evidence obtained from experiments suggests that for leak-age somewhat greater than that specified for UNIDENTIFIED LEAKAGE, the probabil-ity is small that the imperfection or crack associated withexceed such leakage would grow rapidly. However, in all cases, if the leakage to rates the values specified or the leakage is located and known be PRESSURE BOUNDARY LEAKAGE, the reactor will be shut down to allow further investigation and corrective action. The Surveillance Requirements for RCS pressure isolation valves provide added assurance of valve integrity, thereby reducing the probability of gross valve failure and consequent intersystem LOCA. NINE MILE POINT - UNIT 2 B3/4. 4-3
EMERGENCY CORE COOLING SYSTEM BASES ECCS - OPERATING AND SHUTDOWN 3/4.5. 1 8 3/4.5.2 (Continued) The capacity of the system is selected to provide the required core cooling. The HPCS pump is designed to deliver greater than or equal to 516/1550/6350 gpm at differential pressures of 1160/1130/200 psi, respectively. Initially, water from the condensate storage tank is used instead of water injected from the suppression pool into the reactor, but no credit is taken in the safety analyses for the condensate storage tank water. With the HPCS system inoperable, adequate core cooling is assured by the OPERA-BILITY of the redundant and diversified automatic depressur ization system and both the LPCS and LPCI systems. In addition, the reactor core isolation cooling (RCIC) system, a system for which no credit is taken in the safety analysis, will automatically provide makeup water at reactor operating pressures on a reactor low water level condition. The HPCS out-of-service period of 14 days is based on the demonstrated OPERABILITY of redundant and diversified low-pressure core cooling systems. I The Surveillance Requirements provide adequate assurance that the HPCS system will be OPERABLE when required. Although all active components are testable and full flow can be demonstrated by recirculation through a test loop during reactor operation, a complete functional test with reactor vessel injection requires the reactor to be shut down. The pump discharge piping is maintained full to prevent water hammer damage. Upon failure of the HPCS system to function properly after a small-break loss-of"coolant accident, the automatic depressurization system (ADS) automatically .causes selected safety/relief valves to open, depressurizing the reactor so that flow from the low-pressure core cooling systems can enter the core in time to limit fuel cladding temperature to less than 22004f. ADS is conservatively required to be OPERABLE whenever reactor vessel pressure exceeds 100 psig. This pressure is substantially below that for which the low-pressure core cooling systems can provide adequate core co for events requiring ADS. f-iv ADS automatically controls seven selected safe y/relief valves although the safety analysis only takes credit for 'alves. It is, therefore, appro-priate to permitwne valvesto be out of service for up to 14 days without materially reducing system reliability. NINE MILE POINT - UNIT 2 B3/4 5-2
rapidly propagating fracture will be minimized, and that the requirements of General Design Criterion (GDC) 51 are satisfied. 6.3 Emer enc Core Coolin S stem In Section 6.3 of the SER the staff discussed the loss-of-coolant accident (LOCA) analyses results for a lead plant that was representative of NMP-2. The appli-cant had committed to perform the plant-specific LOCA analyses and determine the minimum number of automatic depressurization system (ADS) safety relief valves (SRVs) needed to achieve a rapid depressurization during a small-break LOCA based on plant-specific LOCA analyses. The applicant provided the plant-specific LOCA analyses in FSAR Amendment 20, dated July 1985. The plant-specific LOCA analyses included a spectrum of large and small postulated pipe breaks and the worst single failures based on the lead plant LOCA analyses. The results of the plant-specific LOCA analyses indicate that the most limiting break is a design-basis break in the recirculation suc-tion'iping with a low-pressure core spray (LPCS) diesel generator failure. The plant-specific LOCA analysis shows that the calculated maximum total hydro-gen generation from the chemical reactor of the cladding with water or steam for the most limiting LOCA case is 0.07K, which is well within the 10 CFR 50..46 limit of 1X. Table 6.2 il'Iustrates how the plant-specific LOCA analysis results meet the re-quirements of 10 CFR 50.46. In a small-break emergency core cooling system (ECCS) analysis, a particular ADS flow rate is required to bring the vessel pressure down in a prescribed time to allow the operation of the low-pressure core cooling system following the postulated failure of the high-pressure core spray (HPCS). This ADS flow rate determines the number of ADS valves. Seven SRVs were needed to perform t'e ADS function based on a generic BWR/5 calculation. The NMP-2 plant-specific ECCS analysis is in compliance with the requirements of 10 CFR 50, Appendix K, and has confirmed the adequacy of the seven ADS valve design by using only six ADS valves in the small-break ECCS analysis.
