ML17264A979
| ML17264A979 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 08/05/1997 |
| From: | ROCHESTER GAS & ELECTRIC CORP. |
| To: | |
| Shared Package | |
| ML17264A976 | List: |
| References | |
| AP-RCS.3, NUDOCS 9708130145 | |
| Download: ML17264A979 (83) | |
Text
EOP: TITLE:
REV: 7 AP-RCS.3 HIGH REACTOR COOLANT ACTIVITY PAGE 1 of 6 ROCHESTER GAS AND ELECTRIC CORPORATION GINNA STATION CONTROLLED COPY NUMBER RESPONSIB MANAGER EFFECTIVE DATE CATEGORY 1.0 REVIEWED BY:
9708iSADOCpi45 9'70805 PDR OCR 05000244 F
EOP:
REV 7 AP-RCS.3 HIGH REACTOR COOLANT ACTIVITY PAGE 2 of 6 A. PURPOSE This procedure provides guidance necessary to operate the plant with indication of high reactor coolant activity.
B. ENTRY CONDITIONS/SYMPTOMS
- 1. SYMPTOMS The symptoms of HIGH REACTOR COOLANT ACTIVITY are;
- a. Unexplained increase in letdown line monitor, R-9, or
- b. Sampling indicates I-131 equivalent GREATER THAN 1.0 uCi/gm, or
- c. Sampling indicates gross degassed activity GREATER THAN 20 uCi/gm, or
- d. Sampling, indicates that total specific activity exceeds 100/E.
I' I
EOP: TITLE:
REV: 7 AP-RCS.3 HIGH REACTOR COOLANT ACTIVITY PAGE 3 of 6 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED CAUTION o IF LETDOWN FLOW EXCEEDS 60 GPM, THEN LOCALLY MONITOR D/P ACROSS THE CVCS DI(S) TO VERIFY THAT FLOW IS CONTINUING AND THAT RELIEF VALVE, V-209, HAS NOT LIFTED.
o LETDOWN FLOW THROUGH THE DI'S SHOULD BE LIMITED TO 90 GPM.
- * * * * * * * * * 'k * * * * * * * * * * * * * * * * * * * * * * * * * * * * *
- NOTE: Conditions should be evaluated for site contingency reporting (Refer to EPIP-1.0, GINNA STATION EVENT EVALUATION AND CLASSIFICATION.
1 Verify RCS Activity:
- b. RCS, activity - GREATER THAN b. IF normal activity verified, NORMAL (Check with RP Department THEN direct 1&C to check for normal activity) operability of R-9, letdown line monitor, AND return to normal operations.
y EOP: TITLE:
REV: 7 AP-RCS ~ 3 HIGH REACTOR COOLANT ACTIVITY PAGE 4 of 6 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 2 Increase Letdown Flow To 60 GPM
- a. Verify deborating DI isolated- a. Place AOV-244 in bypass position.
DIVERT VLV CATION DEBOR DI AOV-244 IN BYPASS POSITION
- b. Place letdown controllers in MANUAL at 60% open
~ TCV-130
~ PCV-135
- c. Increase letdown flow as follows:
- 1) Close letdown orifice valve (AOV-200A or AOV-200B)
- 2) Immediately open 60 gpm letdown orifice valve, W AOV-202
- d. Adjust low pressure letdown pressure to approximately 250 psig
- e. Place TCV-130 in AUTO at 105'F
- f. Place PCV-135 in AUTO at 250 psig
- g. Adjust charging pump speed and HCV-142 as necessary
e C)
EOP: TITLE:
REV: 7 AP-RCS.3 HIGH REACTOR COOLANT ACTIVITY PAGE 5 of 6 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 3 Check Letdown Line Monitor, Evaluate conditions to determine R LESS THAN 200 MR/HR whether local radiation emergency ABOVE BACKGROUND exists (Refer to EPIP 1-13, LOCAL RADIATION EMERGENCY).
CAUTION PLACING A NEW DI IN SERVICE MAY RESULT IN A POSITIVE OR NEGATIVE REACTIVITY ADDITION DUE TO A BORON CHANGE.
4 Direct RP Tech To Sample IF DI efficiency is NOT acceptable, Letdown DI Efficiency- THEN place a new mixed bed in DECONTAMINATION FACTOR service (Refer to S-3.2B, PLACING A GREATER THAN 10 MIXED BED DEMINERALIZER IN SERVICE BORON CONCENTRATION DIFFERENT THAN RCS).
5 Evaluate AUX BLDG Radiation Levels:
- a. Direct RP Tech to survey AUX BLDG'.
Check AUX BLDG radiation b. Perform the following:
monitors - NORMAL
- 1) Direct RP Tech to survey AUX
~ R-4 BLDG areas as necessary.
~ R-9
~ R-10B 2) Evaluate conditions to
~ R-13 determine whether local
~ R-14 radiation emergency exists (Refer to EPIP 1-13, LOCAL RADIATION EMERGENCY).
0 0
EOP: 71ILE:
REV: 7 AP-RCS.3 HIGH REACTOR COOLANT ACTIVITY PAGE 6 of 6 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 6 Determine Can If Plant Operation Continue (Consult Plant IF plant shutdown is required,= THEN refer to 0-2.1, NORMAL SHUTDOWN TO staff if necessary)
OPERATION CAN CONTINUE HOT SHUTDOWN.
NOTE: Refer to 0-9.3, NRC IMMEDIATE NOTIFICATION, for reporting requirements.
7 Notify Higher Supervision
-END-
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Item: ADAMS Document Library: ML ADAMS"HQNTAD01 ID 011700281:1
Subject:
'mproved Technical Specifications (ITS), Amendment No.-79; ITS Bases Revision 20; an d Technical Requirements Manual Revision 17. Document Date: 06/07/2001 Body:
ADAMS DISTRIBUTION NOTIFICATION.
A001 - OR Submittal: General Distribution for Power Reactor (10 CFR 50) Submittals
Title:
Improved Technical Specifications (ITS), Amendment No. 79; ITS Bases Revision 20; and Technical Requirements Manual Revision 17.
Docket Number: 05000244 Document Date: 06/07/2001 Author Name: Harding T L Author Affiliation:Rochester Gas 8 Electric Corp Page 1
t 7
Distri-1.txt Addressee Name:
Addressee Affiliation:
Document Type: Manual Availability: Publicly Available Document Sensitivity: Non-Sensitive Comment:
Electronic Recipients can RIGHT CLICK and OPEN the first Attachment to View the Document in ADAMS. The Document may also be viewed by searching for Accession Number ML011700281.
Page 2
ROCHESTER GAS Ec, 8
ELECTRIC CORPORATION DOCUMENT CONTROL Document Transmittal Transmittal: 200106005 053 To: DOCUMENT CONTROL DESK Date: 06/07/2001 US NRC Holder: 1217 ATT: GUY VISSING PROJECT DOCTORATE WASHINGTON DC 20555 l-l LISTED BELOW ARE NEW OR REVISED CONTROLLED DOCUMENTS. REVIEW THE ATTACHED INSTRUCTION LETTER (IF APPLICABLE), AND UPDATE YOUR CONTROLLED DOCUMENT IN ACCORDANCE WITH EP-3-S-0901 Doc Id Doc Title Rev Rev Date Qty TRM TECHNICAL REQUIREMENTS MANUAL FOR 017 06/08/2001 1 GINNA STATION Receipt Acknowledgement and Action Taken:
Signature Date Return ENTIRE FORM to: Rochester Gas & Electric Corporation Ginna Station Records Management 1503 Lake Road Ontario, New York 14519
Rochester Gas and Electric Corporation Inter-Office Correspondence June 7, 2001
Subject:
Improved Technical Specifications (ITS), Amendment No. 79; ITS Bases Revision 20; and Technical Requirements Manual (TRM) Revision 17 Tot Distribution Attached is Amendment No. 79 to the Ginna Station Improved Technical Specifications (ITS). This Amendment revises the ITS requirements for the storage of new and spent fuel within the Spent Fuel Pool. This Amendment was approved by the NRC in a safety evaluation dated December 7, 2000.
