ML032720619
| ML032720619 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 04/01/2003 |
| From: | Alan Blamey NRC/RGN-I/DRS/OSB |
| To: | Conte R NRC/RGN-I/DRS/OSB |
| Conte R | |
| References | |
| 50-354/03-302 50-354/03-302 | |
| Download: ML032720619 (180) | |
Text
84-11-83 1 4 : 4 5 ID=
QUESTION I O Answer choice C states "Raise level to +80 inches usjng natural circulation for heat removal."
Natural circulation removes decay heat from the fuel bundles in the core to the bulk coolant. RWCU, C, RHR, D RHR, or Condensate Transfer can be used to remove decay heat from the bulk coolant to the Main Condenser, RACS, or SACS.
Abnormal procedure HC.OP-ABRPV-0009 Condition E action step E.2 states "Maintain RPV Level 2 80 inches BUT 5 90 Inches." This step is performed if Forced Circulation cannot be established using preferred RHR loops (A or 8) or Reactor Recirculation.
This step is performed when RWCU, C RHR, D RHR, or Condensate Transfer is required for Alternate Decay Heat removal. The conditions of the stem require Alternated Decay Heat removal methods.
Condition E action E.5 states "Evaluate the following systems for alternate decay heat removal:
0 RWCU (Subsequent F)
C RHR (Attachment 1)
D RHR (Attachment 2)
Condensate Transfer (Subsequent G)"
The stem does not provide core exposure history other than shutdown 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> ago.
The student could assume Beginning of Core Life, End of Life or anywhere in between.
Since the stem does not rule out RWCU or Condensate Transfer operation, RWCU can be assumed in service and can be used in conjunction with natural circulation once level has been raised to 80 - 90 inches. The heat removal means is natural circulation removing heat from the fuel bundles to the bulk reactor coolant then to RACS and the Main Condenser. Under normal operation with RWCU rejecting 69 gpm from CRD injection, with some heat removed through RACS and some removed by replacement water from CRD. Based on stem conditions, RWCU is required for Alternate Decay Heat Removal. RWCU is realigned IAW Subsequent Action step F which opens the cooling water supply valve ED-V035 full open and bypasses the Non -Regenerative Heat Exchanger.
Heat removal capability in Alternate Decay Heat Removal Mode is approximately 15 to 16 Million BTU's per hour. ( per System Engineering 8. Down). Reactor decay heat load at the Beginning of Life (80L) during the initial startup from a typical 30 day refueling outage is approximately 13 Million BTUs per hour and rises with full power operation history. If the scram was assumed from the startup from refueling outage before the reactor had any significant full power operation, the decay heat load 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> after the scram would be well within RWCU Alternate Decay Heat Removal capability. Therefore, answer choice C would also be correct.
The students are not required to know the value of BTU's per hour removal rate, or the BTU generation rated of the core at a particular time of core life. From a procedure user point of view, answer choice C is also correct when applied to Subsequent Action step E.
Recommended action is to accept answer choices A or C as correct answers.
98%
P. 02
Given the following conditions:
- The plant is in Operational Condition 4 following a forced shutdown 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> ago.
- RHR Loop A operating in Shutdown Cooling.
- The B RHR pump is Cleared & Tagged for motor replacement.
- The A RHR pump develops a high vibration and trips on overcurrent.
- HC.OP-AB.RPV-0009, Shutdown Cooling, is entered.
Which of the following will be adequate to maintain Operational Condition 4?
_ _ _ pump for
_ heat removal.
Crosstie C or Maximize RWCU bottom head drain flow.
Raise level to +80 inches using natural circulation for heat removal.
0
~ -
_ _ _ I _ _ _ _ _ _
a GSkt with Core Spray from the CST to the RPV.
a IAppl ica tion Hope Creek 02/24/2003 295021A104 nt Evolutions 3
2
- 61. Ability to operate and/or monitor the following as they apply to LOSS.. OF. SHUTDOWN COOLING 3 7 3 7
~ _ _
Crosstie C or D RHR pump for heat removal. correct - RPV-0009 subsequent action E, RHR Pumps C & D may be realigned to provide alternative decay heat removal.
Maximize RWCU bottom head drain flow. -incorrect-Maximizing bottom head drain flow does not provide heat removal adequate to maintain less than 200 degrees.
Raise level to +80 inches using natural circulation for heat removal. -Incorrect-per subsequent action E Natural Circulation does not provide heat removal, only circulation.
