ML20195B893
| ML20195B893 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 11/10/1998 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20195B896 | List: |
| References | |
| NUDOCS 9811160306 | |
| Download: ML20195B893 (9) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION o
i WASHINGTON, D.C. 20555-00()1 f
l BOSTON EDISON COMPANY DOCKE T NO. 50-293 PILGRIM NUCLEAR POWER STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 178 License No. DPR-35
- 1. The Nuclear Regulatory Commission (the Commission or the NRC) has found that:
A.
The application for amendment filed by the Boston Edison Company (the licensee) dated June 26,1998, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; 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 Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.8 of Facility Operating License No. DPR-35 is hereby amended to read as follows:
B.
Technical Soecifications The Technical Specifications contained in Appendix A, as revised through Amendment No.178
, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
t I
9911160306 901110 PDR ADOCK 05000293 p
PM l
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3.
This license amendment is effective as of its date of issuance and shall be implemented within 30 days.
FOR THE NUCLEAR REGULATORY COMMISSION 0*
W Cecil O. Thomas, Director Project Directorate 1-3 Division of Reactor Projects - t/11 Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications
.j Date of issuance: November 10, 1998 I
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1 ATTACHMENT TO LICENSE AMENDMENT NO.178 t
FAClLITY OPERATING LICENSE NO. DPR-35 l
DOCKET NO. 50-293 l
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Replace 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.
I Remove Insert 3/4.1-2 3/4.1-2 3/4.2-20 3/4.2-20 3/4.3-3 3/4.3-3 B 3/4.7-2 B 3/4.7-2 B 3/4.7-4 8 3/4.7-4 i.
B 3/4.7-5 B 3/4.7-5 i
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PNPS Table 3.1.1 REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION REQUIREMENT Operable Inst.
Modes in Which Function Must Be Channels per Operable Trip System ")
- Trip Function Trip Level Setting Refuel Startup/ Hot Run Action")
Minimum Avail.
Standby R
1 1
Mode Switch in Shutdown X
X A
M 1
1 Manual Scram X
X X
A IRM
- 20/125 of full scale XM X
W 3
4 High Flux A
~
3 4
Inoperative XM X
W
,A APRM
" 7)
" 7) 2 3
High Flux X
A or B
")
M N
2 3
Inoperative X
X.
X A or B A
2 3
High Flux (15%)
s 15% of Design Power X
X A or B
~
2 2
High Reactor Pressure s 1063.5 psig X"*
X X
- A W
X*
X A
2 2
High Drywell Pressure s 2.22 psig X
2 2
Reactor Low Water Level 2 11.6 in. Indicated Level X"*
X X
s 38 Gallons X
X X
A 2
2 East 2
2 West 4
4 Main Steam Line X 'X*
X DX*
I W
Isolation Valve Closure 5 0% Valve Closure X
A or C 1
2 2
Turbine Control Valve Fast 2150 psig Control Oil X")
X")
X")
A or D l
Closure Pressure at Acceleration Relay 4
4 Turbine Stop Valve s 10% Valve Closure X")
X")
X")
A or D Closure Revision Amendment No. 45,42,86,92,*17,133,147,151,152,151,164,169 178 3/4.1-2
PNPS TABLE 3.2.C.1 (Cont)
INSTRUMENTATION THAT INITIATES ROD BLOCKS Operable Channels per Tri) Function Trio Function Minimum Available Reauired Operational Conditions Notes SRM Detector not in Startup 3
4 Startup/ Refuel, except trip is bypassed when SRM (1)(4)(6)
Position count rate is 2 100 counts /second or IRMs on Range 3 or above SRM Downscale 3
4 Startup/ Refuel, except trip is bypassed when IRMs (1)(4)(6) on Range 3 or above SRM Upscale 3
4 Startup/ Refuel, except trip is bypassed when th.e (1)(4)(6) 1RM range switches are on Range 8 or above (4)
SRM Inoperative 3
4 Startup/ Refuel, except trip is bypassed when the (1)(4)(6) lRM range switches are on Range 8 or above (4)
Scram Discharge instrument 2
2 Run/Startup/ Refuel (3)(6)
Volume Water Level-High Scram Discharge instrument 1
1 Refuel / Shutdown (3)(6)
Volume-Scram Trip Bypassed Revision Amendment No.13, ? 47,159,169178 3/4.2-20
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LIMITING CONDITION FOR OPERATION -
SURVEILLANCE REQUIREMENT
{
j 3.3 REACTIVITY CONTROL (CONT) 4.3 REACTIVITY CONTROL (Cont)
.B.. Control Rods (Cont)
B.