In addition to the Amendment, the associated changes to the ITS Bases and the Technical Requirements Manual are included. Instructions for the necessary changes to your controlled copy of the ITS are attached.
These changes are considered effective June 8, 2001.
Please contact Tom Harding (extension 3384) ifyou have any questions.
rx-~~
Thomas L. Harding
Attachment A Please replace the following pages of your controlled copy of the ITS as follows:
Volume Section Remove Insert Attachment A to License First Page First Page Attachment A to License LEP-i through LEP-vii LEP-i through LEP-vii II Chapter 3.7 (LCO) 3.7-29 3.7-29 II Chapter 3.7 (LCO) 3.7-30 3.7-30 II Chapter 3.7 (LCO) 3.7-31 3.7-31 II Chapter 3.7 (LCO) 3.7-31a 3.7-31 a II Chapter 3.7 (LCO) 3.7-31b 3.7-31b II Chapter 3.7 (LCO) 3.7-31c II Chapter 3.7 (LCO) 3.7-31d II Chapter 3.7 (LCO) 3.7-31e II Chapter 3.7 (LCO) 3.7-31f II Chapter 3.7 (LCO) 3.7-31g II Chapter 3.7 (LCO) 3.7-3lh II Chapter 3.7 (LCO) 3.7-31i Chapter 3.7 (Bases) B 3.7.12-1 through B 3.7.12-3 B 3.7.12-1 through B 3.7.12-4 Chapter 3.7 (Bases) B 3.7.13-1 through 8 3.7.13-6 B 3.7.13-1 through B 3.7.13-7 Chapter 4.0 4.0-2 4.0-2 Chapter 4.0 4.0-3 4.0-3 III TRM First Page First Page III TRM LEP-i LEP-i III TRM LEP-ii LEP-ii TRM Chapter 3.7 (TR) TR 3.7.7-1 TR 3.7.7-1 TRM Chapter 3.7 (TR) TR 3.7.7-2 TR 3.7.7-2 III TRM Chapter 3.7 (Bases) TR 3.7.7-1 TRB 3.7.7-1 III TRM Chapter 3.7 (Bases) TR 3.7.7-2 TRB 3.7.7-2 III TRM Chapter 3.7 (Bases) TR 3.7.7-3 TRB 3.7.7-3 III TRM Chapter 3.7 (Bases) TR 3.7.7-4 TRB 3.7.7-4 III TRM Chapter 3.7 (Bases) TR 3.7.7-5 TRB 3.7.7-5 III TRM Chapter 3.7 (Bases) TRB 3.7.7-6
a+i R.E. Ginna Nuclear Power Plant Technical Requirements Ma al TRM Revision 16 Responsibl Manager:
George robel Eff ctive Date: 6 /(/goo 6/1/2001 Controlle Copy No.
R.E. Ginna Nuclear Power Plant TRM-1 Revision 6
List of Effective Pages (TRM)
Page Rev. Page Rev. Page Rev. Page Rev.
TOC Section 4.0 15 TR 3.3.4-1 14 TR 3.7.2-1 14 TR 4.0-1 14 15 TR 3.3.4-2 14 TR 3.7.2-2 14 TR 3.3.4-3 14 TR 3.7.2-3 14 TR 3.3.4Q 14 TR 3.7.2C 14 Section 5.0 Section 1.0 TR 1.0-1 14 TR 3.3.4-5 TR 3.3.4-6 TR 3.3.4-7 TR 3.3.5-1 TR 3.3.5-2 14 14 14 14 14 TR 3.7.2-5 TR 3.7.3-1 TR 3.7.3-2 TR 3.7.3-3 TR 3.7A-1 14 14 14 14 14 TR 5.1-1 TR 5.1-2 '6 16 Section 2.0 TR 3.3.5-3 14 TR 3.7.4-2 14 TR 3.7.4-3 14 TR 2.1-1 14 TR 3.7.4-4 14 Section 3.4 TR 3.7.5-1 14 TR 3.7.5-2 14 Section 3.0 TR 3.4.1-1 14 TR 3.7.5-3 14 TR 3.4.1-2 14 TR 3.7.5-4 14 TR 3.0-1 14 TR 3.4.2-1 14 TR 3.7.5-5 14 TR 3.0-2 14 TR 3.4.2-2 14 TR 3.7.6-1 14 TR 3.4.2-3 14 TR 3.7.6-2 14 TR 3.4.3-1 14 TR 3.7.7-1 14 Section 3.1 TR 3.4.3-2 14 TR 3.7.7-2 14 TR 3.4.3-3 14 TR 3.7.8-1 15 TR 3.1.1-1 14 TR 3.4.34 14 TR 3.7.8-2 15 TR 3.1.1-2 14 TR 3.4.3-5 14 TR 3.1.1-3 14 TR 3.4.3-6 14 TR 3.1.2-1 14 Section 3.8 TR 3.1.2-2 14 Section 3.5 TR 3.8.1-1 14 TR 3.8.2-1 14 Section 3.2 TR 3.5-1 14 TR 3.8.2-2 14 TR 3.2-1 14 Section 3.6 Section 3.9 Section 3.3 TR 3.6-1 14 TR 3.9.1-1 14 TR 3.9.2-1 14 TR 3.3.1-1 14 TR 3.9.2-2 14 TR 3.3.1-2 14 Section 3.7 TR 3.9.3-1 '14 TR 3.3.2-1 14 TR 3.9.4-1 14 TR 3.3.2-2 14 TR 3.7.1-1 14 TR 3.9.4-2 14 TR 3.3.3-1 14 TR 3.7.1-2 14 TR 3.7.1-3 14 R.E. Ginna Nuclear Power Plant LEP-i Revision 16
i ist of Effective Pages (TRM Bases)
Page Rev. Page Rev. Page Rev. Page Rev.
TOC Section 3.7 15 TRB 3.7.1-1 14 TRB 3.7.2-1 14 TRB 3.7.3-1 14 Section 2.0 TRB 3.7.4-1 14 TRB 3.7.5-1 14 TRB 3.7.6-1 14 Section 3.0 TRB 3.7.7-1 14 TRB 3.7.7-2 14 TRB 3.7.7-3 14 Section 3.1 TRB 3.7.7-4 14 TRB 3.7.7-5 14 TRB 3.7.8-1 15 Section 3.2 TRB 3.7.8-2 15 TRB 3.7.8-3 15 Section 3.3 Section 3.8 Section 3.4 Section 3.9 Section 3.5 TRB 3.9.1-1 14 TRB 3.9.1-2 14 TRB 3.9.1-3 14 Section 3.6 TRB 3.9.2-1 14 TRB 3.9.3-1 14 TRB 3.9.4-1 14 TRB 3.9.4-2 14 TRB 3.9.4-3 14 TRB 3.9.4Q 14 Section 4.0 Section 5.0 R.E. Ginna Nuclear Power Plant LEP-ii Revision 16
SFP Cooling System TR 3.7.7 3.7 PLANT SYSTEMS TR 3.7.7 Spent Fuel Pool (SFP) Cooling System TR 3.7.7 The SFP cooling system shall be maintained as follows:
- a. The SFP water temperature shall be < 150'F; and
- NOTE-The SFP heat load and SFP cooling capabilities are determined by Nuclear Engineering Services.