) The-be out technical- specification-weal.l- state..that-onIy-.one"ADS valve is permitted to of service during operation. CxAS8<7 R77dcrlab Pd~ghggg The staff concludes that the plant-specific LOCA analyses and the results for NMP-2 are acceptable. On the basis of the above, confirmatory issues 10, 15, and 16 in the SER are considered to be resolved. I
.,6. 4 Control Room Habitabilit S stems Section 6.4 of the SER contains the sentence "To ensure maintenance of a Iow-leakage control room, the staff wil'I require periodic surveillance through Technical Specifications to ensure that 1/8-in. water gauge control room pressurization relative NMP-2 SSER 2 6-3
It, OQ
INSERT to page 6-3, SSER-2 I A subsequent analysis to support entended operation with two SRVs out of service has been docketed as Appendix 15C of the FSAR. This analysis has confirmed the plant's capability to meet 10 CFR 50, Appendix K criteria by using only five ADS valves in the small break ECCS analysis.= The associated technical specification revision, hence, states that two ADS valves are permitted to be out of service during operation for up to 14 days.
Changes to Techni cal Speci fi cati on s Table 4.3.7.5-1 "Accident Monitoring Instrumentation Surveillance Requirements"
II
Subject:
Justification for changes to Technical Specification Table 4.3.7.5-1, "Accident Monitoring Instrumentation Surveillance Requirements" The requested changes are enclosed. The changes clarify the intent of the channel check and channel calibration requirements of primary containment isolation valve position indication. Without the requested change, it can be misinterpreted that the valve must be stroked in order to determine operability of the valve position indication. These changes were discussed with Carl Schulten, of your staff. CHANGE RE(VESTED FOR CLARIFICATION
1 t INSTRUMENTATION MONITORING INSTRUMENTATION ACCIDENT MONITORING INSTRUMENTATION LIMITING CONDITIONS FOR OPERATION 3.3.7.5 The accident monitoring instrumentation channels shown in Table 3.3.7.5-1 shall be OPERABLE. APPLICABILITY: As shown in Table 3.3.7,5-1. ACTION: With one or more accident monitoring instrumentation channels inoperable, take the ACTION required by Table 3.3.7.5-1. SURVEILLANCE RE UIREMENTS 4.3.7.5 Each of the above required accident monitoring instrumentation channels shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3.7.5-1., NINE MILE POINT - UNIT 2 3/4 3-81
TABLE 3.3.7.5-1 ACCIDENT HONITORING INSTRUHENTATION M m HINIHUH APPLICABLE C) REQUIRED NUMBER CHANNELS OPERATIONAL M INSTRUMENT OF CHANNELS OPERABLE CONDITIONS ACTION
- 1. Reactor Vessel Pressure 1, 2 80 C
M 2. Reactor Vessel Mater Level
- a. Fuel Zone 1 2 80,
- b. Wide Range 1, 2 80
- 3. .Suppression Pool Mater Level
- a. Narrow Range 1, 2, 3 83
- b. Wide Range 1, 2, 3 83
- 4. Suppression Pool Mater Temperature 8, 2/Quadrant 4, 1/Quadrant 1, 2 80
- 5. Suppression Chamber Pressure 1, 2 80
- 6. Drywell Pressure 1, 2 80
- 7. Drywell Air Temperature I, 2 80
- 8. Drywell Oxygen Concentration l. 2 80
- 9. Drywell Hydrogen Concentration 1, 2 80 Analyzer and Honitor
- 10. Safety/Relief Valve Position 2/Val ve 1/Valve 1, 2 80 Indicators*
TABLE 3. 3. 7. 5-1 (Continued) ACCIDENT MONITORING INSTRUMENTATION MINIMUM APPLICABLE REQUIRED NUMBER CHANNELS OPERATIONAL INSTRUMENT OF CHANNELS OPERABLE CONDITIONS ACTION
- 11. Drywell High Range Radiation Monitors 2 1, 2, 3 81
- 12. RHR Heat Exchanger Service Water Exchanger 1/Heat Exchanger 1, 2, 3 81 13.
Monitor Flux'/Heat Radiation Monitor Refuel Platform Area Radiation 82
- 14. Neutron APRM 1, 2 80 IRM 1, 2 80 SRM 1 80
- 15. Primary Containment Isolation Valve Position Indication 1, 2 "Acoustic monitoring and tail pipe temperature
- When handling fuel, or components in the fuel pool or reactor cavity.
tNeutron flux indication is sufficient to meet the OPERABILITY requirement of this specification.
Table 3.3.7. 5-1 (Continued) ACCIDENT MONITORING INSTRUMENTATION ACTION ACTION 80 - a. With the number of OPERABLE accident monitoring instrumentation channels less than the Required Number of Channels shown in Table 3.3.7.5-1, restore the inoperable channel(s) to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours.