. Control Rods (Cont)
- 2. The control rod drive housing
- b. When the rod is fully
)
support system shall be in place withdrawn the first time during reactor power operation and subsequent to each when the reactor coolant system is refueling outage or after i
pressurized above atmospheric:
maintenance, observe that
,i pressure with fuel in the reactor.
_ the drive does not go to the vessel, unless all control rods are overtravel position.
i-fully inserted and Specification i
3.3.A.1 is met.
i
- 2. The control rod drive housing i
l support system shall be inspected after reassembly and -
the results of the inspection J
recorded.
3.
- a. No control rods shall be moved
- 3. Prior to control rod withdrawal for when the reactor is below 20%
startup or insertion to reduce ~
rated power, except to shutdown power below 20% of the L
the reactor, unless the Rod operability of the Rod Worth Worth Minimizer (RWM) is Minimizer (RWM) shall be operable. A maximum of two verified by:
rods may be moved below 20%
design power when the RWM is
- a. verifying the correctness of the control rod withdrawal inoperable if all other rods sequence input to the RWM except those which cannot be moved with control rod drive
- computer, pressure are fully inserted.
- b. Control rod patterns and the
- b. performing the RWM sequence of withdrawal or computer diagnostic test.
insertion shall be established l
such that:
l
- 1) when the reactor is critical
- c. verifying the annunciation of g
and below 20% design power the selection errors of at the maximum worth of any least one out-of sequence insequence control rod which control rod in each distinct is not electrically disarmed is RWM group.
less than 0.010 delta k.
- d. verifying the rod block
- 2) and when the reactor is above function of an out of-20% design power the sequence control rod which is L
maximum worth of any control withdrawn no more than three rod, including allowance for a notches.
single operator error, is less than 0.020 delta k.
J P
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. Revision
. Amendment No. 39178 3/4.3 3 l
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BASES:
3/4.7 CONTAINMENT SYSTEMS (Cont)
A.
Primary Containment (Cont)
The maximum permissible bulk suppression pool temperature of 120 F is acceptable since a complete accident blowdown can be accomodated without exceeding the bulk suppression pool temperature limit of 170 F immediately after blowdown. This 170 F LOCA blowdown limit is not a limit for the heatup of the suppression pool after the vesselis depressurized. Specifications have been placed on the envelope of reactor operating conditions so that the reactor can be depressurized in a timely manner to avoid the regime of potentially high pressure suppression chamber loadings. Current Technical Specification limits on suppression pool temperature ensure bulk pool temperature remains within an acceptable range to condense steam discharged to the suppression pool during a LOCA or SRV actuation.
in addition to the limits on temperature of the suppression chamber pool water, operating procedures define the action to be taken in the event a relief valve inadvertently opens or sticks open. This action would include: (1) use of all available means to close the valve, (2) initiate suppression pool water cooling heat exchangers, (3) initiate reactor shutdown, and (4) if other relief valves are used to depressurize the reactor, their discharge shall be separated from that of the stuck-open relief valve to assure mixing and uniformity of energy insertion to the pool.
Because of the large volume and thermal capacity of the suppression pool, the volume and temperature normally changes very slowly and monitoring these parameters daily is sufficient to establish any temperature trends. By requiring the suppression pool temperature to be continually monitored and frequently logged during periods of significant heat addition, the temperature trends will be closely followed so that appropriate action can be taken. The requirement for an external visual examination following any event where potentially high loadings could occur provides assurance that no significant damage was encountered. Particular attention should be focused on structural discontinuities in the vicinity of the relief valve discharge since these are expected to be the points of highest stress.
If a loss-of-coolant accident were to occur when the reactor water temperature is below approximately 330 F, the containment pressure will not exceed the 62 psig code permissible pressure, even if no condensation were to occur. The maximum allowable pool temperature, whenever the reactor is above 212 F, shall be governed by this specification. Thus, specifying water volume-temperature requirements applicable for reactor water temperature above 212 F provides additional margin above that available at 330 F.
Revision Amendment No. 39,53,83, '13 178 B3/4.7-2
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BASES:
3/4.7 CONTAINMENT SYSTEMS (Cont)
A.
