ML20245J112
| ML20245J112 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 04/27/1989 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20245J111 | List: |
| References | |
| NUDOCS 8905040003 | |
| Download: ML20245J112 (47) | |
Text
_ - _ _ -
i (f Mouq[0 UNITED STATES
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g NUCLEAR REGULATORY COMMISSION 5
- j WASHINGTON, D. C. 20555
- %...../
METROPOLITAN EDISON COMPANY j
JERSEY CENTRAL POWER & LIGHT COMPANY PENNSYLVANIA ELECTRIC COMPANY GPU NUCLEAR CORPORATION DOCKET NO. 50-289
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THREE MILE ISLAND NUCLEAR STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 149 License No. DPR-50 1.
The Nuclear Regulatory Comission (the Consnission) has found that:
A.
The applications for amendment by GPU Nuclear Corporation, et al.
(the licensee) dated December 2 and December 19, 1988 and January 31, 1989, comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regula*.4 ns of the Comission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.
The issuance of this amendment will not be inimical to the comon l
defense and security or to the health and safety of the public; I
and E.
The issuance of this amcrdment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements t
have been satisfied.
l l
1 l
'8905040003 890427 ADOCK 0500 89 gDR
e-rp. l 2.
Accordingly, the license is amended by changes to the Technica; Specifications as indicated in the attachment to this license amendment, and paragraph 2.c.(2) of Facility Operating License No. DPR-50 are hereby amended to read as follows:.
(2) Technical Specifications The Technical Specifications contained in Appendix
' A, as revised through Amendment No.149, are hereby incorporated in the license. GPU Nuclear Corporation shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance, to be implemented within 30 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION c
ch
. Stolz, Director Pr ett Directorate I-vision of Reactor Projects I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: Apiil 27, 1989
.o.
gs ATTACHMENT TO LICENSE AMENDMENT NO.149 FACILITY OPERATING LICENSE NO. DPR-50 DOCKET NO. 50-289 Rnplace the following pages of the Appendix A Technical Specifications with the attached pages. The revised pages are identified by amendment number and contain vertical lines indicating the area of change.
Remove Insert 11 11 viii viii 3-13a 3-13a 3-14 3-14 3-18e 3-18c 3-18d 3-18d 3-18g 3-18g 3-23 3-23 3-24 3-24 3-29(Table 3.5-1) 3-29(Table 3.5-1) 3-37 3-37 3-37a 3-37a 3-61 3-61 3-62a 3-62a 3-106 3-106 3-107 3-107 3-109 3-109 4-4 (Table 4.1-1) 4-4(Table 4.1-1) 4-8 (Table 4.1-2) 4-8 (Table 4.1-2) 4-39 4-39 4-40 4-40 4-41 4-41 4-42 4-42 4-44 4-44 4-46 4-46 4-55 4-55 4-55b 4-55b 4-58 4-58 4 4-59 4-61 4-61 4-64 4-64 4-80
.4-80 4-80a 4-81 4-81 4-82 4-82 5-1 5-1 5-10
. 5-10 5,11...
Figurt.5-1.
Figure.5-1 Figure 5-2 Figure 5-2 Figure 5-3 Figure 5-3 t " Figure 5-4 6-2 6-2 6-5 6-5 l
6-8 6-8 6-19 6-19 6-19a 6-19a
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%8 TABLE OF CONTENTS Section Page l
2 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2-1 l
2.1 Safety Limits, Reactor Core 2-1 l
2.2 Safety Limits, Reactor System Pressure 2-4 i
2.3 Limiting Safety System Settings, Protection Instrumentation 2-5 3
LIMITING CONDITIONS FOR OPERATION 3-1 3.0 General Action Requirements 3-1 3.1 Reactor Coolant System 3-la 3.1.1 Operational Components 3-la 3.1.2 Pressurization Heatup and Cooldown Limitations 3 3.1.3 Minimum Conditions for Criticality 3-6 3.1.4 Reactor Coolant System Activity 3-8 3.1.5 Chemistry 3-10 3.1.6 Leakage 3-12 3.1.7 Moderator Temperature Coefficient of Reactivity 3-16 3.1.8 Single Loop Restrictions 3-17 3.1.9 Low Power Physics Testing Restrictions 3-18 3.1.10 Control Rod Operation 3-18a 3.1.11 Reattor Internal Vent Valves 3-18b 3.1.12 Pres urizer Power Operated Relief Valve (PORV) 3-18c anc Block Valve 3.1.13 Reactcr Coolant System Vents 3-18f 3.2 Makeup and Purification & Chemical Addition Systems 3-19 3.3 Emergency Cire Cooling, Reactor Building Emergency Cooline and Reactor Building Spray Systems 3-21 3.4 Decay Heat Removal Capability 3-25 3.4.1 Reacto/ Coolant System Temperature Greater than 250*F 3-25 3.4.2 Reactor Coolant System Temperature 250*F or Less 3-26 3.5 Instrumentation Systems 3-27 3.5.1 Operational Safety Instrumentation 3-27 3.5.2 Control Rod Group and Power Distribution Limits 3-33 3.5.3 Engineered Safeguards Protection System Actuation Setpoints 3-37 3.5.4 Incore Instrumentation 3-38 3.5.5 Accident Monitoring Instrumentation 3-40a 3.5.6 Chlorine Detection Systems 3-40f 3.6 Reactor Building 3-41 3.7 Unit Electrical Power System 3-42 3,8 Fuel Loading and Refueling 3-44 3.9 Deleted 3-46 l
3.10 Miscellaneous Radioactive Materials Sources 3-46 3.11 Handling of Irradiated Fuel 3-55 3.12 Reactor Building Polar Crane 3-57 3.13 Secondary System Activity 3-58 3.14 Flood 3-59 3.14.1 Periodic Inspection of the Dikes Around TMI 3-59 3.14.2 Flood Condition for Placing the Unit in Hot Standby 3-60 3.15 Air Treatment Systems 3-61 3.15.1 Emergency Control Room Air Treatment System 3-61 l
3.15.2 Reactor Building Purge Air Treatment System 3-62a a
3.15.3 Auxiliary and Fuel Handling Building Air Treatment System 3-62c 3
3.15.4 fuel Handling Building ESF Air Treatment System 3-62e j
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LfST OF FIGURES Figure Title 3.5-2K Axial Power Imbalance Envelope for Operation from 40+10/-0 to 100+10/-0 EFPD, THI-l 1
-3.5-2L Axlal Power Imbalance Envelope for Operation after-l 100+10/-0 EFPD, TMI-1 3.5-2M LOCA Limited Maximum Allowable Linear Heat Rate 3.5-1 Incore. Instrumentation' Specification Axial Imbalance Indication, THI-l-3.5-2 Incore Instrumentation Specification Radial Flux Tilt Indication, TMI-l 3.5-3 Incore Instrumentation Specification 3.11-1 Transfer Path to and from Cask Loading Pit
- 4.17-1 Snubber Functional Test - Sample Plan 2 5-1 Extended Plot Plan TMI 5-2 Site Topography 5 Mlle Radius 5-3' Locations of' Gaseous Effluent Release Points and Liquid Effluent Outfalls vili Amendment Nos. J [ M M J47, # 149 i
L l
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. Bases (Continued)
Although some leak rates on the order of gallons per minute may be tolerable from a dose point of view, it is recognized that leaks in the order of drops per minute through any of the barriers of the primary system could be indicative of materials failure such as by stress corrosion cracking.
If depressurization, isolation, and/or other safety measures are not taken promptly,.these small leaks could develop into much larger leaks, possibly into a gross pipe rupture. Therefore, the nature and location of the leak, as well as the magnitude of the leakage, must be considered in the safety evaluation.
J When reactor coolant leakage occurs to the Reactor Building, it is ultimately conducted to the Reactor. Building sump. Although the reactor coolant is safely contained, the gaseous components in it escape to the Reactor Building atmosphere.
There, the gaseous 1
components becorre a potential hazard to plant personnel, during i
inspection tours within the Reactor Building, and to the general public whenever the Reactor Building atmosphere is periodically purged to the environment.
When reactor coolant leakage occurs to the Auxiliary Building, it is collected in the Auxillary Building sump. The gases escaping from reactor coolant leakage within the Auxiliary Building will be collected in the Auxiliary and Fuel Handling Building exhaust ventilation system and discharged to the environment via the unit's Auxiliary and Fuel Handling Building vent. Since the majority of this leakage occurs within confined, separately ventilated cubicles within the Auxiliary Building, it incurs very little hazard to plant personnel.