APPLICABILITY: Whenever any irradiated fuel assembly is stored in the SFP.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. SFP temperature not A.1 Suspend movement of Immediately within limit. irradiated fuel assemblies from the reactor to the SFP.
AND A.2 Initiate action to restore SFP Immediately temperature to within limit.
B. One required SFP cooling B.1 Suspend movement of Immediately system inoperable. irradiated fuel assemblies from the reactor to the SFP.
AND 8.2 Restore a second SFP 14 days cooling system to OPERABLE status.
R.E. Ginna Nuclear Power Plant TR 3.7.7-1 Revision 14
SFP Cooling System TR 3.7.7 CONDITION REQUIRED ACTION COMPLETION TIME C. Both required SFP C.1 Perform TSR 3.7.7.1. Once within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> cooling systems and every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> inoperable. thereafter AND C.2 Restore one SFP cooling Prior to the SFP system to OPERABLE water temperature status. exceeding 120'F SURVEILLANCE REQUIREMENTS.
SURVEILLANCE FREQUENCY TSR 3.7.7.1 Verify SFP temperature is within limit. 24 Hours TSR 3.7.7.2 Verify power available to the two OPERABLE SFP 7 days cooling systems.
R.E. Ginna Nuclear Power Plant TR 3.7.7-2 Revision 14
Spent Fuel f(SFP) Cooling System TRB 3.7.7 B 3.7 PLANT SYSTEMS TRB 3.7.7 Spent Fuel Pool (SFP) Cooling System BASES BACKGROUND The Spent Fuel Pool (SFP) Cooling System is designed to remov'e from the SFP the heat generated by stored spent fuel assemblies.
The SFP Cooling System consists of three cooling loops (see Figure TRB 3.7.7-1). The primary loop (loop B) is made up of the SFP Pump B, SFP Heat Exchanger B, and piping. The backup loops include installed loop A with the SFP Pump A, SFP Heat Exchanger A, and piping, and a standby loop with the Skid-Mounted SFP Pump, SFP Standby Heat Exchanger, and hoses. Service water (SW) circulates through the shell while SFP water circulates through the tubes of all three heat exchangers. The SFP Cooling System loop B is sized for 100% of the design SFP heat load.
The SFP Cooling System loop A and the standby loop are each capable of removing the normal basis heat load and when operated in parallel can remove more than 100% of the design heat load (Ref. 1). Motor-operated valves provide automatic and remote manual isolation of the SW supply to the heat exchangers associated with the SFP Cooling System loops A and 8 and the Component Cooling Water Heat Exchangers. These valves close automatically upon coincidence of safety injection and loss of offsite power. Handwheels are provided for manual operation.
The SFP Cooling System is designed to maintain the pool ( 120'F during normal refueling conditions and s 150'F during full core discharge operations (Ref: 1). The cooling systems can take a suction from either near the surface of the SFP and/or at a point above the irradiated fuel assemblies, such that a failure of any pipe in the system will not drain the pool to a point where the fuel would be exposed. The cooling system return line to the pool also contains a 0.25 inch vent hole located near the SFP surface level to prevent siphoning. Control board alarms exist with respect to the SFP level and temperature. These features all help to prevent inadvertent draining of the SFP.
APPLICABLE For structural integrity reasons, the pool water temperature is not to SAFETY exceed 180'F. In order to provide sufficient time to take corrective action ANALYSES in the event of a SFP Cooling System failure, the pool temperature limit is not to exceed 150'F for all modes of operation including a full core discharge. The requirement for two 100% SFP Cooling Systems is based on being able to provide cooling following either a loss of offsite power or an active single failure within the SFP Cooling System.
R.E. Ginna Nuclear Power Plant TRB 3.7.7-1 Revision 14
Spent Fuel P(SFP) Cooling System TRB 3.7.7
' TR The SFP water temperature is required to be < 150'F. The specified water temperature provides sufficient margin to the assumptions of the SFP structural integrity. As such, it is the maximum allowed while storing irradiated fuel assemblies within the SFP.
To ensure that the SFP temperature can be maintained within the required limit, two SFP Cooling Systems must be OPERABLE. This requirement provides for 100% backup capability assuming either a loss of offsite power or an active single failure within the SFP Cooling System.
SFP Cooling System loop A and loop B are considered OPERABLE when the respective pump has power available and the respective heat exchanger is capable of providing cooling water with the ability to remove the decay heat load of the irradiated fuel assemblies stored within the SFP. The standby loop is considered OPERABLE when the Skid-Mounted SFP Pump has power connected to a temporary power source, the SFP Standby Heat Exchanger has been staged, and the temporary hoses have been connected and leak checked. During a full core offload the SFP Cooling System loop A and standby loop may be operated in parallel to comprise one of the required OPERABLE systems. Loop A and the standby loop have the ability to utilize fire water for cooling in lieu of SW to provide for increased redundancy. The temporary electrical power source for the standby loop pump may also be varied to provide for increased redundancy.
Also included in the determination of OPERABILITY are all necessary support systems not addressed by this TR (e.g., service water, fire water, electrical). Single active failures are not required to be considered within the support systems for the purpose of this TR; however, the support systems must be capable of performing their support function per the definition of OPERABLE-OPERABILITYin ITS Section 1 1. ~
The TR is modified by a note. The SFP heat load from the irradiated fuel assemblies stored within the SFP is a variable based on the total number of assemblies stored, the power history of the individual assemblies, and the time since the assemblies were last irradiated. The heat removal capabilities of the individual SFP cooling loops is also a variable based on the temperature and flow rate of the cooling source, SW or fire water.
The SFP heat load and the SFP Cooling System heat removal capabilities are determined by Nuclear Engineering Services and provided, as necessary, based on plant conditions.
R.E. Ginna Nuclear Power Plant TRB 3.7.7-2 Revision 14
Spent Fuelbt (SFP) Cooling System TRB 3.7.7 APPLICABILITY This TR applies whenever any irradiated fuel assembly is stored in the SFP, to provide sufficient margin to the assumptions of the SFP structural integrity. Specific requirements applicable during a full core discharge are covered by TR 3.9.4.
ACTIONS A.1 and A.2 If the SFP temperature is not within limit, steps should be taken to preclude the assumptions of the SFP structural integrity from being exceeded. Suspending any operation that would increase SFP decay heat load, such as discharging a fuel assembly from the reactor to the SFP, is a prudent action under this condition. Therefore, actions shall be taken immediately to suspend movement of irradiated fuel assemblies from the core to the SFP.
With the potential for exceeding the assumptions of the SFP structural integrity, corrective actions to restore the SFP temperature to within limit shall be initiated immediately.
B.t and B 2 With one of the required SFP Cooling Systems inoperable, suspending any operation that would increase SFP decay heat load, such as discharging an irradiated fuel assembly from the reactor to the SFP, is a prudent action under this condition. Therefore, actions shall be taken immediately to suspend movement of irradiated fuel assemblies from the core to the SFP.
With the suspension of fuel discharge into the SFP then a second SFP Cooling System must be restored to OPERABLE status within 14 days.
In this condition the remaining OPERABLE SFP Cooling System is adequate to remove the decay heat load. The 14 day Completion Time is adequate to perform typical maintenance activities associated with a SFP Cooling System and takes into account the large heat sink capabilities of the SFP.
R.E. Ginna Nuclear Power Plant TRB 3.7.7-3 Revision 14
Spent FuetOtt (SFP) Cooling System TRB 3.7.7 C.1 and C.2 If no SFP Cooling System is OPERABLE, there will be no forced cooling of the SFP and as such the SFP temperature must be monitored more frequently until the cooling is restored. This monitoring is accomplished by performing surveillance TSR 3.7.7.1 to verify SFP temperature is within limit. The Completion Time of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> thereafter is sufficient due to the large heat sink of the SFP and slow heatup rate.