Primary Containment (Cont) l capability of the structure over its service lifetime. Additional margin to maintain the containment in the "as built" condition is achieved by establishing the allowable operationalleak rate. The allowable operationalleak rate is derived by multiplying the maximum allowable leak rate or the allowable test leak rate by 0.75 thereby providing a 25% margin to allow for leakage deterioration which may occur during the period between leak rate tests.
t The primary containment leakage rate testing is based on the guidelines in Regulatory Guide 1.163 dated September 1995, NEl 94 01 Revision 0 dated July 25, 1995, and ANSI /ANS 56.8-1994. Specific acceptance criteria for as-found and as-left leakage rates, as well as methods of defining the leakage rates, are contained in i
the primary containment leakage rate testing program.
i The primary containment leak rate test frequency is based on maintaining adequate assurance that the leak rate remains within the specification. The leak rate test frequency is in accordance with 10CFR50 App. J, Option B and Regulatory Guide 1.163 dated September 1995.
Type A, Type B, and Type C tests will be performed using the technical methods and techniques specified in ANSI /ANS 56.8 - 1994 or other alternative testing methods approved by the NRC.
l A note is included in Surveillance 4.7.A.2.a stating that definition 1.U is not i
applicable. The 25% allowable extension of surveillance intervals is already included in the primary containment leakage rate testing program; therefore, an additional 25% is not allowed.
The penetration and air purge piping leakage test frequency, along with the containment leak rate tests, is adequate to allow detection of leakage trends.
Whenever a bolted double gasketed penetration is broken and remade, the space between the gaskets is pressurized to determine that the seals are performing properly, it is expected that the majority of the leakage from valves, penetrations and seals would be into the reactor building. However, it is possible that leakage into other parts of the facility could occur. Such leakage paths that may affect significantly the consequences of accidents are to be minimized. The personnel air lock is tested at 10 psig, because the inboard door is not designed to shut in the opposite direction.
Primary Containment isolation Valves Double isolation valves are provided on lines penetrating the primary containment and open to the free space of the containment. Closure of one of the valves in each line would be sufficient to maintain the integrity of the pressure suppression system.
Automatic initiation is required to minimize the potential leakage paths from the containment in the event of a loss of coolant accident.
Revision Amendment Nc.
- 13,136,197, # 72 178 B3/4.7-4
4 BASES:
9 3/4.7 CONTAINMENT SYSTEMS (Cont)
A.
Primary Containment (Cont)
Group 1 - process lines are isolated by reactor vessel low-low water level in order to allow for removal of decay heat subsequent to a scram, yet isolate in time for proper operation of the core standby cooling systems. The valves in group 1 are also closed when process instrumentation detects excessive main steam line flow, low pressure, main steam space high temperature, or reactor vessel high water level.
Group 2 - isolation valves are closed by reactor vessel low water level or high drywell pressure. The group 2 isolation signal also " isolates" the reactor building and starts De standby gas treatment system. It is not desirable to actuate the group 2 isolation signal by a transient or spurious signal.
Group 3 - isolation valves can only be opened when the reactor is at low pressure and the core standby cooling systems are not required. Also, e nce the reactor vessel could potentially be drained through these process lines, these valves are closed by low water level.
Group 4 and 5 - process lines are designed to remain operable and mitigate the consequences of an accident which results in the isolation of other process lines.
The signals which initiate isolation of group 4 and 5 process lines are therefore indicative of a condition which would render them inoperable.
Group 6 - process lines are normally in use and it is therefore not desirable to cause spu,lous isolation due to high drywell pressure resulting from non-safety related causes. To protect the reactor from a possible pipe break in the system, isolation is provided by high temperature in the cleanup system area or high flow through the inlet to the cleanup system. Also, since the vessel could potentially be drained through the cleanup system, a low level isolation is provided.
Group 7 - The HPCI vacuum breaker line is designed to remain operable when the HPCI system is required. The signals which initiate isolation of the HPCI vacuum breaker line are indicative of a break inside containment and reactor pressure below that at which HPCI can operate.
The maximum closure time for the automatic isolation valves of the primary containment and reactor vessel isolation control system have been selected in consideration of the design intent to prevent core uncovering following pipe breaks outside the primary containment and the need to contain released fission products following pipe breaks inside the primary containment.
In satisfying this design intent an additlenal margin has been included in specifying maximum closure times. This margin permits identification of degraded valve performance, prior to exceeding the design closure times.
In order to assure that the doses that may result from a steam line break do not exceed the 10CFR100 guidelines, it is necessary that no fuel rod perforation resulting from the accident occur prior to closure of the main steam line isolation valves.
Analyses indicate that fuel rod cladding perforations would be avoided for main steam valve closure times, including instrument delay, as long as 10.5 seconds.
Revision Amendment Nc.
- 13,167 178 B3/4.7-5