Inject with Core Spray from the CSTs to the RPV. -incorrect-This is not an approved method of Alternate DHR.
ABRPVSE007 (R) Explain the bases for Subsequent Actions and the information contained in the Discussion Bases Section of Shutdown
)
i Cooling.
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...~
~
~ _-__..
None Significantly Modified c
- Q61 Tuesday, March 04,2003 11:30:37 AM I
Page 15 of 58
PSEG Internal Use Only 7
SUBSEQUENT OPERATOR ACTIONS (continued)
I
~
CONDITION established using preferred RHR loops or Reactor Recirculation.
E. Forced Circulation CANNOT be
[CD-693A9 CD-1784 CD-973B9 CD-lOOA, CD-O76B, CD-065x1
- Time:
i F. RWCU is required for Alternate Decay Heat Removal.
(CIb900EJ Time:
Hope Creek ACTION 3
NOTE I**
2 E. 1 MONITOR temperatures IA W DL-0026 s.
2 E.2 MAINTAIN RPV LVL 1 80 inche 3 E.3 RPV LVL reaches 90 inches, Ti-IEN CLOSE the MSNs.
2 E.4 ENSURE T.S. cool down limits are not exceeded.
3 a RWCU (Subsequent F) [CD-900E]
P a C RHR (Attachment 1) 0 D RHR (Attachment 2) 0 CONDENSATE TMNSFER R E.6 the vessel head is removed, AND the Reactor Cavity is flooded, THEN maximize Fuel Pool Cooling:
P e ENSURE two Fuel Pool Cooling pumps 0
a ENSURE SACS flow aligned through 0 F.l ENSURE RWCU is in service. (B are in service. (EC)
BOTH Fuel Pool Cooling heat exchangers.
L 0 F.2 FULLY OPEN ED-VO35.
[3 F.3 Enecessary,
\\
THEN Bypass the Regenerative heat exchanger to maximize decay heat removal.
(BG)
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Page 11 of 45 Rev. 0
- 4.
You may bring pens, pencils, and calculators into the examination room. Use black ink to ensure legible copies; dark pencil should be used only if necessary to facilitate machine grading.
- 5.
Print your name in the blank provided on the examination over sheet and the answer sheet. You may be asked to provide the examiner with some form of positive identification.
- 6.
Mark your answers on the answer sheet provided and do not leave any question blank.
Use only the paper provided and do not write on the back side of the pages. If you are using ink and decide to change your original answer, draw a single line through the error, enter the desired answer, and initial the change.
- 7.
If you have any questions concerning the intent or the initial conditions of a question, do operation or training references, you should answer the question based on the actual Restroom trips are permitted, but only one applicant at a time will be allowed to leave.
Avoid all contact with anyone outside the examination mom to eliminate even the appearance or possibility of cheating.
When you complete the examination, assemble a package including the examination questions, examination aids, answer sheets, and scrap paper and give it to the NRC examiner or proctor. Remember to sign the statement on the examination cover sheet indicating that the work is your own and that you have neither given nor received assistance in completing the examination. The scrap paper will be disposed of immediately after the examination.
After you have turned in your examination, leave the examination area as defined by the proctor or NRC examiner. If you are found in this area while the examination is still in progress, your license may be denied or revoked.
plant.
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- 8.
- 9.
- 10.
- 11.
Do you have any questions?
PART C - GENERIC OPERATING TEST GUIDELINES (CATEGORIES A. B. AND C)
- 1.
If you are asked a question or directed to perform a task that is unclear, you should not hesitate to ask for clarification.
- 2.
The examiner will take notes throughout the test to document your performance, and sometimes the examiner may take a short break for this reason. The amount of note-NUREG-1 021, Revision 8, Supplement 1 2 of 5
4 ATTACHMENT 2 LICENSEE COMMENTS Written Question 10 Answer choice C states Raise level to +80 inches using natural circulation for heat removal.
Natural circulation removes decay heat from the fuel bundles in the core to the bulk coolant.
RWCU, C RHR, D RHR, or Condensate Transfer can be used to remove decay heat from the bulk coolant to the Main Condenser, RACS, or SACS.
Abnormal procedure HC.OP-AB.RPV-0009 Condition E, action step E.2 states Maintain RPV level greater than or equal to 80 inches but less than or equal to 90 inches. This step is performed if forced circulation cannot be established using preferred RHR loops (A or 6) or reactor recirculation. This step is performed when RWCU, C RHR, D RHR, or condensate transfer is required for alternate decay heat removal. The conditions of the stem require alternate decay heat removal methods to be used.