Actions must also be initiated to restore one SFP Cooling System to
,OPERABLE status prior to the SFP temperature exceeding 120'F. The 120'F is not a safety requirement but is a limit set for normal operation and provides adequate margin to the 150'F limit.
SURVEILLANCE TSR 3.7.7.1 REQUIREMENTS This TSR verifies that the SFP temperature is within the required limit.
The temperature in the SFP must be checked periodically to ensure the SFP structural integrity is met. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency is appropriate due to the large volume of water in the pool and the relatively slow heatup rate.
Verification of SFP water temperature is normally accomplished by the use of TIA-635, which also provides a SFP high temperature alarm to alert the operators of an increasing SFP temperature.
T~SR 3.7.7.
Verification that the required pumps are OPERABLE ensures that an additional SFP pump can be placed in operation, if needed, to maintain decay heat removal. Verification is performed by verifying proper breaker alignment and power available to the required pumps. The Frequency of 7 days is considered reasonable in view of other administrative controls available and has been shown to be acceptable by operating experience.
REFERENCES 1. UFSAR, Section 9.1.3.
R.E. Ginna Nuclear Power Plant TRB 3.7.7-4 Revision 14
Spent Fuetept (SFP) Cooling System TRB 3.7.7 I
8664 I
Spent Fuel gl SW Pool Standby HX Skid-mounted SFP Pump l SW/Fire 781 782 I
l 8614 aes 8662 gf s r a +e a r i SFP HX A 8654 SFP Pump A t SW/Fire 8663 SFP HX B SFP Pump B 8657 Legend:
Loop e Flowpalh Loop A Flowpath For illustration only Standby Loop Flowpath Figure TRB 3.7.7-1 SFP Cooling System R.E. Ginna Nuclear Power Plant TRB 3.7.7-5 Revision 14
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Item: ADAMS Package Library: ML ADAMS"HQNTAD01 ID: 011590048
Subject:
Revision to Emergency Plan Implementing Procedures R. G. Ginna Nuclear Power Plant
. Document Date: 06/04/2001 Body:
ADAMS DISTRIBUTION NOTIFICATION.
A045- OR Submittal: Emergency Preparedness Plans, Implementing Procedures, Corre
-'pondence
Title:
Revision fo Emergency Plan Implementing Procedures R. G. Ginna Nuclear Power Plant.
Docket Number: 05000244 Document Date: 06/04/2001 Page 1
Distri-l.txt Author Name: Poifleit P S Author Affiliation:Rochester Gas 8: Electric Corp Addressee Name: Vissing G S Addressee Affiliation:NRC/Document Control Desk Addressee Affiliation:NRC/NRR/ DLPM/LPDI Document Type: Emergency Preparedness-Emergency Plan Implementing Procedures Availability: Publicly Available Document Sensitivity: Non-Sensitive Comment:
Electronic Recipients can RIGHT CLICK and OPEN the first Attachment to View the Document in ADAMS. The Document may also be viewed by searching for Accession Number ML011590048.
Page 2
~ t iiruuzzz t Zii!"'I Zsuu'!g',
zzvuzz ROCHESTER GAS AND ELECTRIC CORPORATION o 89 EAST AVENUE, ROCHESTER, N.Y. 14649-0001 TELEPHONE ARE* cooE 716 546.2700 June 4, 2001 U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Attn: Mr. Guy S. Vissing (Mail Stop 14D11)
Project Directorate I-1
Subject:
Revision to Emergency Plan Implementing Procedures R.E. Ginna Nuclear Power Plant Docket No. 50-244
~ C Gentlemen:
In accordance with 10 CFR 50.4(b)(5), enclosed are revisions to Ginna Station Emergency Plan Implementing Procedures (EPIPs).
We have determined, per the requirements of 10 CFR 50.54(q), that these procedure changes do not decrease the effectiveness of our Nuclear Emergency Response Plan.
Very tr ours, I~ I jl I 5 ~
Peter S. Polfleit Corporate Nucl r E ergency Planner Enclosures 5 Q 5
xc: USNRC Region 1 (2 copies of letter and 2 copies of each procedure)
Resident Inspector, Ginna Station (1 copy of letter and 1 copy of each procedure)
RGSE Nuclear Safety and Licensing (1 copy'of letter)
Dr. Robert C. Mecredy (2 copies of letter only)
PSP/jtw
PROCEDURE REVISION NUMBER EPIP 1-5 43 EPIP 2-1 19 EPIP 2-5 EPIP 2-17 EPIP 3-1 15 EPIP 4-7 17 EPIP 5-1 21 EPIP 5-7 32
I (t
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Item: ADAMS Document Library: ML ADAMS"HQNTAD01 ID: 011630044:1
Subject:
R. E. Ginna Nuclear Power Plant Technical Requirements Manual TRM, Revision 'l6. Doc ument Date: 05/31/2001 Body:
ADAMS DISTRIBUTION NOTIFICATION.
A001 - OR Submittal: General Distribution for Power Reactor (10 CFR 50) Submittals
Title:
R. E. Ginna Nuclear Power Plant Technical Requirements Manual TRM, Revision 16.
Docket Number: 05000244 Document Date: 05/31/2001 Page 1
Distri- l.txt Author Name:
Author Affiliation:Rochester Gas & Electric Corp Addressee Name: Vissing G Addressee Affiliation:NRC/Document Control Desk Addressee Affiliation:NRC/NRR/DLPM/LPD1 Document Type: Manual Document Type: Technical Specification Availability: Publicly Available Document Sensitivity: Non-Sensitive Comment:
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ROCHESTER GAS 6 ELECTRIC CORPORATION DOCUMENT CONTROL Document Transmittal Transmittal: 200105037 053 To: DOCUMENT CONTROL DESK Date: 05/31/2001 US NRC Holder: 1217 ATT: GUY VISSING PROJECT DOCTORATE 1-1 WASHINGTON DC 20555 LISTED BELOW ARE NEW OR REVISED CONTROLLED DOCUMENTS. REVIEW THE ATTACHED INSTRUCTION LETTER (IF APPLICABLE), AND UPDATE YOUR CONTROLLED DOCUMENT IN ACCORDANCE WITH EP-3-S-0901 Doc Id Doc Title Rev Rev Date Qty TRM TECHNICAL REQUIREMENTS MANUAL FOR 016 06/01/2001 1 GINNA STATION Receipt Acknowledgement and Action Taken:
Signature Date Return ENTIRE FORM to: Rochester Gas & Electric Corporation Ginna Station Records Management 1503 Lake Road Ontario, New York 14519
Rochester Gas and Electric Corporation Inter-Office Correspondence May 31, 2001
Subject:
Technical Requirements Manual (TRM) Revision 16 To: Distribution Attached are the revised pages for the Ginna Station Technical Requirements Manual (TRM). The change, included within Revision 16 is summarized as follows:
Table TR 5.1-1 was revised to add a new more conservative, letdown flow rate based, requirement for reactor coolant DOSE EQUIVALENTI-131 specific activity as the result of ACTION Report 2000-0450.
Revision 16 of the TRM is considered effective June 1, 2001. Instructions for the necessary changes to your controlled copy of the ITS are as follows:
Volume Section Remove Insert III TRM Cover Page Cover Page 111 TRM LEP-i LEP-I III TRM LEP-ii LEP-ii III TRM Chapter 5.0 TRM 5.1-1 TRM 5.1-1 III TRM Chapter 5.0 TRM 5.1-2 TRM 5.1-2 Please contact Tom Harding (extension 3384) ifyou have any que tions.