In regard to the surveillance specification 4.2.7, the isolation valves may be tested at a reduced pressure is accordance with the Franklin Research Center Report titled " Primary Coolant System Pressure Isolation Valves for THI-1" (FRC Task 212) dated October 24, 1980, Section 2.2.2.
When reactor coolant leakage occurs to the nuclear services closed cooling water system, the leakage, both gaseous and liquid, is contained because the nuclear services closed cooling water system surge tank is a closed tank that is maintained above atmospheric pressure. The leakage would be detected by the nuclear services closed cooling water system monitor and by purge tank liquid level, both of which alarm in the control room. Since the nuclear services closed cooling water system's only potential contact with reactor coolant is in the sample coolers, it is considered not to be a hazard.
However, if reactor coolant leakage to this receptor occurred and the surge tank's relief valve discharged, radioactive q
gases could be discharged to the environment via the unit's
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auxiliary and fuel handling building vent.
i l
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Order dtd. 4/20/81 3-13a Amendment no 149
i Bases (Continued)
When reactor coolant leakage occurs to the intermediate cooling closed cooling water system, the leakage is indicated by both the intermediate cooling water monitor (RM-L9) and the intermediate cooling closed cooling water surge tank liquid level indicator, both of which alarm in the control room.
Reactor coolant leakage to this receptor ultimately could result in radioactive gas leaking to the environment via the unit's auxiliary and fuel handling building vent by way of the atmospheric vent on the surge. tank.
l Nhen reactor coolant leakage occurs to either of the decay heat closed cooling water systems, the leakage is indicated by the affected system's radiation monitor (RM-L2 or RM-L3 for system A and B, respectively) and surge tank liquid level indicator, all four of
-which alarm in the control room.
Reactor coolant leakage to this receptor _ ultimately could result in radioactive gas leaking to the environment via the unit's auxiliary and fuel handling building vent i
by way of the atmospheric vent on the surge tank of the affected I
system.
Assuming the existence of the maximum allowable activity in t'ie reactor coolant, a reactor coolant leakage rate of less than one gpm unidentified leakage within the reactor or auxiliary building or any of the closed cooling water systems indicated above, is a conserva-tive limit on what is allowable before the guide lines of 10 CFR 20 would be exceeded.
This is shown as fol, lows: if the specific activity of the reactor coolant is 130/E uCi/ml and the gaseous i
portion of it (as identified by.UFSAR Table 11.1-2) is discharged to l
l the environment via the unit's auxiliary and fuel handling building vent, the yearly whole body dose resulting from this activity at the site boundary, using an annual average X/Q = 4.$ x 10-6 sec/m3, is 0.34 rem. This may be compared with the 10 CFR 20 guideline of 0.5 rem / year whole body dose.
When the reactor coolant leaks to the secondary sides of either steam generator, all the gaseous components and a very small fraction of the ionic components are carried by the steam to the main condenser.
The gaseous components exit the main condenser via the unit's vacuum pump which discharges to the condenser vent past the condenser off-gas monitor. The condenser off-gas monitor will detect any radiation, above background, within the condenser vent.
However, buildup of radioactive solids in the secondary side of a steam generator and the presence of radioactive ions in the conden-sate can be tolerated to only a small degree. Therefore, the appear-ance of act.ivity in the condenser off-gas, or any other possible in-dications of primary to secondary leakage such as water inventories, j
condensate demineralized activity, etc., shall be considered positive indication of primary to secondary leakage and steps shall be taken to determine the source and quantity of the leakage.
l Amendment No. M M g W,149 3-14
3 3.1.12 Pressurizer Power Operated ' Relief Valve (PORV) and Block Valve
' Applicability Applies to the settir.;s, and conditions for isolation of the'PORV.
Objective
.To prevent the possibility of inadvertently overpressure:ing or depressurizing the Reactor Coolant System, i
Specification 3.1.12.1 The PORY shall not be taken out of service, nor shall it be isolated from the system (eu ept that the PORY may be isolated to limit leakage to within the. limits of Specifi-i cation 3.1.6) unless one of the following is in effect:
a.
High Pressure Injection Pump breakers are racked out or MU-V16A/B/C/D and MU-V217 are closed.
b.
Head of the Reactor Vessel is removed.
Tavg is above 332*F.
c.
l 3.1.12.2 The PORV settings shall be as follows, within the tolerances of 125 psi and 112*F:
Above 275'F - 2450 psig Below 275'F - 485 psig 1
3.1.12.3 If the reactor vessel head is installed and Tavg is <332*F, I
High Pressure Injection Pump breakers shall not be racked I
in unless:
l a.
MU-V16A/B/C/D and MU-V217 are closed, and b.
Pressurizer level is < 220 inches.
If pressurizer level is > 220 inches, restore level to < 220 inches within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
3.1.12,4 PORV and Block Valve The PORY and the associated block valve shall be OPERABLE i
during HOT STANDBY, STARTUP, AND POWER OPERATION:
a.
With the PORV inoperable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore the PORV to OPERABLE status or close the associated block valve and remove power from the block valve; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 N rs, b
With the PORY block valve inoperable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore the PORV block valve to OPERABLE status or close the PORY (verify closed) and remove power frv.n the PORV; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
Amendment No. % ) [, 149 3-18c
With either the PORY or block valve inoperable, c.
restore the inoperable valve to operable status prior to startup from the next cold shutdown unless the cold shutdown oscurs within 90 Effective Full Power Days (EFPD1 ef the end of the fuel cycle.
If a cold snutdown occurs within this 90-day period,. restore the inoperable valve to operable status prior to the startup for the next fuel cycle.
Bases If the PORY is removed from service, sufficient measures are incorporated to prevent severe.overpressurization by either-l eliminating the high pressure sources or flowpaths or assuring that the RCS is open to atmosphere.
In order to prevent exceeding leakage rates specified in T.S. 3.1.6., the PORY may be isolated.
The PORY setpoints are specified with tolerances assumed in the bases for Technical Specification 3.1.2.
With RCS temperatures less than 332*F and the makeup pumps running, l
the high pressure injection valves are closed and pressurizer level is maintained less than 220 inches to prevent severe overpressurization in the event of any single failure.
l Both the PORY and the PORY block valve should be operable during the HOT STAND 8Y, STARTUP, AND POWER OPERATION. If either the PORY or the PORY block valve are inoperable, the PORV discharge line should be isolated to prevent potential uncontrolled RCS depressurization.
For protection from severe overpressurization during HPI testing, refer to Section 4.5.2.1.c.
A endment No.[,149 l
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. Bases The safety function enhanced by this venting capability is core cooling.
For events beyond the present design basis, this venting capability will substantially increase the plants ability to deal with large quantitles of noncondensible gas which could interfere with. natural circulation (1.e., core cooling).
The reactor vessel head vent (RC-V42 & RC-V43 in series) provides the capability of venting noncondensible gases from the majority of the reactor vessel head as well as the Reactor Coolant hot legs (to the elevation of the top of the outlet nozzles) and cold legs (through vessel internals leakage paths, to the elevation of the top of the inlet nozzles).
This vent is routed to containment atmosphere.
Venting for the pressurizer steam space (RC-V28 and RC-V44 in series) has been provided to assure that the pressurizer is l
available for Reactor Coo.lant System pressure and volume control.
This vent is routed to the Reactor Coolant Drain Tank.
Additional venting capability has been'provided for the Reactor-Coolant hot leg high points (RC-V40A, B,'RC-41A, B), which normally cannot be vented through the Reactor vessel head vent or pressurizer steam-space vent.
These vents relieve to containment atmosphere through a rupture disk (set at low pressure).
i The above vent systems are seismically designed and environmentally qualified in accordance with the May 23, 1980 Commission Order and Memorandum per NUREG-0737 Item II.B.l.
The high point vents do not fall within the scope of 10 CFR 50.49, since the vents are not relied upon during or following any design basis event.
The power operated valves (2 in series in each flow path) which are powered from emergency buses fail closed on loss of gower. All vent valvt?
for the reactor vessel head vent, pressurizer vent and loop B high point vent are powered from the class IE "B" bus.
The vent va!/es for the loop A high point vent are powered from the class IE "A" bus.
The power operated valves are controlled in the Control Room.
The individual vent path lines are sized so that an inadvertent valve opening will not constitute a LOCA as defined in 10 CFR 50.46(c)(1).
These design features provide a high degree of assurance that these vent paths will be available when needed, and that inadvertent operation or failures will not significantly hamper the safe operation of the plant.
Amendment No.
149
4 3.3.3
' Exceptions to 3.3.2-shall be as follows:
E a.
