ML20054D636
| ML20054D636 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 03/20/1982 |
| From: | Vassallo D Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20054D631 | List: |
| References | |
| TAC-48771, NUDOCS 8204230219 | |
| Download: ML20054D636 (8) | |
Text
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UNITED STATES E
h NUCLEAR REGULATORY COMM!SSION O
ep WASHINGTON, D. C. 20555 klCM*****f" BOSTON EDISON COMPANY DOCKET NO. 50-293 PILGRIM NUCLEAR POWER STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 56 License No. OPR-35 1.
The Nuclear Regulatory Comission (the Commission) has found that:
A.
The application for amendment by the Boston Edison Company (the licensee) dated February 19, 1982 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),
and the Commission'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 A::t, 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 comon 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 Spec-ifications as. indicated in the attachment to this license amendment and paragraph 3.B of Facility Operating License No. OPR-35 is hereby amended to read as follows:
B.
Technical Soecifications The Technical Specifications contained in ppendices A and B, as revised through Amendment No. 56, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
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This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
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Domenic B. Vassallo, Chief Operating Reactors Branch #2 Division of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: March 20,1982 h
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ATTACHMENT TO LICENSE AMENDMENT NO. 56 FACILITY OPERATING LICENSE NO. DPR-35 DOCKET NO. 50-293 Replace the following pages of the Appendix "A" Technical Specification 7 with the enclosed pages. The revised page is identified by Amendment Number and contains a vertical line indicating the area of change.
Remove Replace 58a 59 59 126 126 127 127 145 145 4
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g Surveillance Instrumentation Table 3.2.F. (Continued)
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P Minimum # of Operable Instrument Instrument Type Indication
,m Channels Instrunent and Range Notes 1/ Valve See Note (6)
Tall Pipe Temperature Thermocouple (6)
Indication e
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1CTIS FOR TABLE 3.2.T_
s From and after the date that one of these, parameters is reduced to one indication, continued operation is permissible during the l
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(1) succeeding thirty days unless such instrumentation is sooner made operable.
Trom and after the date that one of these parameters is not indi-cated in the control room, continued operation is permissible
~ (2) h instrumentation is
,during the succeeding seven days unless suc i
sooner made operable.
If the requirements of notes (1) and (2) cannot be met, an orderly shutdown shall be initiated and the reactor shall (3) be in a Cold Shutdown Condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
These surveillance instruments are considered to be redundant to (4) each other.
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At a minimum, the primary or back-up* parameters shall be operable for With both pri=ary l
(3) sach valve when the valves are required to be operable.bac 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
to operable status within 31 days or be in a shutdown mode within and The following instruments are associated with the safety / relief & safett valves:
Secondary Primary w
Tail Pipe Te=perature Thernoccuple g,g Acoustic Monitor TE6271-3 203-3A ZT-203-3A TE6272-3 203-35 IT-203-35 i\\
TE6273-3 203-3C ZT-203-3C TI6276-3 203-3D ZT-203-3D TI6274-3 203-4A IT-203-4A TI6275-B 203-43 ZT-203-4B
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- See Note (6) tail pipe temperature, one of the At a minimtsn, the above listed SR (6) dual thennoccuples, will be operable for each valve when the valves If a thermocouple becomes inoperable, it are required to be operable.
shall be returned to an operable. condition within 31 day's or the reactor shall be placed in a shutdown made within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
e Amendment No. M 56 59 m
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t,IMIT,Jyc CONDITION FOR 0?! RATION SURVETIUNCE REQUIR.W_ INT
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b' 3.6.C Coolant chemistry (Cont'd) 4.6 i
pcver operation is per=1ssible l
only during the succaeding seven days.
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3.
If the con'ditions in'1.or 2 above cannot be mat, an orderly shutdown shall be initiated and the rasctor 4
. shall be in a Cold Shutdeva Condi-tion within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
D.,
Safety and Relief Valves [
D.
Safety and Relief Valves l
1.
During reactor power operating 1.
At least one safety valve and two conditicas and prior to reactor relief /safet7 valves shall be i
startup from a Cold Condition, or checked or replaced with bench Whenever reactor coolant pressure checked valves enca per operating i
is grestar than 104 psig *and tam-cycle. All valves vill be testad perature greater than 340 T, both every two cycles, safety valves and the safacy modes of all relief valves shall be op-The set point of the safety valves arabla.
shall be as specified in Specifi-cation 2.2.
2.
If Specification 3.6.D.1 is not met, an orderly shutdown shall be in-2.
At least one of the relief / safety iriated and the reactor coolant valves shall be disassenblad and pressure shall be belosr 104 psig inspected each refuelist outage.
within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Note: Technical Specifications 3.6.D.2 - 3.6.D.5 3.
apply only when two Stage Target-Whenever the safety relief valves Rock SRVs are installed.
are required to be operable, the discharge pipe temperature of each safety relief valve shall be 3.
