ML20116M557
| ML20116M557 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 11/16/1992 |
| From: | Butler W Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20116M559 | List: |
| References | |
| NUDOCS 9211200377 | |
| Download: ML20116M557 (11) | |
Text
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= UNITED STATES T
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- NUCLEAR REGULATORY COMMISSION 2*
- WASHINoToN, D.C. 20805 -
/
BOSTON EDISON COMPA'NY DOCKET NO. 50-293-PILGRIM NUCLEAR POWER STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.-143 License No. DPR-35 1.
The Nuclear Regulatory Comission (the Commission or the NRC) has folad i
that:
t A.
The application for amendment filed by the Boston Edison Company (the' licensee) dated October 7, 1991 and supplemented October 26,' 1992, 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; 8.
The facility will operate in conformity with the application,1the provisions of the Act, and the rules and. regulations of the Commission; C.
There is reasonable assurance:
(i) that tne activities authorized by this amendment can be conducted without endangering the health and safety of_ the public, and (ii) that such activitiesiwill: be conducted in compliance with the Commission's regulations set forth in 10-CFRL Chapter I; 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 reqtiirementsihave been satisfied.
2.
Accordingly,Lthe license is amended by changes to the Technical Specifica7
-tions as indicated in the attachment-to this license amendment, and.
. aragraph.3.B of Facility-operating License No. 0PR-35 is~hereby= amended p
to read as follows:
9211200377 921116 PDR ADOCK 05000293 F
PDR a
a
. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.143, are hereby incorporcted in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of its'date of issuance and shall be implemented within 30 days.
FOR THE NUCLEAR REGULATORY COMMISSION
/dte &A WalterR. Butler,> Director Project Directorate I-3 Divisica of Reactor Projects - I/II Office of Nuclear Reactor Regulatico
Attachment:
Changes to the Technical Specifications Date of Issuance: November 16, 1992
-u-.
ATTACHMENT 10 LICENSE AMENDMENT NO.143 FACillTY OPERATING LICENSE NO. DPR QQ[EET NO. 50-291 Replace the following paae of the License with the 'sttached page.
The revised page contains vertical lines indicating the area of change.
Remove Jg
-3a-
-3a-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.
Remove Insert 11 11 198 198 206 206 206 a 206 A 206 b 206 B 206 c 206 c-1 206 d 206 e 206 e-1 Text moved 206 e-2 to page 206 206 f.
206'f-l 206 f-2 206 g 206 h 206 i 206 i-1 206 j 206 j-1 Text moved to page 206B 209 209 212 212
-3a-3.F Fire Protection Boston Edison shall implement and maintain in effect all provisions of the approved fire protection program as described in tht Final Safety Analysis Report for the facility and as approved in the SER dated December 21, 1978 as supplemented subjact to the following provision:
Boston Edison may make changes to the approved fire protection program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.
3.G PAL,..lcal Protec* ion The licensee shall fully implement and maintain in effect all prc"isions of the Commission-approved physical security, guard training and qualification, and safeguards contingency plans including amcodments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 278!7 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The plans, which contain Safeguards Information protec.ed under 10 CFR 73.21, are entitled:
" Pilgrim Nuclear Power Station Physical Security Plan," with revisions submitted through September 18, 1987: " Pilgrim Nuclear Power Station Guard Training and Qualification Plan,".with revisions submitted through September 24, 1984; and " Pilgrim Nuclear Power Station Safeguards Contingency Plan," with revisions submitted through February 15, 1984. Changes made in accordance with 10 CFR 73.55 shall be implemented in accordance with the schedule set forth therein.
Amendment No. 143
Surveillance Pace No.
3.7 00NTAINHENT SYSTEMS 4.7 152 A.
Primary Containtaent n
152 B.
Standby Cas Treatment-System and B
158 Control Room High Efficiency Air Filtration System C.
159 3.8 RADI0 ACTIVE EFFLUENTS 4.8 177 A.