Both core flood tanks shall be operable.at all times.
b.
Both the motor operated valves associated with the core flood tanks shall be fully opened at all times.
c.
One, reactor building cooling fan and associated I
cooling unit shall be permitted to be out-of-service for seven days.
3.3.4 prior to initiating maintenance on-any of.the components, the duplicate (redundant) component shal.1 be tested to assure operability.
Bases The requirs;ments of Specification 3.3.1 assure that, before the reactor can be.made critical, adequate engineered safety features are operable. Two engineered safeguards makeup pumps, two decay heat removal pumps and two decay heat removal coolers (along with their i
respective cooling water systems components) are specified.
- However, only one of each is necessary to supply emergency coolant to the reactor in the event of a loss-of-coolant accident. Both core flooding tanks are required because a single core flooding tank has insufficient inventory to reflood the core for hot and cold line breaks.
The operability of the barated water storage tank (BWST) as part of the ECCS ensures that a sufficient. supply of borated water is available for' injection by the ECCS in the event of a LOCA l
(Reference 2).
The limits on BWST minimup volume and boron concentration ensure that 1) sufficient water is available within containment to permit recirculation cooling flow to the core, and 2) the reactor will remain at least one percent suberitical at 70*F without any control rods in the core following mixing of the BWST and RCS water volumes (Reference 3).
i The contained water volume limit of 350,000 gallons includes an allowance for water not usable because of tank discharge location.
The limits on contained water volume, NaOH concentration and boron concentration ensure a pH value of between 8.5 and 11.0 of the solution sprayed within containment after a design basis accident.
The minimum pH of 8.5 assures that iodine will remain in solution while the maximum pH of 11.0 minimizes the potential for caustic damage to mechanical systems and components.
Redundant heaters maintain the borated water supply at a temperature greater than 40'F.
3-23 Amendment No.149
= - -
1 The post-accident reactor building emergency cooling may be accomplished by three emergency cooling units, by two spray systems, j
or by a combination of one emergency cooling unit and one spray system. The specified requirements assure that the required post-accident components are available.
l The iodine removal function of the reactor building spray system requires'one spray pump and sodium hydroxide tank contents.
The spray system utilizes common suction-lines with the decay heat removal system.
If a single train of equipment is removed from either system, the other train must be assured to be operable in each system.
When the reactor is critical, maintenance is allowed per Specification 3.3.2 and 3.3.3 provided requirements in Specification-3.3.4 are met which assure operability of the duplicate components.
The specif.ied maintenance times are a maximum. Operability of the specified components shall be based on the results of testing as l
required.by Technical Specification 4.5.
I An allowable maintenance period of up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> may be utilized if.
the operability of equipment redundant to that removed from service is demonstrated immediately prior to removal.
In the event that the need for emergency core cooling should occur, i
operation of one makeup pump, one decay heat removal pump, and both core flood tanks will protect the core.
In the event of a reactor coolant system rupture their operation will limit the peak clad temperature to less than 2,300'F and the metal-water reaction to that representing less than 1 percent of the clad.
Two nuclear service river water pumps and'two nuclear service closed cycle cooling pumps are required for normal operation. The normal operating requirements are greater than the emergency requirements
.following a loss-of-coolant.
a-REFERENCES.
(1) Updated FSAR, Sec'. ion 6.1 - Emergency Core Cooling System (2) Updated FSAR, Section 14.2.2.3 - Large Break LOCA (3) Updated FSAR, Section 14.2.2.1 - Fuel Handling Accident
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3.5.3 ENGINEERED SAFEGUARDS PROTECTION SYSTEM ACTUATION SETPOINTS Applicability:
This specification applies to the engineered safeguards protection system actuation setpoints.
Objective:
To provide for automatic initiation of the engineered safeguards protection system in the event of a breach of Reactor Coolant System integrity.
Specification:
3.5.3.1 The engineered safeguards protection system actuation setpoints and permissible bypasses shall be as follows:
Initiating Signal Function Setpoint High Reactor Building Reactor Building Spray 1 30 psig Pressure (1)
Reactor Building Isolation 1 30 psig High-Pressure Injection 1 4 psig low-Pressure Injection 1 4 psig j
i Start Reactor Building i
Cooling & Reactor Building Isolation 1 4 psig Low Reactor Coolant High Pressure Injection 2 1600(2) and System Pressure 1 500(3) psig Low Pressure Inject %n 1 1600(2) and 1 500(3) psig Reactor Building Isolation 1 1600 psig(2) 4.16 kv E.S. Buses 1
Undervoltage Relays Degraded Voltage (5)
Switch to Onsite Power Source and load shedding 3595 volts (4)
Degraded grid timer 10 sec (5)
Loss of voltage Switch to Onsite Power i
Source and load shedding 2400 Volts (6) l Loss of voltage timer 1.5 sec (7)
(1) May be bypassed for reactor building leak rate test.
l (2) May be bypassed below 1775 psig on decreasing pressure and is automatically reinstated before 1800 psig on increasing pressure.
(3) May be bypassed below 925 psig on decreasing pressure and is automatically reinstated before exceeding 950 psig on increasing pressure.
Amendment No. [ [ [ [ 1493 37
c.
.M n'
i (4) Minimum allowed setting is 3560 v. Maximum allowed setting is 3650 v.
(5) Minimum allowed time is 8 sec. maximum allowed time is 12 sec.
(6) Minimum allowed setting is 2200 volts, maximum allowed setting is 2860 volts (7) Minimum allowed time is 1.0 second, maximum allowed time is 2.0 seconds.
Bases High Reactor Building Pressure, l
The basis for the 30 psig and 4 psig setpoints for the high pressure signal is-to establish a setting which would be reached in adequate i
time in the event of a LOCA, cover a spectrum of break sizes and yet be far enough above normal operation maximum internal pressures to prevent spurious initiation.
Low Reactor Coolant System Pressure The basis for the 1600 and 500 psig low reactor coolant pressure setpoint for higa and low pressure injection initiation is to establish a value which is high enough such that protection is provided for the entire spectrum to break sizes and is far enough below normal operating pressure to prevent spurious initiation.
Bypass of HPI below 1775 psig and LPI below 925 psig, prevents ECCS actuation during normal system cooldown.
4.16 XV ES Bus Undervoltage Relays The basis for the degraded grid voltage relay setpoint is to protect the safety related electrical equipment from loss of function in the event of a sustained degraded voltage condition on the offsite power system.
The timer setting prevents spurious transfer to the onsite so.rce for transient conditions.
The loss of voltage relay and timers detect loss of offsite power condition and initiate transfer to the onsite source with minimal time delay.
l l
i l
3-37a Amendment No. M, M M [ 149 i
m-----------e.-----
a---
y i
~ 3.15 AIR TREATMENT SYSTEMS L
3.15.1 EMERGENCY CONTROL ROOM AIR TREATMENT SYSTEM Applicability
]
Applies to the emergency control room air treatment system l
and its associated filters.
4 Objective l
To specify minimum availability and efficiency for the emergency.
L control room air treatment system and its associated filters.
Specifications 1
3.15.1.1 Except as specified in Specification 3.15.1.3 below, both l
emergency treatment systems, AH-E18A fan and associated filter AH-F3A and AH-E188 fan and associated filter AH-F3B shall be operable at all times, per the requirements of Specification 3.15.1.2 below; when containment integrity is required and when irradiated fuel handling operations are in progress.
3.15.1.2 a.
The results of the in-place 00P and halogenated hydrocarbon tests at design flows on HEPA filters and charcoal absorber banks shall show <0.05% 00P penetration and <0.05% halogenated hydrocarbon penetration, except that.the DOP test will be conducted with profilters installed.
b.
The results of laboratory carbon sample analysis shall l
show290% radioactive methyl iodide decontamination efficiency when tested at 125'F, 95% R.H.
The fans AH-EISA and B shall each be shown to operate c.
within 1 4000 CFM of design flow (40,000 CFM).
3.15.1.3 From and after the date that one control room air treatment system is made or found to be inoperable for any reason, reactor operation or irradiated fuel handling operations are permissible only during the succeeding 7 days provided the redundant system is demonstrated to be operable per 4.12.1.1 and 4.12.1.3 within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and daily thereafter.
3.15.1.4 From the date that both control room air treatment systems are made or found to be inoperable or if the inoperable 3
system of 3.15.1.3 cannot be made operable in 7 days, irradiated fuel handling operations shall be terminated in 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and reactor shutdown shall be initiated and the reactor shall be in cold shutdown within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> l
Amendment No. M g, [ 149 1
1
,( h.:
3.15.2 REACTOR BUILDING PURGE AIR' TREATMENT SYSTEM Applicability Applies to the reactor building purge air treatment system and its associated filters.