If the temperature of any safety logged daily.
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relief discharge pipe exceeds 212 F 0
during normal reactor power operation 4.
for a, period of greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, Instrumentation shall be calibrated an engineering evaluation shall be and checked as indicated in Table _.
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s 4.2.F.
performed justifying continued opera-tion for the corresponding temp.
increases, and a Report shall be 5.
Notwithstanding the above, as a issued per T.S. Section 6.9.B.1 which shall address the actions that have minimum safety relief valves that have been in service shall be i
been taken or a schedule of actions to be taken, tested in the as-found condition during both Cycle 6 and Cycle 7.
4.
Any safety relief valve whose dis-charge pipe temperature exceeds 2120F for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or more shall be removed at the next cold shutdown of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or more tested in the as-found condi-
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i tion, and recalibrated as necessa.ry prior to reinstallation.
Power opera-
[i tion shall not continue beyond 90 days 126 Anendment No.fl 56
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1 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENT 3.6.D Safety and Relief Valves (Cont'd) from the initial discovery of dis-charge pipe temperatures in excess of 212oF for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> without prior NRC approval of _the engineering E.
Jet Pumpe evaluation delineated in 3.6.D.3.
Whenever there is recirculation flow 5.
The limiting conditions of operation with the reactor in the startup or for the instrumentation that monitors run modes, jet pump operability shall tail pipe temperature are given in be ' checked daily by verifying that Table 3.2.F.
the following conditions do not oc-cur simultaneously:
1.
The two recirculation loops have a flow i= balance of 15.
E.
Jet Ptmpa or more when the pumps are 1.
Whenever the reactor is in the startup or run modes, all jet 2.
The indicated value of core pumps shall be operable.
If it is flow rate varies from the determined that a jet pump is value derived from loop flow inoperable, an orderly shutdown i
measurements by more than 10%.
shall be initiated and the reactor shall be in a Cold Shutdown Condi-3.
The diffuser to lower plenum tion within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
differential pressure reading on an individual jet pump varies from established' jet pump P characteristics by more than 10%.
i F.
Jet Pump Flow Mismatch F.
Jet Pump Flow Mismatch 1.
Whenever both recirculation pumps Recirculation pump speeds shall be are in operation, pump speeds shall, checked and logged at least once be maintained within 10% of each per day.
other when power level is greater than 80% and within 15% of each other when power level is less chan or equal to 80%.
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, Structural Integrity G.
Structural Integrity 1.
The structural integrity of the The nondestructive inspections listed primary system boundary shall in Table 4.6.1 shall be performed as be maintained at the level re-specified. The results obtained from quited by the ASME Boiler and compliance with this specification Pressure Vessel Code, section will be evaluated after 5 years and X%, " Rules for Inservice In-the conclusions of this evaluation spection cf Nuclear Power will be reviewed with AEC.
l Plant Components," 1974 i e 7,4-J i
, 6. h Amendment No. /d '56 g
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3.6.D and 15.6.D
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Safety and Relief valves As discussed in Subsection 4.4.6 of the Final Gafety Analysis,Repert, -
design of the m al - system pressure relief system is intanded te protect the nuclear system from,overpressurination in the event of the safety valve sining transient..' An 1:yii. net scram is assumed because JLM Boiler and Pressure vessel Code, Sectica III,. requires that protectics syste=s directly related to the valve s4-d g transient must not be credited with action in deter-4 d g valve relieving capacity. Atotalofkrelief/safetyvalvesa=d2safetyvalvesis provided by the design.
Experience in safety valve operation shows that a testing of at least Sc5 of the safety valves per refueling outage is adequate to detect fai1"res or deterioration. The tolerance value of +1% is in accer-
. :...- ith Section III of the ASME Eciler and Pressure vessel Code.
An ama.Q.is m., been performed which shows that with all safety valves set 1% higher, the reacter coolant pressure safety 11=10 of 1375 psig is oct exceeded.
The relief / safety valves have two fu=c'tions; i.e., power elief or self-actuated by high pressure. Power relief is a solenoid actuated function (Automatic Pressure Relief)' in which exter=al instramen-tation sign =1= of coincident high drywell pressure and icv-los water level initiate the valves to open. This function is discussed in Specification 3 5.D..
In addition,. the valves can be operated
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Pilgrim's experience with 2 stage safety / relief valves has demonstrated that mininnwn leakage exists when the ta11 pipe temperature is 215 Farenheit.
0 Therefore, a reporting requirement triggered by a temperature of 212.F is conservative, and assures timely reporting before leakage reaches signifi-cant proportions'.
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