Liquid Effluents Concentration A
177 B.
Radioactivo Liquid Effibent B
177 Instrumentation C.
Liquid Radwaste Treatment C
178 D.
Gaseous Effluents Dose Rate D
179 E.
Radioactive Gaseous Effluent E
180 I.istrumentation F.
Gaseous Effluent Treatment F
181 G.
182 H.
Mechanical Vacuum Pump H
183 3.9 AUXILIARY ELECTRICAL SYSTEM 4.9 194 A.
Auxiliary Electrical Equipment A
194 8.
Operation with Inoperable Equipment 196 3.10 CORE ALTERATIONS 4.10 202 A.
Refueling Interlocks A
202 B.
Core Monitoring B
2U2 C.
Spent Fuel Pool Hater Level C
203 D.
Multiple Control Rod Removal D
203 3.11 REACTOR FUEL ASSEMBLY 4.11 205a A.
Average Planar Linear Heat A
205a Generation Rate (APLHCR)
~B.
Linear Heat Generation Rate (LHGR)
B 205b C.
Hinimum Critical Power Ratio (HCPR)
C 205c 1
D.
Power / Flow Relationship D
205d 3.12 FIRE PROTECTION - ALTERNATE SHUTDOHN 4.12 206 PANELS Amendment No.
75, 27, 45, 84, 89, 773, Ild, 733, 143 11
BMil:
l t
3.9 The general objectivo of this Specification is to assure an adequate source of electrical power to operate the auxiliaries during )lant operation, to operate facilities to cool and lubricate the plant during slutdown, and to operate the engineered safeguards following an accident.
There are three sources of a c elect.. cal energy available; namely, the startup transformer, the diesel I
geneators and the shutdown transformer.
Ihn d c supply is required for switchgear and engineered safety feature systems.
Specification 3.9. A states the r1guired availability of a c and d-c power,
i.e., an active off-site a-c source, a back up source of off site a-c power and the maximum amount of on-site a c and d c sources.
The diesel fuel supp1', consists of two (2) 25,0nD gallon tanks.
Level instrumentation providas operators the information necessary to ensure a niinimum supply of 19,800 gallons in each tank.
Auxiliary power fcr PNPS is supplied from two sources; either the unit auxiliary transformer or the startup transformer.
Both of these transformers are sized to carry 100% nf the auAlliary load.
if the startup transformer is lost, the unit can continue to operate since the unit auxiliary transformer is in service, the shutdown transformer is available, and both diesel generators tre operational.
If the startup and shutdown traasformers are both lost, the reactor power level must be reduced to a value whereby the unit could safely reject the load and continue to supply auxiliary electric power to the station.
In the normal mode of o)eration, the tartup transformer is energized, two diesel generators and tie snutdown transformer are operable. One diesel generator may be allowed out of service based on the availability of
)ower frun the startup transformer, the shutdown transformer and the fact t1at one diesel generator carries sufficient engineered safeguards equipment to cover all breaks. With the shutdown transformer and one diesel generator out of service, both 345kV supply lines must be available for the startup transformer.
Upon tha loss of one on-sito and one off-ite power source, power would be available from the other immediate off-si e power source and the one operable on-site diesel to carry sufficient engineered safeguards equipment to cover all breaks.
In addition to these two power sources, ren,aval of the Isolated Phase Bus flexible connecto:s would allow backfeed of power thrnugh the main transforner to the unit auxiliary transformer and provide power to carry the full station auxiliary load.
L time required to perform this operation is comparable to the time the reactor could remain on RCIC operation before controlled depressurization need be initiated.
A battery charger is supplied with each of the 125 and 250 volt batteries and, in addition, (1) a 125 volt shared back up br.ttery charger is supplied which Amendment %, 143 198
LIMITING CONDITION FOR OPERATION SURVEILLANCE REOUIREMENTE 3.12 Fire Protection 4.12 Fire Protection Altarnate Shutdown Panels Alternate Shutdown Panels 1.