Objective To specify minimum availability and efficiency for the reactor building purge air treatment system and its associated filters.
Specification i
3.15.2.1 Except as specified in Specification 3.15.2.3 below, the Reactor Building Purge Air Treatment System filter AH-F1 shall be operable as defined by the Specification below at all times when containment integrity is required unless the Reactor Building purge isolation valves are closed.
3.15.2.2 a.
The results of the in-place 00P and halogenated l
hydrocarbon tests at maximum available flows on HEPA filters and charcoal adsorber banks for AH-F1 shall i
show 1;ss than 0.05% DOP penetration and less than
~0.05% halogenated hydrocarbon penetration; except that the DOP test will be conducted with prefilters installed.
b.
The results of laboratory carbon sample analysis for l
the reactor building purge system filter carbon shall show greater than or equal to 90% radioactive methyl' iodide decontamination efficiency when tested at 250*F,95% R.H.
3.15.2.3 From and after the date that the filter AH-Fl in the reactor building purge system is made or found to be inoperable as defined by Specification 3.15.2.2 above, the Reactor Building purge isolation valves shall be closed until the filter is made operable.
I Bases The Reactor Building Purge Exhaust System filter AH-F1 while normally used to filter all reactor building exhaust air.
It is necessary to demonstrate operability of these filters to assure readiness for service if required to mitigate a fuel handling I
accident in the Reactor Building and to assure that 10CFR50 Appendix I limits are met.
Reactor Building purging is required to be terminated if the filter is not operable.
3-62a M [J[149 Amendment No.
l l
l
a
~3.22 RADIOACTIVE-EFFLUENTS' i:
3.22.1 LIQUID EFFLUENTS 3.22.1'.1 CONCENTRATION LIMITING CONDITION FOR OPERATION-3.22.1.1 ' The concentration of_ radioactive material release'd at I
anytime from the unit to unrestricted areas (see Figure 5-3) shall l
be_ limited to the concentrations specified in 10 CFR Part 20, Appendix B, Table II Column 2 for radionuclides other than dissolved or entrained noble gases.
For dissolved or entrained noble gases, the concentration shall be limited to 3 x 10-3 uCf/cc total activity.
APPLICABILITY: At all times ACTION:
With the concentration of radioactive material released from the a.
unit _to unrestricted areas exceeding the above limits, imediately restore concentration within.the above limits.
b.
If.acticn "a" cannot be met, then be in:
- 1. At least HOT STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />,
- 2. At least HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and l
- 3. At least COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
l BASES-This specification is provided to ensure that the concentration of radioactive materials released in liquid waste effluents from the unit to unrestricted areas will be less than the concentration levels specified in 10'CFR Part 20, Appendix B Table II.
This limitation provides additional assurance that the levels of radioactive materials in bodies of water outside the site will not result in exposures with (1) the Section II.A design cbjectives of Appendix I,10 CFR Part 50, to a MEMBER OF THE PUBLIC and (2) the limits of 10 CFR Part 20.106 (e) to the population.
The concentration limit for noble gases is based upon the assumption the Xe-135 is the controlling radioisotope and its MPC in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication 2.
3-106 Amendment No.
k.149
____ w _ _ _ _ _ -
RADI0 ACTIVE EFFLUENTS
]
DOSE LIMITING CONDITION FOR OPERATION i
3.22.1.2 The dose or' dose commitment to a MEMBER OF.THE PUBLIC from I
I radioactive materials in liquid effluents released from the unit to the site boundary (see Figure 5-3) shall be limited:
l l
- a. During any calendar quarter to < 1;5 mrem to the i
total body and to 5 5 mrem to aiiy organ.
]
- b. During any calendar year to 5--3 mrem 'to the total body and to g 10 mrem to any organ.
APPLICABILITY: At all times ACTION:
l
- a. With the calculated dose from the release of radioactive 4
materials in liquid effluents exceeding any_ of the above limits, prepare and submit to the NRC Region I Administrator within 30 days, a Special Report which identifies-the
.l cause(s) for exceeding the limit (s) and defines the-corrective actions to be taken to reduce the releases of 1
radioactive materials in liquid effluents during the remainder of the current calendar quarter and during the subsequent 3 calendar cyarters so that the cumulative dose or dose commitment to any individual from such releases during
.i these four calendar quarters is within 3 mrem to the total body and 10 mrem to any organ.
This Special Report shall also include (1) the result of radiological 1' analyses of the drinking water source, and (2) the radiological impact on l
finished drinking water supplies with regard to the requirements of 40 CFR 141, Safe Drinking Water Act.
BASES This specification is provided to implement the requirements of Sections II. A, III.A, and IV. A of Appendix I,10 CFR Part 50.
The Limiting Condition for Operation implements the guides set forth in Section II.A of Appendix I.
The ACTION statements provide the l
required operating flexibility and at the same time implement the a
guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in liquid effluents will be kept "as low as is reasonably achievable". Also, for fresh water sites with drinking water supplies which can be potentially affected by i
plant operations, there is reasonable assurance that the operation l
of the facility will not result in radionuclides concentrations in i
the finished drinking water that are in excess of the requirements of 10 CFR 20. The dose calculations in the ODCM implement 3-107 J
J2[,k149 Amendment No.
l
\\
l a
b l
.,. t RADIOACTIVE EFFLUENTS LIQUID RADWASTE TREATMENT' SYSTEM LIMITING CONDITION FOR OPERATION 3.22.1.3 The appropriate portions of the liquid radwaste treatment system-shall be used to reduce the radioactive materials in liquid wastes prior _ to their discharge when the pro-jected doses due to the liquid effluent from the unit to unrestricted areas (see Figure 5-3) would exceed 0.06 mrem l
j to the total body or 0.2 mrem to any organ in any calendar i
month.
APPLICABILITY: At all times ACTION:
a.
With radioactive liquid waste being discharged without treatment and in excess of the above limits, prepare and submit to the NRC Region I Administrator within 30 days, a Special Report which includes the following information:
4 1.
Explanation of why liquid radwaste was being discharged 4
without treatment, identification of any inoperable equipment or subsystems, and the. reason for in-operability,-
2.
Action (s) taken to restore the inoperable equipment to OPERABLE status, and,
-3.
A summary description of action (s) taken to prevent a recurrence.
BASES The requirement that the appropriate portions of this system be used, when specified, provides assurance that the releases of radioactive materials in liquid effluents will be kept as low
-as is reasonably achievable.
This specification implements the i
requirements of 10 CFR Part 50.36a, General Design Criterion 60 1
of Appendix A to 10 CFR Part 50 and the design objective given l
in Section II.D of Appendix I to 10 CFR Part 50.
The intent of
- Section II.D. is to reduce effluents to as low as is reasonably achievable in a cost effective manner.
This LCO satisfies this intent by establishing a dose limit which is a small fraction (25%)
of Section II.A of Appendix I, 10 CFR Part 50 dose requirements.
This margin, a factor of 4, constitutes a reasonable reduction.
i l
3-109 Amendment No.
k d 149
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TABLE 4.1-2 MINIMUM: EQUIPMENT TEST FREQUENCY Item Test Frequency 1.
Control Rods Rod drop times of all Each Refueling shutdown full length rods 2.
Control. Rod Movement of each rod Every'two weeks, when Movement reactor is critical
- 3. ' Pressurizer Setpoint*
50% each refueling Safety Valves period 4.
Main Steam Setpoint 25% each refueling Safety Valves period l
5.
Refueling System Functional Start of each Interlocks refueling period 6.
Main Steam (See Section 4.8)
Isolation Valves 7.
Reactor Coolant Evaluate Daily, when reactor System Leakage coolant system temperature is greater.than 525'F 8.
Deleted 9.
Spent Fuel Functional Each refueling period Cooling System prior to fuel handling
- 10. Intake Pump (a) Silt Accumulation-Each refueling period House Floor Visual inspection (Elevation of Intake Pump 262 ft. 5 in.)
House Floor (b) Silt Accumulation Quarterly Measurement of Pump House Flow
- 11. Pressurizer Block Functional **
Quarterly Valve (RC-V2)
- The setpoint of the pressurizer code safety valves shall be in accordance with ASME Boller and Pressurizer Vessel Code,Section III, Artic1s 9, Winter, 1968.
- Function shall be demonstrated by operating the valve through one complete cycle of full travel.
4-8
[
149 j
Amendment No.
a.