Thomas L. arding
I I R.E. Ginna Nuclear Power Plant Technical Requirement anual TRM Revisio 15 Respo sible Manager:
.~ George r el Effective Date: I<)/c>/ 8 10/11/2000 Controlled Copy R.E. Ginna Nuclear Power Plant TRM-1 Revision 15
0 s f 0
List of Effective Pages (TRM)
Page Rev. Page Rev. Page Rev. Page Rev.
TOC Section 4.0 15 TR 3.3.4-1 14 TR 3.7.2-1 14 TR 4.0-1 14 15 TR 3.3.4-2 14 TR 3.7.2-2 14 TR 3.3.4-3 14 TR 3.7.2-3 14 TR 3.3.4-4 14 TR 3.7.2-4 14 Section 5.0 Section 1.0 TR 3.3.4-5 14 TR 3.7.2-5 14 TR 3.3.4-6 14 TR 3.7.3-1 14 TR 5.1-1 14 TR 1.0-1 14 TR 3.3.4-7 14 TR 3.7.3-2 14 TR 5.1-2 14 TR 3.3.5-1 14 TR 3.7.3-3 14 TR 3.3.5-2 14 TR 3.7.4-1 14 Section 2.0 TR 3.3.5-3 14 TR 3.7.4-2 14 TR 3.7.4-3 14 TR 2.1-1 14 TR 3.7.4-4 14 Section 3.4 TR 3.7.5-1 14 TR 3.7.5-2 14 Section 3.0 TR 3.4.1-1 14 TR 3.7.5-3 14 TR 3.4.1-2 14 TR 3.7.5-4 14 TR 3.0-1 14 TR 3.4.2-1 14 TR 3.7.5-5 14 TR 3.0-2 14 TR 3.4.2-2 14 TR 3.7.6-1 14 TR 3.4.2-3 14 TR 3.7.6-2 14 TR 3.4.3-1 14 TR 3.7.7-1 14 Section 3.1 TR 3.4.3-2 14 TR 3.7.7-2 14 TR 3.4.3-3 14 TR 3.7.8-1 15 TR 3.1.1-1 14 TR 3.4.3-4 14 TR 3.7.8-2 15 TR 3.1.1-2 14 TR 3.4.3-5 14 TR 3.1.1-3 14 TR 3.4.3-6 14 TR 3.1.2-1 14 Section 3.8 TR 3.1.2-2 14 Section 3.5 TR 3.8.1-1 14 TR 3.8.2-1 14 Section 3.2 TR 3.5-1 14 TR 3.8.2-2 14 TR 3.2-1 14 Section 3.6 Section 3.9 Section 3.3 TR 3.6-1 14 TR 3.9.1-1 14 TR 3.9.2-1 14 TR 3.3.1-1 14 TR 3.9.2-2 14 TR 3.3.1-2 14 Section 3.7 TR 3.9.3-1 14 TR 3.3.2-1 14 TR 3.9.4-1 14 TR 3.3.2-2 14 TR 3.7.1-1 14 TR 3.9.4-2 14 TR 3.3.3-1 14 TR 3.7.1-2 14 TR 3.7.1-3 14 R.E. Ginna Nuclear Power Plant LEP-i Revision 15
List of Effective Pages (TRM Bases)
Page Rev. Page Rev. Page Rev. Page Rev.
TOC Section 3.7 15 TRB 3.7.1-1 14 TRB 3.7.2-1 14 TRB 3.7.3-1 14 Section 2.0 TRB 3.7.4-1 14 TRB 3.7.5-1 14 TRB 3.7.6-1 14 Section 3.0 TRB 3.7.7-1 14 TRB 3.7.7-2 14 TRB 3.7.7-3 14 Section 3.1 TRB 3.7.7-4 14 TRB 3.7.7-5 14 TRB 3.?.8-1 15 Section 3.2 TRB 3.7.8-2 15 TRB 3.7.8-3 15 Section 3.3 Section 3.8 Section 3.4 Section 3.9 Section 3.5 TRB 3.9.1-1 14 TRB 3.9.1-2 14 TRB 3.9.1-3 14 Section 3.6 TRB 3.9.2-1 14 TRB 3.9.3-1 14 TRB 3.9.4-1 14 TRB 3.9.4-2 14 TRB 3.9.4-3 14 TRB 3.9.4-4 14 Section 4.0 Section 5.0 R.E. Ginna Nuclear Power Plant LEP-i Revision 15
Te rary LCO Requirements TR 5.1 5.0 ADMINISTRATIVECONTROLS TR 5.1 Temporary LCO Requirements This section defines additional administrative restrictions placed on systems, structures, and components (SSCs) which are addressed within the ITS in order to ensure that they remain OPERABLE with respect to their ITS requirements. These are temporary requirements only until the ITS can be modified via a license amendment request, plant modification with a subsequent bases change, etc. Table TR 5.1-1 contains the listing of temporary LCO requirements and all associated background information.
R.E. Ginna Nuclear Power Plant TR 5.1-1 Revision 14
T rary LCO Requirements TR 5.1 Table TR 5.1-1 Temporary LCO Requirements Action Required Actions Item Report ff Temporary LCO Requirement If Not Met Effective Dates 5.1.1 97-0783 a. Verify key switches for a. Declare associated 6/1 9/97 until LCO 3.5.2 is MOVs 852A and 852B ECCS train inoperable changed are in 'off position" every per LCO 3.5.2.
31 days.
5.1.2 97-0894 a. Verify NaOH System tank a. Declare NaOH system 6/1 9/97 until LCO 3.6.6 is NaOH solution inoperable per LCO changed.
concentration is 6 35% by 3.6.6.
weight every 184 days.
R.E. Ginna Nuclear Power Plant TR 5.1-2 Revision 14
k Distri75.txt Distribution Sheet g~ 14~%<a'g Priority: Normal From: E-RIDS1 Action Recipients: Copies:
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Item: ADAMS Document Library: ML ADAMS"HQNTAD01 ID: 011440392:1
Subject:
Revision 20 to Nuclear Emergency Response Plan R.E. Ginna Nuclear Powe r Plant.
Body:
ADAMS DISTRIBUTION NOTIFICATION.
Electronic Recipients can RIGHT CLICK and OPEN the first Attachment to View the Document in ADAMS. The Document may also be viewed by searching f or Accession Number ML011440392.
A045 OR Submittal: Emergency Preparedness Plans, Implementing Proced ures, Correspondence Docket: 05000244 Page 1
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'l! IVIIIIIIIIII 1066 Igj)g IiIIIII ROCHESTER GAS AND ELECTRIC CORPORATION o 89 EAST AVENUE, ROCHESTER, N.Y. 14649-0001 TELEPHONE AREA coDE 7I6 546-2700 May 18, 2001 U.S. Nuclear Regulatory Commission:'
Document Control Desk Washington, DC 20555 Attn: Mr. Guy S. Vissing (Mail Stop 14D11)
Project Directorate I-1
Subject:
Revision to Nuclear Emergency Response Plan R.E. Ginna Nuclear Power Plant Docket No. 50-244 Gentlemen:
In accordance with 10 CFR 50.4(b)(5), en'closed are corrections to Revision 20 of the Ginna Station Nuclear Emergency Response Plan (NERP).
We have determined, per the requirements of 10 CFR 50.54(q), that the changes do not decrease the effectiveness of the Nuclear Emergency Response Plan.