Condition E, action E.5 states Evaluate the following systems for alternate decay heat removal:
RWCU (subsequent F)
C RHR (Attachment I) 0 D RHR (Attachment 2)
Condensate Transfer (Subsequent G)
The stem does not provide core exposure history other than shutdown 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> ago. The student could assume Beginning of Core Life, End of Core Life or anywhere in between.
Since the stem does not rule out RWCU or condensate transfer operation, RWCU can be assumed in service and can be used in conjunction with natural circulation once reactor level has been raised to 80 - 90 inches. The heat removal means is natural circulation removing heat from the fuel bundles to the bulk reactor coolant, then to RACS and the main condenser. Under normal operation with RWCU rejecting 69 gpm from CRD injection, with some heat removed through RACS and some removed by replacement water from CRD. Based on stem conditions, RWCU is required for alternate decay heat removal. RWCU is realigned in accordance with subsequent action step F, which opens the cooling water supply valve ED4035 full open and bypasses the non regenerative heat exchanger.
Heat removal capability in alternate decay heat removal mode is approximately 15 to 16 Million BTUs per hour. (Per System Engineering 6. Down). Reactor decay heat load at the Beginning of Life (BOL) during the initial startup from a typical 30 day refueling outage is approximately 13 Million BTUs per hour and rises with full power operation history. If the reactor automatic shutdown was assumed to occur during a startup from a refueling outage, before the reactor had any significant full power operation, the decay heat load 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> after the automatic shutdown would be well within RWCU Alternate Decay Heat Removal capability. Therefore, answer choice C would also be correct.
5 The students are not required to know the value of BTUs per hour removal rate, or the BTU generation rate of the core at a particular time of core life. From a procedure user point of view, answer choice C is also correct when applied to Subsequent Action step E.
Recommended action is to accept answer choices A or C as correct answers.
6 ATTACHMENT 3 NRC RESOLUTION OF LICENSEE COMMENTS Written Question: I O Comment: The question provides a condition in which the plant has been shutdown for 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> when residual heat removal (RHR) shut down cooling is lost. The applicant must determine which condition will be adequate to maintain the plant in operational condition 4, in accordance with HC.OP-AB.RPV-0009, "Shutdown Cooling." The correct answer was (a) to crosstie "C" or "D" RHR pump for core decay heat removal. Answer (c) Raise level to + 80 inches using natural circulation for heat removal was recommended to also be accepted as a correct answer. The basis for accepting (c) is that the reactor water clean up system would be in service with a normal line up and rejecting water to the condenser at 69 gpm following a reactor shutdown. If the plant has just started up following a 30 day refueling outage then the decay heat load would be approximately 13 million BTU per hour. The RWCU heat exchanger, in the alternate decay heat removal lineup will have a capacity of 15 to 16 million BTUs per hour.
Therefore, the plant will be able to be maintained in operational condition 4, after increasing level to + 80 inches mode of operation in accordance with the procedure.
reconfiguring the RWCU system to the alternate decay heat removal NRC Resolution: Based our review the only correct answer is (a) "crosstie "C" or "Dl RHR pump for heat removal." This is based on the stem of the question which states that 'Which of the following will be adeauate to maintain Operational Condition 4." Answer (a) is the only answer that is sufficient to have enough heat removal capacity to maintain the plant in operational condition 4. Answer (c) states raise level to +80 inches using natural circulation for heat removal. This action may initially keep the plant in operational condition 4, but will not be able to maintain the plant in operational condition 4 for long term. Raising level to + 80 inches only enhance heat transfer from the fuel to the coolant, but it will not remove the decay heat from the vessel. Another system(s) will be required together with raising level to remove the decay heat from the vessel. Therefore, (a) is the only correct answer.
ADAMS DOCUMENT COVER SHEET DOCUMENT TITLE: NRC Letter Transmitting Certificates ESTIMATED PAGE COUNT:
AVAILABILITY: PUBLICALLY AVAILABLE KEY WORD: NRR-079 DOCUMENT SENSITIVITY: Non-Sensitive SECURITY:
Rl-DRS-BS........... Access Level: Owner NRC Users............ Access Level: Viewer R l -DRS-OSB.........Access Level: Viewer DPC....................... Access Level: Owner C KB......................