4.5 EMERGENCY LOADING SEQUENCE AND POWER TRANSFER. EMERGENCY CORE COOLXNG SYSTEM & REACTOR BUlLDING COOLING SYSTEM PERIODIC TESTING 4.5.1 Emergency Loading Sequence Applicability: Applies to periodic testing requirements for safety actuation systems.
Objective:
To verify that the emergency loading sequence and automatic power transfer is operable.
Specifications:
4.5.1.1 Sequence and Power Transfer Test
- a. During each refueling interval, a test shall be conducted to demonstrate that the emergency loading sequence and power transfer is operable.
- b. The test will be considered satisfactory if the following pumps and fans have been successfully started and the following valves have completed their travel on preferred y
power and transferred to the emergency power as evidenced by the control board component operating lights, and either the station computer or pressure / flow indication.
-M.
U. Pump
-D. H. Pump and D. H. Injection Valves and D. H. Supply Valves
-R. B. Cooling Pump
-R. B. Ventilators
-D. H. Closed Cycle Cooling Pump
-N. S. Closed Cycle Cooling Pump
,0. H. River Cooling Pump
-N. S. River Cooling Pump
-D. H. and N. S. Pump Area Cooling Fan
-Screen House Area Cooling Fan
-Spray Pump. (Initiated in coincidence with a 2 out of 3 R. B.
30 psig Pressure Test Signal.)
-Motor Driven Emergency Feedwater Pump
- c. Following successful transfer to the emergency diesel, the diesel generator breaker will be opened to simulate trip of the generator then reclosed to verify block load on the reclosure.
4.5.1.2 Seauence Test
- a. At intervals not to exceed 3 months, a test shall be conducted to demonstrate that the emergency loading sequence is operable, this test shall be performed on either preferred power or emergency power.
- b. The test will be considered satisfactory if the pumps and fans listed in 4.5.1.lb have been successfully started and the valves listed in 4.5.1.1b have completed their travel as evidenced by the control board component operating lights, and either the' station computer or pressure / flow indication.
AmendmentNo.[
149
8
~N.
g l'
l i
Bases i
- 1 The emergency loading sequence and automatic power transfer controls the operation of the pumps associated with the j
emergency core cooling system and Reactor Building cooling system.
k REFERENCES (1)
Updated FSAR Section 7.0 (2)
Updated FSAR Section 1.4 (3) Speci fication 4.6.1b l
(
{
O Amendment [,149 4-40
i,..
4.5.2-
-EMERGENCY CORE C00LENG SYSTEM Applicability: -Applies.to periodic testing requirement for.
emergency core cooling systems.
' Objective:
To verify. that the emergency core cooling systems are ' operable.
_. Specification:
[
4.5.2.1
.High Pressure Injection.
L
- a. During each refueling. interval and following maintenance or modification that affects system flow characteristics, system pumps and system high point vents shall be vented, and a.
system test shall be conducted to demonstrate that the system '
is ' operable.
l The M. U. Pump and its required supporting auxiliaries will be started manually by the operator and a test signal will be applied to the high pressure injection. valves MU-V-16A, B, C, D to demonstrate actuation of the high pressure injection system for emergency core cooling operation,
- b. The test will be considered satisfactory if the valves have completed their travel and the M. U. Pumps are running as evidenced by the control board component operating lights.
1 Minimum acceptable injection flow must be greater than or equal to 500 gpm per HPI. pung when pump discharge pressure is 600 psig or greater (the pressure between the pump and flow limiting device) and when the RC pressure is equal to or less than 600 psig.
- c. Testing which requires HPI flow thru MU-V16A, B, C, D shall be conducted only under either of the following conditions:
- 1) T avg shall be greater the 332*F.
- 2) Head of the Reactor Vessel shall be removed.
i 4.5.2.2 Low Pressure Injection
- a. During each refueling period and following maintenance or modification that affects system flow characteristics, system pumps and high point vents shall be vented, and a system test shall be conducted to demonstrate that the system is operable.
The auxiliaries required for low pressure injection are all included in the emergency loading sequence specified in 4.5.1.
I
- b. The test will be considered satisfactory if the decay heat pumps listed in 4.5.1.1b have been successfully started and the decay heat injection valves and the decay heat supply valves have completed their travel as evidenced by the control board component operating lights.
Flow shall be verified to be equal or greater than the flow assumed in the Safety Analysis for the single corresponding RCS pressure used in the test.
AmendmentNo.%
149 4-41
o -
c' When the. Decay Heat System'is required to'be operable, the co# rect position of DH-V-19A/B shall be verified by obser-vation within fourt hours of each valve stroking operation or.
valve maintenance which effects the position indicator.
4.5.2.3
. Core Flooding-g
- a. During each refueling period, a system test shall be conducted to demonstrate proper operation of the system.
During depressurization of the Reactor Coolant System, l
verification shall be made that the check and isolation valves in the core cooling flooding tank discharge lines operate properly,
- b. The test will be considered *. satisfactory.if control board indication.of core flooding tank level verifies that all valves have opened.
l 4.5.2.4 Component Tests
- a. At intervals.not to exceed 3 months, the components required for emergency core cooling will be tested,
- b. The test will be considered satisfactory if the pumps and fans hcve been successfully started and the valved have completed their travel as evidenced by the control board component operating lights, and either the station computer or pressure / flow indication.
Bases The emergency core cooling systems are the principal reactor safety features in the event of a loss of coolant accident. The removal of heat from the core provided by these systems is designed to limit core damage.
The low pressure injection pumps are tested singularly for operability by opening the borated water storage tank outict valves and the bypass valves in the borated water storage tank fill line.
This allows water to be pumped from the borated water storgage tank through each of the injection lines and back to the tank.
l The minimum acceptable HPI/LPI flow assures proper flow and flow split between injection legs.
]
Hith the reactor shutdown, the valves in each core flooding lines are checked for operability by reducing the reactor coolant system pressure until the indicated level in the core flood tanks verify that the check and isolation valves have opened.
4-42 Amendment No.
149
y,.
., ~
r
- b. Reactor Building Cooling and Isolation Systems
- l. During each refueling period., a system test shall be conducted L
.to demonstrate proper operation of the system. A test signal l
will. actuate the R.B. emergency cooling system valves to demonstrate _ operability of the coolers.
- 2. The test will be considered satisfactory if the valves have completed their expected travel as evidenced by the control board component operating lights, and either the station computer or local verification.
l 4.5.3.2 Component Tests
- a. At intervals not to exceed three months, the components I
required for reactor building cooling and isolation will be tested,
- b. The test will be considered satisfa torv if the valves have completed their expected' travel as evidenced by the control board component operating lights, and either the station computer or local verification.
Bases The reactor building cooling and isolation systems and reactor building spray system are designed to remove the heat in the containment atmosphere to prevent the building pressure from exceeding the design pressure.
The delivery capability of one reactor building spray pump at a time can be tested by opening the valve in the line from the borated water storage tank. opening the corresponding valve in the test line, and starting the corresponding pump.
With the pumps shut down and the borated water storage tank outlet closed, the reactor building spray injection valves can each be opened and closed by the operator action. With the reactor building spray inlet valves closed, low pressure air can be blown through the test connections of the reactor building spray nozzles to demonstrate that the flow paths are open.
The equipment, piping, valves and instrumentation of the reacter building cooling system are arranged so that they can be visually Inspected. -The cooling units and associated piping are located outside the secondary concrete shield.
Personnel can enter the reactor building during power operations to inspect and maintain this equipment.
The reactor building fans are normally operating periodically, constituting the test that these fans are operable.
Reference f
(1) FSAR, Section 6 I
h 4-44 Amendment No.
149
4-4.6 EMERGENCY POWER SYSTEM PERIODIC TESTS Appl'es-to periodic testing and surveillance Applicability:
i requirement of.the emergency power system.-
Objective:
To verify that the emergency power system will respond promptly and properly when required.
Specification:
The following tests and surveillance shall be performed as stated:
4.6.1 Diesel Generators a.
Manually-initiate start of the diesel generator, followed by manual synchronization with other power sources and assumption of load by the diesel generator up to the name-plate rating (3000 kw).
This test will be conducted every month on each diesel ger.erator.
Normal plant operation will not be effected.
b.
Automatically start and loading the emergency diesel l
generator in accordance with Specification 4.5.1.1.b/c including the following.
This test will be conducted every refueling interval on each diesel generator.
(1) Verify that the diesel generator starts from ambient condition upon receipt of the ES signal and is ready to load in 1 0 seconds.
1 (2) Verfiy that the diesel block loads upon simulated loss of offsite power in 130 seconds.