Very truly yours, Peter S. Pol i Corporate Nuc ar Emergency Planner Enclosures xc: USNRC Region 1 (2 copies of letter'and 2 copies of the plan)
Resident Inspector, Ginna Station (1 copy of letter and 1 copy of the plan)
RGSE Nuclear Safety and Licensing (1 copy of letter)
Dr. Robert C. Mecredy (2 copies of letter only)
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OROCHESTER GAS S ELECT4k CORPORATION INTEROFFICE CORRESPONDENCE Date: May 17, 2001
Subject:
Changes to Rev. 20 of the Nuclear Emergency Response Plan (NERP)
From: Peter PolfleIt (Ext. 6772)
To: .Controlled Copyholders Attached please find the corrected pages for Revision 20 of the Nuclear Emergency Response Plan. Please remove/add the pages as directed below:
Remove Re lace With 25 25 27 27 29 29 90 90 91 91 92 92 98 98 99 99 100 100
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Table of Contents continued
~Pa eNo.
6.0 Emergency Facilities 61 6.1 Emergency Response Facilities (RG8 E) 61 6.2 Communications System 69 6.3 Assessment Facilities 72 6.4 Protective Facilities 88 6.5 First Aid and Medical Facilities 91 7.0 Maintaining Nuclear Emergency Preparedness 92 7.1 Training and Drills 92 7.2 Review and Revision of the Plan and Procedures 98 7.3 Emergency Equipment and Supplies 98 7.4 Auditing 99 8.0 Recovery 100 8.1 Recovery Actions 100
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The company's Emergency Support Organization is charged with the responsibility of bringing together a cohesive company managerrient and technical team. This organization will be calling upon the maximum resources available within the company and the entire nuclear industiy for the purpose of (1) assuring the safe shutdown and recovery of Ginna Station following an accident condition and, (2) minimizing the impact of the situation on the health and safety of the public.
The Ginna Station short-term responding organization is made up of site personnel, those on-shift and those immediately available from the plant staff complement. The Emergency Support Organization will be available to the Ginna Station Plant Manager for implementation of long-term recovery operations.
The company's Emergency Support Organization is established under the leadership of a single individual called the EOF/Recovery Manager. The EOF/Recovery Manager is supported by various technical and advisory disciplines including Engineering, Facilities and Personnel, Nuclear Operations, Public Communications and Energy Delivery.
4.2.4 Im lementation of the Offsite Emer enc Or anization:
Ginna Station procedures provide that the EOF/Recovery Manager (or qualified alternate) is notified and provided with details concerning the emergency, the emergency classification, station status, and immediate corporate assistance, if any, which may be required. The EOF/Recovery Manager has the authority to activate the Emergency Operations Facility.
The Emergency Support Organization (Figure 4.2C) will be activated under the Alert, Site Area Emergency and General Emergency categories although it may be activated under other categories. Upon activation of the Emergency Support Organization, the EOF is established in the basement of the leased office building at 49 East Avenue, adjacent of the main corporate office building. A Joint Emergency News Center (JENC) will also be activated in the main corporate office building, The JENC will be used to coordinate all news releases and press conferences with the appropriate Federal, State and local authorities.
Personnel who have responsibilities in the EOF will be notified of an incident and the need for them to report to the EOF according to an approved procedure. Each individual assigned to the EOF will have a designated alternate who will be notified as necessary.
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Desigri ahd construction support with the responsibility of coordinating the activities of the company, AIE, NSSS supplier and construction forces on proposed station modifications or other design and construction support required for response and recovery.
- 4. Advisory support functions with advisory support consisting of, senior representatives of the NSSS supplier and special consultants as necessary.
- 5. Public Communications'taff with the responsibility of providing administration, logistics, communications, and personnel support for the response and recovery operations.
- 6. Administration and logistics with the responsibility of providing administration, logistics, communications, and personnel support for the response and recovery operations.
The EOF will be organized as shown on the attached Emergency Support Organization Chart, Figure 4.2C. The positions will be filled by trained individuals as listed in Section 4.2.6.
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4.2.6 The following positions are the principal managers and coordinators which make up the team in the EOF and Joint Emergency News Center in response to a nuclear emergency at Ginna Station. The responsibilities for each position are given in EPIP 4-7 and EPIP 5-7.
4.2.6.1 EOF/Recovery Manager assumes overall command and control of the recovery operation of the Ginna facility.
4.2.6.2 Energy Distribution Liaison provides advisor technical support, supplementary and complementary to onsite personnel using consultants and in-house technical experts, as necessary. Directs electric system operations to meet the emergency.
4.2.6.3 Nuclear Operations Manager assists EOF/Recovery Manager in coordinating activities of the offsite organization to support site activities.
4.2.6.4 Engineering Manager coordinates the design and construction activities of the utility, A/E, NSSS Supplier, construction forces, and outside vendors.
4.2.6.5 Facilities and Personnel Manager provides administrative, logistic, communications, and personnel support for the recovery operation.
4.2.6.6 News Center Manager directs implementation of procedures governing JENC facility activation, operations and public information functions.
4.2.6.7 Corporate Spokesperson advises the media and the community through the Public Communications function of current and potential plant conditions.
4.2.6.8 Offsite Agency Liaison advises County, State, and Federal Agency Representatives in the EOF or outside of conditions related to the incident.
'4 4.2.6.9 EOF Dose Assessment Manager advises EOF/Recovery Manager of projected doses and Protective Action Recommendations (PARS) to limit exposure to the public in the affected Emergency Response Planning areas.
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6.4.5 Contamination Control Provisions:
The plant site is divided into two categories, the Clean Area and the Radiologically Controlled Area. Entry to and exit from the Radiologlcally Controlled Area is normally through the designated Access Control Point. Any area in which radioactive materials and radiation are present shall be surveyed, classified, roped and conspicuously posted with the appropriate radiation caution sign. These then become Radiologically Controlled Areas and proper access is provided and controlled. Plant procedures provide the radiation or contamination levels at which an area is declared a Radiologically Controlled Area or removed from radiation control status.
The general arrangement of the service facilities is designed to provide adequate personnel decontamination and change areas. The clean locker room is used to store items of personal clothing not required or allowed in the Controlled Area.
The Access Control Point is employed as a protective clothing change area. A supply of clean protective clothing for personnel is maintained in this area, and there is provision for collection of used protective clothing.
All personnel will survey themselves on leaving the Controlled Area using equipment provided at the Access Control Point. A decontamination shower and washroom are located adjacent to the Access Control Point.
Personnel decontamination kits with instructions posted for their use are available in the dispensary described in Section 6.5, First Aid and Medical Facilities.
In the event of a site evacuation, provisions for decontamination are available at the Survey Center.
A 1000-gallon holding tank is available to contain decontamination water from a sink and shower located in the Survey Center. Decontamination water will be sampled prior to transfer, treatment or disposal.
6.4.6 Protective E ui ment and Su lies:
Personnel entering the Controlled Area may be required to wear protective clothing.
The nature of the work to be done governs the selection of protective clothing to be worn by individuals. The protective apparel available are shoe covers, head covers, gloves and coveralls. Additional items of specialized apparel such as plasticsuits, face shields, and respirators are available for operations involving high levels of contamination. In all cases, Radiation Protection personnel shall evaluate the radiological conditions and specify the required items of protective clothing to be worn.
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Respiratory protective devices are required wherever an airborne radiation area exists or is expected. In such cases, Radiation Protection personnel monitor the airborne concentrations and specify the necessary protective devices according to concentration and type of airborne contaminants present.
Available respiratory devices include full face air purifying respirators (filter type both negative and pressured powered air purifying units). Air line supplied respirators of pressure demand type are used as well as constant flow hoods. Self-Contained Breathing Apparatus, using full face masks and pressure demand regulators are also available.