- b. With the number of OPERABLE accident monitoring instrumentation channels less than the Minimum Channels OPERABLE requirements of Table 3. 3. 7. 5-1, restore the inoperable channel(s) to OPER-ABLE status within 48 hours or be in at least HOT SHUTDOWN within the next 12 hours.
ACTfON 81 - With the number of OPERABLE accident monitoring instrumentation channels less than required by the Minimum Channels OPERABLE re-quirement, either restore the inoperable channel(s) to OPERABLE status within 72 hours, or:
- a. Initiate the preplanned alternate method of monitoring the appropriate parameter(s), and abilityy
- b. In lieu of another report required by Specification 6.9.2, prepare and submit a Special Report to the Commission pur-suant to Specification 6.9.2 within 14 days following the event outlining the action taken, the cause of the inoper-and the plans and schedule for restoring the system to OPERABLE status.
.ACTION 82 - With the number of OPERABLE accident monitoring instrumentaiton channels less than the Minimum Channels OPERABLE requirements of Table 3.3.7.5-1, suspend movement of fuel or components in the fuel pool or reactor cavity, or, initiate the preplanned alter-nate method of monitoring the appropriate parameter(s). ACTION 83 - a. With the number of OPERABLE accident monitoring instrumentation channels less than the Required Number of Channels shown in Table 3.3.7.5-1, restore the inoperable channel(s) to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
- b. With the number of OPERABLE accident monitoring instrumentation channels less than the Minimum Channels OPERABLE requirement of Table 3.3.7.5-1, restore the inoperable channel(s) to OPERABLE status within 48 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
Action 84 - Take the ACTION required by Specification 3.6.3. NINE MILE POINT " UNIT 2 3/4 3-84
TABLE 4.3.7.5-1 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE RE UIREMENTS I m APPLICABLE Q oM CHANNEL CHANNEL OPERATIONAL INSTRUMENT CHECK CALIBRATION CONDITIONS I
- l. Reactor Vessel Pressure 1, 2 C
- 2. Reactor Vessel Water Level M a. Fuel Zone 1, 2
- b. Wide Range 1, 2
- 3. Suppression Pool Water Level
- a. Narrow Range R 1, 2, 3
- b. Wide Range R 1, 2, 3
- 4. Suppression Pool Water Temperature R" 1, 2
- 5. Suppression Chamber Pressure R* 1, 2
- 6. Drywell Pressure R 1, 2
- 7. Drywell Air Temperature RA 1. 2
- 8. Drywell Oxygen Concentration R 1, 2
- 9. Drywell Hydrogen Concentration Analyzer and Monitor Q*A 1, 2
- 10. Safety/Relief Valve Position Indicators R l. 2
- 11. Drywell High Range Radiation Monitors Rt 1, 2, 3
- 12. RHR Heat Exchanger Service Water Radiation Monitor R 1, 2, 3
- 13. Refuel Platform Area Radiation R Flux Monitor'eutron 14.
- a. APRM 1, 2
- b. IRM 1, 2
- c. SRM 1
- 15. Primary Containment Isolation Valve Position Indication
- 1, 2
TABLE 4. 3.7. 5-1 (Continued) ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE RE UIREMENTS TABLE NOTATIONS
" Excludes sensors; sensor comparison shall be done in lieu of sensor calibration. "* Using sample gas containing:
- a. One volume percent hydrogen, balance nitrogen.
- b. Four volume percent hydrogen, balance nitrogen.
The CHANNEL CALIBRATION shall consist of an electronic calibration of the channel, not including the detector, for range decades above 10 R/hr and a one point calibration check of the detector below 10 R/hr with an installed or portable gamma source. tt When handling fuel or components in the fuel pool or reactor cavity. f4'ed orggauvl ov Mogp~nt >ndic,afov Shall ba. Wre$ ied as iMi~4'g vcl<4'- +sirius. 0
+eiie The pasih oniodioation verification W wod >he rquiienients of risoiB senti'on ~ xh/v-3eao shall su/ice. in ineHiMy the reguireuent ashen pargorroed at this Pcpeney NINE MILE POINT " UNIT 2 3/4 3-86
ltd ~"
)~)~i) i))>>"* ~ s'> "l I) X P ugly I~~'~w g~
'1 ltd)w $ ', g gi,9 s
'I l' i" sA p
\ r
,'P ~ 4' i ,"'I 1 ~ t.. gr 1 I
,. ) ~ .,
~
) j;",,),<i'>
4 r~p;
. ic<gq~ ~ f N i>k, t,
(zan p>$ 'll p l1l
'4 le y yon gl J
Changes to Technical Specifications in the Area of Administrative Controls
l'f, E. y 8
Subject:
Changes to Technical Specifications in the area of Administrative Controls The requested changes and justification for these changes were submitted to you in a letter dated July 21, 1986. Enclosed is a copy of that letter for your information. CHANGE REQUESTED FOR CERTIFICATION
I d}}