(3)
The diesel eperates with the permanently connected and auto connected load for 25 minutes.
(4) The diesel engine does not trip when the generator breaker is opened while carrying emergency loads.
(5) The diesel generator block loads and operates for 15 minutes upon reciosure of the diesel generator breaker.
1 c.
Each diesel generator shall be given an inspection at least annually in accordance with the manufacturer's recommendations for this class of stand-by service.
4.6.2 Station Batteries, a.
The voltage, specific gravity, and liquid level of each cell will be measured and recorded monthly.
b.
The voltage and specific gravity of a pilot cell will be measured and recorded weekly.
c.
Each time data is recorded, new data shall be compared with old to detect signs of abuse or deterioration.
Amendment No.
149 4-46
= _ _ _ _ _ _ _ _
a:
.r 4.12 AIR TREATMENT SYSTEM 4.12.1-EMERGENCY CONTROL ROO AIR TREATMENT SYSTEM Applicability Applies to the emergency control room air treatment system and
-associated components.
Objective To verify that this system and associated components will be able to perfctm its design functions.
Specification 4.12.1.1 At least every refueling interval or once every 18 months, whichever comes first, the pressure drop across the combined HEPA filters and charcoal adsorber banks of AH-F3A and 38 shall be demonstrated to be less than 6 inches of water at system design flow rate (110%).
4.12.1.2 a.
The tests and sample analysis required by l'
Specification 3.15.1.2 shall be performed initially and at least once per year for standby service or after every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of system operation and following significant painting, steam, fire or chemical release in any ventilation zone communicating with the system that could contaminate the HEPA filters or charcoal adsorbers, b.
00P testing shall be performed after each complete or l
partial replacement of the HEPA filter bank or after any structural maintenance on the system housing which could affect the HEPA filter bank bypass leakage.
c.
Halogenated hydrocarbon testing shall be performed l
1 after each complete or partial replacement of the charcoal adsorber bank or after any structural maintenance on the system housing which could effect the charcoal adsorber bank bypass leakage, d.
Each AH-E18A and B (AH-F3A and B) fan / filter circuit i
shall be operating at least 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> every inonth.
4.12.1.3 At least once per refueling interval or once every 18 months, whichever comes first, automatic initiation of the Control Building isolation and recirculation Dampers AH-D28, 37, 39, and 36 shall be demonstrated as operable.
4.12.1.4 An air distribution test shall be performed on the HEPA filter bank initially, and after any maintenance or testing that could affect the air distribution withir: the system.
The air distribution across the HEPA filter bank shall be uniform within 120%. The test shall be performed at 40,000 cfm (210%) flow rate.
Amendment No.
149 4-55
,e
- )
4.12.2' REACTOR BUlLDING PURGE AIR TREATMENT SYSTEM
(
Applicability:' Applies to the reactor building purge air treatment system and associated components.
Objective:
To verify that this system and associated components L
will be able to perform its design functions.
Specification 4.12.2'.1 At least once per refueling interval or once per 18 months, whichever comes first it shall be demonstrated that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 6 inches of water at system design flow rate (110%).
4.12.2.2 a.
The tests and sample analysis required by Specification l-3.15.2.2, shall be performed initially, once per re-fueling interval or 2 years, whichever comes first, or within 30 days prior to the movement of irradiated fuel in containment and following significant paint-Ing, steam, firc, or chemical release in any ventila-1 tion zone communicating with the system that could contaminate the HEPA filters or charcoal adsorbers.
b.
00P testhg shall be performed after each complete or l
partial replacement of a HEPA filter bank or after any structural maintenance on the system housing which could affect HEPA frame bypass leakage.
c.
Halogenated hydrocarbon testing shall be performed I
after each complete or partial replacement of a charcoal adsorber bank or after any structural maintenance on the system housing which could affect the charcoal adsorber bank bypass leakage, d.
The DOP and halogenated hydrocarbon testing shall be l
performed at the maximum available flow considering physical restrictions, i.e., purge valve position, and 3
i gaseous radioactive release criteria.
e.
Each refueling, AH-E7A&B shall be shown to operate within 2 5000 cfm of design flow (50,000 cfm) with purge valves fully open.
4.12.2.3 An air distribution test shall be performed on the HEPA filter bank initially and after any maintenance or testing that could affect the air distribution within the system.
The air distribution across the HEPA filter bank shall be uniform within 120%. The test shall be performed at 50,000 cfm (110%) flow rate with purge valves fully open.
Bases j
Pressure drop across the combined HEPA filters and charcoal l
adsorbers of less than 6 inches of water at the system design flow rate will indicate that the filters and adsorbers are not clogged by excessive amounts of foreign matter. Pressure drop should be
{
determined at least once every refueling interval to show system performance capability.
J l
/
4 55b Amendment No.
JO[ 149
o i
4.15.
f1AIN STEAM SYSTEM INSERVICE INSPECTION-Applicability This technical specification applies to the inservice inspection of those welds in the main steam system identified as Numbers 3, 4, and 5 of Figure 6, Supplement 2, Part IX and Number 3 of Figure 9, Supplement 2, Part IX.
Objective The objective of this inservice inspection program is to provide assurance of the continuing integrity of that portion of the main steam system'in which a. postulated failure would produce pressures in excess of the compartment wall and/or slab capacities.
Specification j
4.15.1.
The four weld joints identified above shall be 100 percent ultrasonically inspected in accordance with the ASME Code,Section XI, Rules for-Inservice Inspection of Nuclear Reactor Coolant Systems dated January 1,1970 as modified by the Winter 1970 Addenda and the provisions of Appendix IX-3400 of Section III of the ASME Boiler and Pressure Vessel Code.
Insper'. ions are to be performed prior to startup and subsequently at 3-1/2 year intervals (or i
nearest refueling outage),
i Prior to initial plant operation a preoperational inspection of the identified weld joints will be performed and any data acquired will be recorded to form a baseline on which to compare results of subsequent inspections.
Upon completion of several inspection cycles, the.
techaical benefit of the inspection program frequency will i
be reviewed.
The conclusions of this review shall be I
submitted to the NRC for evaluation.
l Bases Calculations reveal that postulated breaks in the main steam lines I
I al: the containment penetrations in small compartments No. 2 and No.
5-could produce pressures in excess of wall and/or slab capacities.
I l
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4-58 l
Amendment No. 149 l
'6:
- d.
-4.16
. REACTOR'!NTERNALS VENT VALVES SURVEILLANCE Applicability Applies to Reactor Internals Vent Valves.
Objective To verify.that no reactor internals vent valve is stuck in the open position and that each valve continues to exhibit freedom of movement.
Specification Item Test Frequency 4.16.1 Reactor Internals Demonstrate Operability Each Refueling Vent Valves By:.
Shutdown l
- a. Conducting a remote visual inspection of visually accessible sur-faces of the valve body and disc sealing faces and evaluating any observed surface irregu-larities.
- b. Verifying that the valve is not stuck in an open position, and
- c. Verifying through manual actuation that the valve is fully open with a force of < 400 lbs. (applied vertically upward).
Bases Verifying vent valve freedom of movet.<nt insures that coolant flow does not bypass the core through rhactor internals vent valves during operation and therefore insures the conservatism of Core Protection Safety limits as delineated in Figures 2.1-1 and 2.1-3.
and the flux / flow trip setpoint.
4-59 AmendmentNo.J8/. 149
o i
.]
SHOCK SUPPRESSORS (SNUBBERS)
SU9VEILLANCE REQUIREMENTS (Continued) a
- c. Refueling Outage Inspections At least once each refueling cycle during shutdown, a visual inspection shall be performed of all safety.related j
snubbers attached to sections _of. safety systems piping l
that have. experienced unexpected, potentially damaging j
transients as determined from a review of operational data and a visual inspection of-the systems.
- d. Visual Inspection Acceptance Criteria l
Visual inspections shall verify: (1) that there are no
{
visible indications of damage or impaired operability and (2) attachments to the foundation or supporting structure are secure.
Snubbers which appear inoperable as a result j
of visual inspections may be determined OPERABLE for the purpose of establishing the next visual inspection in-l terval, provided that: (1) the cause of the rejection is clearly established and remedied for that particular snubber and for other snubbers that may be generically susceptible, and (2) the affected snubber is functionally tested in the as found condition and determined OPERABLE per Specification 4.17-If.
When the reservoir outlet port of a snubber is found to be uncovered by fluid, the snubber shall only be declared operable if functional testing in both extension and ietraction directions is satisfactory and an engineering evaluation concludes that this snubber is operable.