Alternate shutdown aanels for the The alternate shutdown panels following systems s1all be shall be demonstrated to be OPERABLE:
OPERABLE according to the following:
1.
Core Spray 2.
RHR 1.
The motor operated valves 3.
RBCCH of the core spray system 4.
Salt Service Hater shall be opereted from the 5.
HPCI alternate shutdown panels 6.
RCIC once each cycle.
7.
Automatic Depressurization 8.
Diesel Generators 2.
The motor operated valves of the RHR system shall be APPLICABILITY:
operated once each cycle utilizing the MCC B-17 At all times that the system is alternate power source, required to be OPERABLE.
3.
The pumps of the SSH ACTION:
system shall be operated from the alternate Hith any of the alternate shutdown shutdown panels once each panels inoperable, cycle.
a) Immediately verify that fire 4.
The pumps and motor detection with automatic fire operated valves of the suppressicn for the Cable Spreading RBCCH system shall be Room is Operable.
If fire operated from the detection with automatic fire alte nate shutdown panels supprcssion cannot be determined once each cycle, operable, within one (1) hour from the time the system is determined 5.
Alternate shutdown panel to be inoperable, estdlish a capability for the RCIC continuous fire Hatch with backup and HPCI systems shall be fire suppression.
verified to be OPERABLE once each cycle, b) Itinedtately verify that the fire detector zones listed on Table 3.12 6.
After each refueling are operable for the respective outage and prior to system fire zone (s) for which the startup, perform a test 4
panel (s) provided alternate from the alternate shutdown capability, shutdown panel to verify that the relief valve If a fire detection zone cannot be l solenoids of the Automatic deteri.ined operable, establish an Depressurization System hourly fire watch patrol to it sect (ADS) actuate.
the affected zone (s).
7.
Once each refueling outage, the diesel generator control circuits shall be isolated from the Cable Spreading Room and the diesel generator started.
Amendment No. 143 206
.t f
Table 3.12 i
Fire Detector Zones Associated with.
Alternate Shutdown Panels t
Alternate Shutdown System Fire Zone Detection Panel /Det. Zones.
l t
Core Spt-ay 1.1 &.2 C-224/4A RHR 1.1 &.2 C223/3C RBCCH 1.21 &.22 C-222/2A & 2B SSH 5.1 &.2 &.3 N/A HPCI 1.3 &.4 C-223/3D & 3E.
RCIC 1.5 C-223/3A & 3B I
ADS 1.1 &.2 C-224/4A DGS 4.1 &.3 C-93/1 & 2 1
i Amendment No. 143.
g 206A -- l -
+
BASIS:
I 3/4.12 Fire Protectiga The alternate shutdown system, independent of cabling and equipment in the Cable Spreading Room, is provided to effect safe shutdown of Pilgrim in the event of a fire in the Cable Spreading Room. This is accomplished by installing isolation switches for safety-related equipment that will provide the capability for the plant operators to reach a safe shutdown condition.
These switches will isolate their associated equipment from the CSR cables, thus transfer control from the Control Room to the Irei emergency shutdown stations outside the CSR.
These isolation switches o located in alternate shutdown panels and are located as close as practical to the equipment or switchgear they serve.
An emergency shutdown procedure, which is compatible with the design modifications and slant operator availability, provides step-by-step actions to initiate safe s1utdown operation. Operator actions to isolate safety-related cables passing through the CSR is initiated as soon as a fire which is not immediately extinguishable is detected and confirmed in the CSR.
Alternate shutdown panels are provided for the following systems:
a.
Core Spray b.
RHR c.
RBCCH d.
Salt Service Hater e.
HPCI f.
RCIC g.
Automatic Depressurization System h.
Diesel Generators Inoperability of the above listed systems does not require eritry into LCO action statements for the alternate shutdown panels.
A surveillance frequency of once per cyc't s considered prudent and more frequent testing not warranted.