For use in an emergency, equipment and supplies are located in the Control Room, Technical Support Center, Respiratory Protection Facility and the Survey Center.
Equipment categories are given in Appendix D.
6.4.7 Emer enc Vehicles:
In the event it becomes necessary to make use of automotive equipment, a number of vehicles will be available. These include company-owned vehicles and personal vehicles. A small delivery truck and a small work truck are assigned to the Station.
The ground maintenance garage, nearby, is assigned a 4-wheel drive truck. Use of personal vehicles is allowed by the company policy regarding paid mileage for company use. Lastly, a large and diverse fleet of vehicles is available from the Rochester Gas and Electric Physical Services Department.
6.5 First Aid and Medical Facilities First aid and medical provisions include both onsite and offsite facilities. The latter are described in Section 4.3, Augmentation of the Emergency Organization. A dispensary onsite contains sinks, a toilet, a bed, a stretcher, and miscellaneous first aid equipment and supplies. Decontamination kits can be obtained from the main Radiation Protection group. Personnel decontamination kits and bioassay collection kits are available at Rochester General Hospital and Newark Wayne Community Hospital.
Auxiliary Operators and Security Officers are trained in first aid procedures using Red Cross Multi-Media or an equivalent program. An administrative procedure establishes a First Aid Team and the actions to be followed in the event of illness or injury at Ginna Station.
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0 7.0 Maintainin NuclearEmer enc Pre aredness:
Formalized training program(s) have been established to ensure that all personnel who actively participate in the Nuclear Emergency Response Plan (NERP) maintain their familiarity with the plan and their required response. A radiation emergency exercise shall be conducted at least annually, with emphasis placed upon orderly implementation of the emergency plan.
Personnel trained for onsite response to a radiation emergency are part of the regular plant staff and are trained to specific responsibilities within the emergency organization. Training is documented by the Department Manager, Nuclear Training, and Corporate Nuclear. Emergency Planner. Any emergency plan work by consultants will be under the control of, and reviewed by, the Corporate Nuclear Emergency Planner.
Exercises shall be evaluated by the Corporate Nuclear Emergency Planner and reviewed by the Plant Operations Review Committee (PORC), thereby assuring the effectiveness of the plan throughout the lifetime of the R.E. Ginna facility.
7.1 Training classes on the emergency plan shall be conducted annually (+I- 3 months) for all RG8E personnel who may actively participate in the radiation emergency plan.
Details of the training programs are established in TR C.22 (Nuclear Emergency Response Plan Training Program). Training will include a demonstration of their ability to perform the functions to which they may be assigned by participating in a Drill or Exercise at least once every two years. During drills, on-the-spot corrections of erroneous performance may be made, followed by a critique or corrective action.
Efforts will also be made to vary the timing of the exercises such that back-shifts will be involved once every six years (i.e. between 6 p.m. and 4 a.m.). Attempts will be made to have some drills unannounced.
Specialized training will be provided for:
- 1. Technical Support Center assignees
- 2. Operation Support Center assignees
- 3. First Aid Teams
- 4. Survey Teams
- 5. Emergency Operations Facility personnel 92
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72 Annual Review and Revision of the Plan and Procedures; Annual review and revision of the Nuclear Emergency Response Plan (EPIP 5-6) will occur following the annual QA audit. Revisions to the Plan are subject to approval by PORC and review by the Nuclear Safety and Audit Review Board (NSARB).
Revisions to the Plan and Emergency Plan Implementing Procedures (EPIP) may be the result of drills, exercises, training or routine surveillance. The Plan and Emergency Plan Implementing Procedures are reviewed by the EPIP Committee. EPIP Procedure changes are controlled using the guidance in the administrative procedures A-205.2 and the IP-PRO series.
Emergency procedure changes are controlled so that only current copies are available for use. Revised procedures are distributed to a list of controlled copy holders with receipt verification. Shift Operators and licensed staff are made aware of revisions during regularly scheduled training coordinated by the Nuclear Training Department.
Emergency telephone numbers are kept up-to-date through quarterly review and distribution of revisions.
7.3 Emer enc E ui mentandSu lies:
The operational readiness of all items of emergency equipment and supplies will be assured through monthly inspections of emergency equipment. The implementing procedure includes inspecting and testing of equipment stored in the Control Room, Survey Center, Technical Support Center, Radiation Protection office, JENC and EOF.
Also included is the procedure for testing the operability of the equipment.
Necessary transportation for offsite surveys will be a personal car supplied by one member of the team. Since most employees commute by private car because of the remoteness of the facility, no lack of vehicles is anticipated. Company policy provides for mileage reimbursement.
7.3.1 Siren Tests and o erabili The Ginna Emergency Sirens shall be activated at intervals not to exceed one year (+ or - Three calendar Months). This test will be considered successful if no more than 10% (10 of 96) of the sirens fail to operate properly. Any time it is found that more than 25% (24 of 96) of the sirens are inoperable for more than One Hour, Then a One-Hour notification willbe made to the NRC. Using procedure 0-9.3 Attachment 3 (reference: 10CFR50.72(b)(1)(v)).
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7.4 ~Auditin:
The Nuclear Emergency Response Plan, its implementing procedures, equipment, training and interface with State and county authorities is audited at a 12 Month frequency 3 months and audit results are reviewed by the NSARB. Follow-up implementation tasks are assigned through the use of the ACTION Reports. Audit documents are retained for 5 years.
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t Recoverec After the initial emergency response actions are concluded (i.e., the plant is in cold shutdown and under control), a decision to begin the recovery phase will be initiated. A number of considerations will enter into the decision to begin the recovery phase and dismantle the Emergency Response Organization. The decision to enter the recovery phase will be made by the EOF/Recovery Manager in consultation with his support managers, the Plant Operation Review Committee, and the NSARB and onsite personnel. EPIP 3-4, "Emergency Termination and Recovery" will be used by the organization to transition from a response organization to a recovery organization.
The decision to enter the recovery phase should be based upon a comprehensive review of station parameters and conditions. These should include, but are not limited to the following:
- 1. The initial emergency response actions are concluded (i.e., the station is in cold shutdown and under control).
- 2. Station parameters of operation no longer indicate a potential or actual emergency exists.
- 3. The reactor shutdown conditions are stable.
- 4. The reactor containment building integrity is intact.
- 5. The release of radioactivity from the station is controllable and no longer exceeds permissible levels, and no danger to the general public from the above source(s) is credible.
- 6. Radioactivity wastesystems and decontamination facilities are operable to the extent needed.
- 7. A reactor heat sink is available and operating.
- 8. The integrity of power supplies and electrical equipment needed for the station to be capable of sustaining itself in a long teim shutdown condition is intact.
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~ ~ ~ ~ I' Distribution Sheet Distri-l.txt g~ /tl gyyi5<
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Item: ADAMS Document Library: ML ADAMS HQNTAD01 ID: 011280194:1
Subject:
R.E. Ginna Revision to Emergency Plan Implementing Procedures.
Body:
ADAMS DISTRIBUTION NOTIFICATION.
Electronic Recipients can RIGHT CLICK and OPEN the first Attachment to View the Document in ADAMS. The Document may also be viewed by searching for Accession Number ML011280194.
A045 - OR Submittal: Emergency Preparedness Plans, Implementing Procedures, Corre Page 1
Distri-l.txt spondence Docket: 05000244 Page 2
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ROCHESTER GAS AND ELECTRIC CORPORATION o 89 EAST AVENUE, ROCHESTER, N.K 14649-0001 TCLCPHONC ARcA coDc 7le 546-2700 April 30, 2001 U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Attn: Mr. Guy S. Vissing (Mail Stop 14D11)
Project Directorate I-1
Subject:
Revision to Emergency Plan Implementing Procedures R.E. Ginna Nuclear Power Plant Docket No. 50-244 Gentlemen:
In accordance with 10 CFR 50.4(b)(5), enclosed are revisions to Ginna Station Emergency Plan Implementing Procedures (EPIPs).