- e. functional Tests
- At least once each refueling interval during shutdown, a representative sample of snubbers shall be tested using one of the following sample plans.
The sample plan shall be selected prior to the test period and cannot be changed during the test period.
The NRC Regional Administrator shall be notified in writing of the sample plan selected prior to the test period, or the semple plan used in the prior test period shall be used:
- 1) At least 10% of the total each type of snubber in use in the plant shall be functionally tested either in-place or in a bench test.
For each snubber of a type that does not meet the functional test acceptance l
criteria of Specification 4.17.lf, an additional 10%
of that type of snubber shall be functionally tested
)
until no more failures are found or until all snubbers of that type have been functionally tested; or 1
1 The four 550,000 lb reactor coolant pump snuhbers are not included.
The functional test program for reactor coolant pump snubbers is implemented in accordance with the schedule and other requirements of the snubber testing program.
Amendment No. g J
, %, 149 d
I 1
p.
SHOCK SUPPRESSORS (SNUBBERS)
SURVEILLANCE REQUIREMENTS (Continued)
- 1. Snubber Seal Service Program A snubber seal service life program shall be developed' whereby the seal service life of hydraulic snubbers is J
monitored to ensure that the service life is not exceeded-between surveillance inspections.
The designated service life for the various seals shall be established based on engineering information. -The seals shall be replaced so f
that the indicated service life will not be exceeded i
during a period when the snubber is required to be OPERABLE.
The seal replacements shall be documented and the documentation shall be retained in accordance with-l Specification 6.10.2.m.
I l
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l 4-64 l
l Amendment No. lg kt6,149 j
c l
l t
r e
2.
A sef ss:ic occurrence greater than the Operating Basis Earthquake.
3.
A loss of coolant accident requiring actuation of the engineering safeguards, or 4.
A najor rain steam line or feedwater line break.
4.19.4 Acceptance Criteria
- a. As used in this Specification:
1.
Imperfection means an exception to the dimensions, finish or contour of a tube from that required by fabrication drawing or specifications.
Eddy current testing indications below 20% of the nominal tube wal thickness, if detectable, may be considered as imperfections.
2.
Degradation means a service-induced cracking, wastage, wear or general corrosion occurring on either inside l
or outside of a tube.
3.
Degraded Tube means a tube containing imperfections 20% of the nominal wall thickness caused by degradation.
4.
% Degradation means the percentage of the tube wall thickness affected or removed by degradation.
5.
Defect means an imperfection of such severity that it I
exceeds the repair limit.
A tube'containing a defect I
is defective.
l 6.
Repair Limit means the extent of degradation at or beyond which the tube shall be repaired or removed from service because it may become unserviceable prior to the next inspection.
This limit is equal to 40% of the nominal tube wall thickness.
l 7.
Unserviceable describes the condition of a tube if it leaks or contains a defect large enough to affect its structural integrit,y in the event of an Operating l
Basis Earthquake, a loss of coolant accident, or a i
steam line or feedwater line break as specified in 4.19.3.c., above.
l 8.
Tube Inspection means an inspection of the steam generator tube from the bottom of the upper tubesheet completely to the top of the lower tubesheet, except as permitted by 4.19.2.b.2, above.
Amendment No.
, 149
?
s 4.19.4 Acceptance Criteria (Continued)
- b. The steam generator shall be determined OPERABLE after completing the corresponding actions (removal from service by plugging, or repair by kinetic expansion, sleeving, or uther methods, of all tubes exceeding the repair limit and all tubes containing throughwall cracks) required by Table 4.19.2.
4.19.5 Reports a.
Following the completion of each inservice inspection of steam generator tubes, the number of tubes repaired
)
or removed from service in each steam generator shall j
be reported to the NRC within 15 days.
I b.
The complete results of the steam generator tube inservice inspection shall be reported to the NRC j
within 3 months following completion of the inspection.
This report shall include:
1.
Num5er and extent of tubes inspected.
2.
Location and percent of wall-thickness penetration for each indication of an imperfection.
3.
Identification of tubes repaired or removed from service.
c.
Results of steam generator tube inspections which fall j
into Category C-3 require notification in accordance with 10 C' R 50.72 prior to resumption of plant opers; tar.
The written followup of this report shall provide a description of investigations conducted to 4
determine the cause of the tube degradation and corrective measures taken to prevent recurrence in accordance with 10 CFR 50.73.
Bases The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be maintained.
The proavam for inservice inspection of steam generator tubes is based on modification of Regulatory Guide 1.83, Revision 1.
In-service inspection of steam generator tubing is essential in order to maintain surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions.
Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken.
Amendment No. M J M pf, I M J d, 149 4-81
o o
Bases (Continued)
The Unit is expected to be operated in a manner such that the primary and secondary coolant will be maintained within those chenistry limits found to result in negligible corrosion of the stean generator tubes.
If the primary or secondary coolant chemistry is not maintained within these chemistry limits, localized corrosion may likely result.
The extent of stears geurator tube leakage due to cracking would be limited by the.seccidary coolant activity, Specification 3.1.6.3.
The extent of cracking during plant operation would be limited by the limitation of total steam generator tube leakage between the pri: nary coolant system and the secondary coolant system (primary-to-secondary leakage = 1 gpm).
Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which
.the leaking tubes will be located and repaired or removed from service.
Wastage-type defects are unlikely with proper chemistry treatment of the primary or the secondary coolant.
However, even if a defect would develop in service, it will be found during scheduled in-service steam generator tube examinations.
Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect degradation that has penetrated 20% of the original tube wall thickness.
Removal from service by plugging, or repair by kinetic expansion, sleeving, or other methods, will be required for degradation equal to or in excess of 40% of the tube nominal wal). thickness.
I Where experience in similar plants with similar water chemistry, as documented by USNRC Bulletins / Notices, indicate critical areas to be inspected, at least 50% of the tubes inspected should be from these critical areas.
First sample inspections sample size may be modified subject to NRC review and approval.
f Amendment No. M [ I g, I49 4-82 I
o-- -
, au r
-5.0 DESfGNFEATURES 5.1' SITE I
Applicability Applies to the' location and extent of the exclusion boundary, restricted area, and low populat1on zone.
Objective To define the above by location and distance description.
Specification 5.1.1 The Three Mile Island Nuclear Station Unit 1 is located in an area of low population density about ten miles southeast of Harrisburg, PA.
It is in Londonderry Township of Dauphin l
County, Pennsylvania, about two and one-half miles north of the southern tip of Dauphin County, where Dauphin is coterminal with York and Lancaster Counties. The station is located on an island approximately three miles in length situated in the Susquehanna River upstream from York Haven i
Dam.
Figure 5-1 is an extended plot plan of the site showing l
the plant orientation and immediate surroundings. The Exclusion. Area as defined in 10 CFR 100.3, is a 2,000 ft.
radius, including portions of Three Mile Island, the river surface around it, and a portion of Shelley Island, which is j
owned by Met Ed. The minimum. distance of 2,000 ft. occurs on the shore of the mainland in a due easterly direction from the plant as shown on Figure 5-1 for the Exclusion Area.
Figure 5-3 showing the physical' location of the fence defines the " Restricted Area" surrounding the plant. The minimum distance of the " Restricted Area" is approximately 560 feet and is from the centerline of the TMI Unit 2 Reactor Building i
to a point on the westerly shoreline of Three Mile Island.
I The minimum distance to the outer boundary of the low l
population zone is two miles as shown on T.S. Figure 5-2, which also depicts the site topography for a radius of five miles.
T.S. Figure 5-3 depicts the locations of gaseous effluent release points and liquid effluent outfalls (as tabularized on page 5-10), and the meteorological tower location (designated as ' weather tower' on the figure).
5-1 Amendment No. [ [ 149
)
1 1
- a
- r ELEVATIONS FOR GASEOUS EFFLUENT RELEASE POINTS
-(See Figure 5-3) l j
Unit 1 Stack 483' 7" Unit 1 Turbine Building 425' 4"
{
l Unit 1 Fuel Handling Building 348' ESF Vent Stack LOCATIONS OF LIQUID EFFLUENT OUTFALLS PURSUANT TO NPDES (See Figure 5-3)
Outfall No.
Description DSN 001 Main Station Discharge DSN 002 Emergency Discharge from Unit 2 (if DSN 001 is blocked)
'a DSN 003 Emergency Discharge from Unit 1 (if DSN 001 is blocked) l DSN 004 Emergency Discharge from Unit 1 (if Unit 1 NDCT blocked)
DSN 005 Stormwater and yard drainage and I
dewatering of natural draft cooling towers, maintenance dredging l
desiltation and basin dewatering, fire brigade training facility runoff, fire service water runoff.