The freqG ng of once per refueling outage for testing the diesel generators prevents unnecessarily rendering them inoperable during normai power operation. The frequency of once per refueling outage for the Automatic Depressurization System is consistent with the existing surveillance frequency for this system.
Requiring this surveillance to be performed during a refueling outage will also assure that plant conditions will allow for safe access to the ADS solenoids.
(The next page is 206K) l Amendment No. Ild,143 206B l
m 6.0 ACHINISTRATIVE CONTROLS 2.
When the unit is in an operational mode o'ter than cold shutdown or refueling, a person holding a Senior Reactor Operator License shall be present in the control room at all times.
In addition to this Senior Operator, a Licensed Operator or Senior Operator shall be present at the controls when fuel is in the vessel.
3.
At least two Licensed Operators shall be pr9sent in the control room during reactor startup, schdoled reactor shutdown and during recovery from reactor trips.
4.
An individual qualified in radiation protection procedures shall be on site when fuel is in the reactor.
5.
ALL CORE ALTERATIONS performed while fuel is in the reactor vessel after the initial fuel loading shall be directly supervised by either a licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling who has no other concurrent responsibilities during this operation.
6 Deleted 7.
The Chief Operating Engineer, Nuclear Hatch Engineers, and Nuclear Operations Supervisors she.11 hold a Senior Reactor Operator License.
The Nuclear Plant Operators shall hold a Reactor Operator License.
6.3 UNIT STAFF OUAUFICATIONS The qualifications with regard to educational and experience backgrounds of the unit staff at the time of appointment to the active position shall meet the requirements as described in the American Nailonal Standards Institute N18.1-1971, " Selection and Training of Personnel for Nuclear Power Plants." In addition, the individual performing the function of Radiation Protection Hanager shall meet or exceed the qualifications of Regu16 tory Guide 1.8, September,1975.
6.4 TRAINING A retraining and replacement training program for the unit staff shall be maintained under the direction of the Nuclear Training Department Manager.
The training programs for the licensed personnel shall meet or exceed the requirements and recommendations of Section 5.5 of ANSI N18.1-1971 and 10CFR Part 55.
Amendment No. 41, 46, 69, 88, 122, 125. 132, 143 209
r e 6.5 A.6 RESPONSIBitITIES (Continued) e.
Review of facility operations to detect potential safety
- hazards, f.
Review of the Station Security Plan and implementing procedures and changes to the plan and procedures.
g.
Review of the Emergency Plan and implementing procedures and changes to the plan and procedures.
h.
Performance of special reviews and investigations and reports thereon as requested by the Nuclear Safety Review and Audit Committee (NSRAC) Chairman.
1.
Investigation of all violatiens of the Technical Specifications and shall prepare and forward a report covering evaluation and recor.mendations to prevent recurrence to the Station Director, the NSRAC Chairman, 3ru +he Senior Vice President - Nuclear, j,
% e the Station Fire Protection Program and 4njiews t'ng procedures and changes to the Program and implementing procedures.
The O'C Chairman may appeint subcommittees composed of personnel who are not members of ORC to perform staff work necessary to th6 efficient functioning of ORC.
7.
AUTHORITY a.
Recommend in writing to the Station Director the approval or disapproval 6 items considered under 6.5.A.6(a) through (d) above.
b.
Render determinations in writing with regard to whether or not each item considered under 6.S.A.6(a) through (d) above constitutes an unreviewed safety question.
c.
Provide written notification within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to the Station Director, the Nuclear Safety Review and Audit Committee, and the Senior Vice President - Nuclear of disagreement between the ORC Members and the ORC Chairman. The Station Director shall have responsibility for resolution of such disagreements.
8.
RECORDS The ORC shall maintain writtsn minutes of each meeting and copies shall be forwarded to the Station Director and the NSRAC Chairman.'
Amendment No. 29, 46, 69, 88, 122, 125, 132, 143 212.
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