We have determined, per the requirements of 10 CFR 50.54(q), that these procedure changes do not decrease the effectiveness of our Nuclear Emergency Response Plan.
Very truly yours,
\
Pete S. o erat Corporate Nucle Emergency Planner Enclosures xc: USNRC Region 1 (2 copies of letter and 2 copies of each procedure)
Resident Inspector, Ginna Station (1 copy of letter and 1 copy of each procedure)
RGSE Nuclear Safety and Licensing (1 copy of letter)
Dr. Robert C. Mecredy (2 copies of letter'only)
PSP/jtw gLoIIgkO/f'/
PROCEDURE REVISION NUMBER EPIP 2-2 10 EPIP 2-11 15 EPIP 2-12 18 EPIP 4-7 16
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Item: ADAMS Document Library: ML ADAMS"HQNTAD01 ID: 010920077:1
Subject:
Emergency Operating Procedures R.E. Ginna Nuclear Power Plant Docket No. 50-244 Body:
ADAMS DISTRIBUTION NOTIFICATION.
Electronic Recipients can RIGHT CLICK and OPEN the first Attachment to View the Document in ADAMS. The Document may also be viewed by searching for Accession Number ML010920077.
Page 1
A002 - OR Submittal:Inadequate Core Cooling (Item II.F.2) GL 82-28 Docket: 05000244 Page 2
AND A Subsidiary of RGS Energy Group, Inc.
ROCHESTER GAS AND ELECIRIC CORPORATION ~ 89 EAST AVENUE, ROCHESTER, N.Y. Id649CXNI ~ 716.771-3250 www.rge.corn JOSEPH A. WIDAY VICE PRESIDENT E PlANT MANAGER GINNA STATION March 26, 2001 U.S. Nuclear Regulatory Commission Document Control Desk Attn: Guy S. Vissing Project Directorate I Washington, D.C. 20555
Subject:
Emergency Operating Procedures R.E. Ginna Nuclear Power Plant Docket No. 50-244
Dear Mr. Vissing:
As requested, enclosed are Ginna Station Emergency Operating Procedures.
Very truly yours, 6'.
J s p A.Wida JAW/jdw xc: U.S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406-1415 Ginna USNRC Senior Resident Inspector Enclosure(s):
AP Index AP-PRZR.1, Rev 12
4 REPORT NO. 01 GINNA NUCLEAR POWER PLANT 03/15/01 PAGE: 1 REPORT: NPSP0200 PROCEDURES INDEX DOC TYPE: PRAP ABNORMAL PROCEDURE PARAMETERS: DOC TYPES - PRAP STATUS: EF QU 5 YEARS ONLY:
PROCEDURE EFFECT LAST NEXT NEWER PROCEDURE TITLE REV DATE REVIEW REVIEW ST AP-CCW. 1 L GE INTO THE COMPONENT COOLING LOOP 014 01/09/01 05/01/98 05/01/03 EF AP-CCW. 2 LOSS OF DURING POWER OPERATION 014 05/18/00 08/17/99 08/17/04 EF AP-CCW. 3 LOSS OP CCW P SHUTDOWN 012 05/18/00 08/17/99 08/17/04 AP-CR. 1 CONTROL ROOM INACCESSI ITY 016 01/11/00 01/11/00 01/11 5 EF AP-CVCS. 1 CVCS LEAK 012 05/01/98 05/01/98 /01/03 EF AP-CVCS. 3 LOSS OF ALL CHARGING FLOW 002 02/11/00 02/ 99 02/26/04 EF AP-CW. 1 LOSS OF A CIRC WATER PIMP 010 07/16/9 05/01/98 05/01/03 EF AP-ELEC. 1 LOSS OP 12A AND/OR 12B BUSSES 020 /08/00 05/01/98 05/01/03 EF AP-ELEC. 2 SAPEGUARD BUSSES LOW VOLTAGE OR SYSTEM LOW FREQUENCY 009 03/22/99 03/22/99 03/22/04 EF AP-ELEC.3 LOSS OF 12A AND/OR 12B TRANSFORMER (BELOW 350 F) 008 09/08/00 05/01/98 05/01/03 EF AP-ELEC. 14/16 LOSS OF SAFEGUARDS BUS 14/16 03 03/15/01 06/09/97 06/09/02 EP AP-ELEC.17/18 LOSS OF SAPEGUARDS BUS 17/18 002 0/18/99 06/09/97 06/09/02 EP AP-FW. 1 PARTIAL OR COMPLETE LOSS OF MAIN FEEDWATE 012 02/11 0 02/27/98 02/27/03 EP AP-IA.1 LOSS OF INSTR~i AIR 017 12/02/99 0 01/98 05/01/03 EF AP- PRZR. 1 ABNORMAL PRESSURIZER PRESS 011 12/02/99 12/02/ 12/02/04 EF AP-RCC.1 CONTINUOUS CONTROL RO ITHDRAWAL/INSERTION 006 02/24/96 05/14/98 05 /03 EF AP-RCC.2 RCC/RPI MAL CTION 008 11/16/98 02/06/97 02/06/02 F AP-RCC.3 DROP ROD RECOVERY 004 11/16/98 02/27/98 02/27/03 EF AP-RCP. 1 RCP SEAL MALFUNCTION 013 06/09/00 05/01/98 05/01/03 EF AP .1 REACTOR COOLANT LEAK 015 09/08/00 05/01/98 05/01/03 EF AP-RCS.2 LOSS OF REACTOR COOLANT FLOW 010 12/14/98 05/01/98 05/01/03 EF AP-RCS.3 HIGH REACTOR COOLANT ACTIVITY 007 08/05/97 08/05/97 08/05/02 EF AP-RCS.4 SHUTDOWN LOCA 011 12/02/99 05/01/98 05/01/03 EP AP-RHR.1 LOSS OF RHR 015 02/08/01 05/01/98 05/01/03 EF
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REPORT NO. 01 GINNA NUCLEAR POWER PLANT 03/15/01 PAGE: 2 REPORT: NPSP0200 PROCEDURES INDEX DOC TYPE: PRAP ABNORMAL PROCEDURE PARAMETERS: DOC TYPES - PRAP STATUS: EF QU 5 YEARS ONLY:
PROCEDURE EFFECT LAST NEXT NUMBER PROCEDURE TITLE DATE REVIEW REVIEW ST AP-RHR. 2 LOSS OF RHR WHILE OPERATING AT RCS REDUCED INVENTORY CONDITIONS 009 10/13/00 03/31/00 03/31/05 EF AP-SG. 1 STEAM GENERATOR TUBE LEAK 000 09/08/00 09/08/00 09/08/05 EF AP-SW. 1 SERVICE WATER LEAK 015 10/18/99 06/03/98 06/03/03 EF AP-TURB. 1 TURBINE TRIP WITHOUT RX TRIP REQUIRED 010 02/12/99 10/10/97 10/10/02 EF AP- TURB. 2 TURBINE LOAD REJECTION 017 02/11/00 05/13/98 05/13/03 EF AP-TURB. 3 TURBINE VIBRATION 010 02/11/00 02/10/98 02/10/03 EF AP>> TURB. 4 LOSS OF CONDENSER VACUUM 014 05/01/98 05/01/98 05/01/03 EP AP-TURB. 5 RAPID LOAD REDUCTION 005 06/09/00 06/09/00 06/09/05 EF TOTAL FOR PRAP 32
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