5-10 Amendment No. [ k 149
4 TABLE 6.2-1 MINIMUM SHIFT CREW COMPOSITION ('
')
' LICENSE CATEGORY QUALIFICATIONS Tave > 200' Tave < 200*
SR0('")
2 1(')
1.
1 R0('")
2 1
l
.Non-Licensed' Auxiliary Operator 2
1 Shift Technical Advisor 1(")
None Requi ed l
t (i)
Does not include the Licensed Senior Reactor Operator or l
' Senior. Reactor Oparator Limited to Fuol Handling, supervising (a) irradiated fuel he.ndling and transfer activities onsite, and f.b)-all unirradiated fuel handling and transfer activities j
)
to and from the Reactor Vessel.
I (ii)
May be on'a different shift rotation than licensed personnel.
(iii) ^Except for the Shift Supervisor, shift crew composition may b6 l
ene less than the minimum requirements for a period of time not to excted 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accomodate unexpected i
absence of on-duty shift crew memb'ets provided immediate action is taken to restore the shift crew composition to I
within the minimum requirements of Table 6.2-1.
This provision does not permit any shift crew position to be l
unmanned upon shift change due te an incoming shift crewman being late or absent.
{
(iv)
Pursuant to the requirements of 10 CFR 50.54(m).
[
j 1
6-2 Amendment No. [ [ [ 149
c s
L 6.5.1.9
.The Emergency Plan and implementing procedures shall. be l
reviewed by a knowledgeable individual (s)/ group other than
'the individual (s)/ group which prepared them.
6.5.1.10. A knowledgeable individual (s)/ group shall review every unplanned onsite release of radioactive material to the environs including the preparation and forwarding of I
reports to the Vice President Tf11-1 covering evaluations, recommendations and disposition of the corrective action to prevent recurrence.
6.5.1.11 Major changes to radwaste systems shall be reviewed by a knowledgeable individual (s)/ group other than the individuals (s)/ group which prepared them.
6.5.1.12 Individuals responsible for reviews performed in l
accordance with 6.5.1.1 through 6.5.1.4 shall include a determination of whether or not additional cross-disciplinary review is necessary.
If deemed necessary, such review shall be performe!1 by the appropriate personnel. Individuals responsible for reviews considered under 6.5.1.1 through 6.5.1.5 shall render determinations in writing with regard to whether or not 6.5.1.1 through 6.5.1.5 constitute an unreviewed safety question.
RECORDS' t
6.5.1.13 Written records of activities performed under Specifi-cations 6.5.1.1 through 6.5.1.11 shall be maintained.
QUALIFICATIONS
'a 6.5.1.14 Responsible Technical Reviewers shall meet or exceed the qualifications of ANSI /ANS 3.1 of 1978 Section 4.6, or 4.4 for applicable disciplines, or have 7 years of appropriate experience in the field of his specialty.
Credit toward experience will be given for advanced degrees on a one-to-one basis up to a maximum of two years. Responsible Technical Reviewers shall be designated in writing.
6.5.2 INDEPENDENT SAFETY REVIEW FUNCTION 6.5.2.1 The Vice President of each division within GPU Nuclear Corporation shall be responsible for ensuring the l
independent safety review of the subjects described in 6.5.2.5 within his assigned area of safety review responsibility, as assigned in the GPUN Review and Approval Matrix.
6.5.2.2 Independent safety review shall be completed by an individual / group not having direct responsibility for the performance of the activities under review, but who may be from the same functionally cognizant organization as the individual / group performing the original work.
6.5.2.3 GPU Nuclear Corporation shall collectively have or have access to the experience and competence required to independently review subjects in the following areas:
Amendment No. [g[)d,149 6-5
a m* -
z&
i E
'dures et'least'once'per 24 months.
f..
The' Process Control Program and implementing procedures' for solidification of radioactive wastes at least once per 24 months.
- j. The' performance of activities required by the Quality Assurance Program to meet criteria of Regulatory Guide 4.15, December,1977 at least once per 12 months.
u k.
Any other area of unit' operation considered appropriate I
by the 10SRG or-the Office of the President - GPUNC.
6.5.3.2-Audits of the following shall be performed under the -
cognizance of the vice president responsible for technical support:
An independent fire protection and loss prevention a.
program inspection and audit shall be performed annually utilizing either qualified licensee l
personnel or an outside fire protection firm.
h.
An inspection and audit of the fire protection and loss prevention program, by an outside qualified fire consultant at. intervals no greater than 3 years.
RECORDS 6.5.3.3 Audit reports encompassed by sections 6.5.3.1 and 6.5.3.2 shall be forwarded for action to the management positions responsible for the areas audited within 60 days after completion of the audit.
Upper management shall be informed per the Operation Quality Assurance Plan.
6.5.4 INDEPENDENT s0NSITE SAFETY REVIEW GROUP (IOSRG) STRUCTURE I
g l
6.5.4.1 The IOSRG shall be a full-time group of engineers, experienced in nuclear power plarJc engineering, operations and/or technology, independent of the unit staff, and located on site.
j ORGANIZATION i
6.5.4.2 a.
The IOSRG shall consist of a manager and a minimum staff of 3 members who meet the qualifications of 6.5.4.5.
Group expertise shall be multi-disciplined.
(
b.
In the event of an unanticipated vacancy in the IOSRG l
staff, the number of staff can be two (2) members for a period of not to exceed six (6) months while the vacancy is being filled.
I c.
The IOSRG shall report to the director responsible for i
nuclear safety assessrr. ant.
I 6-8 Amendment No.((k[ 149
r.
.r I
6.9.4.2.5 L
The Radioactive Efflu'ent Release Reports shall include the instrumentation not returned to OPERABLE status within 30 days per TS 3.21.1.b and TS 3.21.2.b.
l i
6.9.'4.3 The following information shall be included in the Radioactive-Effluent Release Report to be submitted 60 days after January 1 of each year.
6.9.4.3.1 The Radioactive Effluent Release Report to be submitted 60 days after January 1 of each year shall include an annual summary of hourly meteorological data collected over the l
previous year.
This annual summary may be either in the form of an hour-by-hour listing of wind speed, wind i
direction, atmosphere stability, and precipitation (if measured) on magnetic tape, or in the form of joint frequency distribution of wind speed,-wind direction, and atmospheric stability.
6.9.4.3.2 The Radioactive Effluent Release Report to be submitted 60 days after January 1 of each year shall include an assessment of the radiation doses due to the radioactive liquid and gaseous effluents released from the unit or station during the previous calendar year.
6.9.4.3.3 The Radioactive Effluent Release Report to be submitted 60 days after January 1 of each year shall include an assessmant of the radiation doses from radioactive liquid and gaseous effluents to MEMBERS OF THE PUBLIC due to their activities inside the site boundary (Figure 5-3) l during the report period. All assumptions used in making these assessments (i.e., specific activity, exposure time and location) shall be included in these reports. The meteorological conditions conc'utrent with the time of release of radioactive materials in gaseous effluents (as determined by sampling frequency and measurement) shall be used for determining the gaseous pathway doses.
The 4
assessment of radiation doses shall be performed in accordance with the OFFSITE DOSE CALCULATION MANUAL (00CM).
l 6.9.4.3.4 The Radioactive Effluent Release Report to be submitted 60
{
days after knuary'1 of each year shall also include art assessment of radiation doses to the likely most exposed I
real individual from reactor releases and other nearby j
uranium fuel cycle sources including doses from primary I
effluent pathways and direct radiation for the previous 12 consecutive months to show conformanck with 40 CFR 190 i
" Environmental Radiatien Protection Standards for Nuclear Power Operation". Acceptable methods for calculating the dose contributions from Liquid and gaseous effluents are given in Regulatory Guide 1.109, Rev. 1.
6-19 Amendment No. [ [ J 6, 149
r : ":'4 L
i 6.10
' RECORD RETENTION 6.10.1 The following records shall be retained for at least five years:
f
' Records of normal station operation including power-a.
levels and periods.of operation at each power level..
b.
Records of principal maintenance activities, including inspection,. repairs, substitution, or replacement of principal items of equipment related to nuclear safety.
j i
c.
All REPORTABLE EVENTS 1
-d.
Records of periodic checks, tests and calibrations.
]
Records of reactor physics tests and other special e.
tests related to nuclear safety.
f.
Changes to procedures required by Specification 6.8.1.
g.
Records of solid radioactive shipments.
l a.
1 i
i i
6-19a l
Amendment No. g g J21f' M % 149
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