.Access Leve I : Owner Accession No:
ADAMS DOCUMENT COVER SHEET ESTIMATED PAGE COUNT:.
AVAl LAB1 LI w+PwuG&:Y 99 mmtmmx-KEY WORD DOCUMENT SENSITIVITY: W SECURITY:
Rl-DRS-BS........... Access Level: Owner NRC Users............ Access Level: Viewer R I -DRS-OSB.........Access Level : Viewer NRR-DI PM -1OLB...Access Level: Viewer D P C.......................Access Level : Owner CKB....................... Access Level: Owner
@Zii
612000 Operator Licensing Interaction No.:
Report on Interaction (ROI) 03-01 Subject :
Type of Action:
Waiver:
X (deferral) Policy Interpretation:
Request for HQ Action:X From:
R. Conte, Chief Operational Safety Branch Date:
1 /23/03 To:
D. Trimble, Chief, IOHS Proposed Due Date:
2/24/03 Info.:
ADAMS PKG: ML022680541 ROI Accession # ML030580309 Backsround / Issue:
Mr. Terry Beltz is an applicant for the Hope Creek LSRO license exam in March 2003. Prior to taking the exam, Mr. Beltz will not meet the required six months of site specific (Hope Creek) experience: Specifically, PSEG Nuclear, LLC requests that all six months of the site specific experience be deferred until after the LSRO examination.
Deferral of Six Months of Site Specific Experience Until After Completion of the LSRO Exam.
RIIDRSIOSB Recommended Action / Resolution:
Region I recommends that Mr. Beltz be allowed to take the exam in March 2003, and defer the required six months of site specific experience until after the LSRO exam. The license: will be issue'cl'after successful completion of the LSRO exam, completion of the required six months of site specific (Hope Creek) experience and PSEG has certified, in writing, that he completed the six months of site specific training.
Final Action / Resolution:
The request to defer the six months on-site experience requirement is approved. Complete the licensing action per Section D.3.c of ES-501 of NUREG-1021 after Hope Creek provides documentary evidence thai tional Safety Branch IOLB CH:
Page I of 1
Exam Item Review Form 8 ITEM No' Course Name:
SAP Event #:
NOHOGNRCEX SAP Event #:
50763913 DATE EXAM ADMINISTERED:
04/07/03 LESSON PLAN NO.:
N/A NUMBER OF TRAINEES EXAMINED:
3 LSRO NRC Written Examination CORRECTIVE ACTION AVERAGE REV1 EW SCORE % CONCLUSION ADMl N I STERED BY:
A. Faulkner 3
4 5
10 33.3 G
Question acceptable 33.3 G
Question acceptable 0.0 G
Question acceptable 66.7 F
Accept A or C Question acceptable I
I 2 66.7 G
33.3 I
G See attached explanation.
Question acceDtable 25 26 15 1
66.7 I
66.7 G
Question acceptable 66.7 Q ues t io n acceD ta b le G
G 44 45 47 I Question acceptable 66.7 G
66.7 G
66.7 G
16 I
66.7 I
G I Question acceptable 40 I
66.7 I
G
~~
~~
Question acceptable 42 66.7 I
G Question acceptable 49 I
33.3 I
G Question acceptable Question acceptable Question acceptable Question acceptable 50 I
33.3 I
G Question acce ~ t a ble
REVIEW CONC LUSl ONS:
A. Insufficient training for the objective tested.
B. Objective not adequately covered in the lesson plan.
C. Poorly worded or invalid enabling objective.
D. Poorly worded or invalid test item or answer.
E. incorrect answer in the exam key.
F. More than one correct answer.
G. Question acceptable.
H. Other:
REVIEW PERFORMED BY:
CHANGE INITIATED BY:
TRAINING SUPERVISOR:
-Ad-----
DATE: +/F/&~
EXAMINEES NOTIFIED OF RESULTS BY:
DATE: A44
QUESTION IO Answer choice C states Raise level to +80 inches using natural circulation for heat remova I
.I Natural circulation removes decay heat from the fuel bundles in the core to the bulk coolant. RWCU, C, RHR, D RHR, or Condensate Transfer can be used to remove decay heat from the bulk coolant to the Main Condenser, RACS, or SACS.
Abnormal procedure HC.OP-AB.RPV-0009 Condition E action step E.2 states Maintain RPV Level 2 80 inches BUT 5 90 inches. This.step is performed if Forced Circulation cannot be established using preferred RHR loops (A or B) or Reactor Recirculation.
This step is performed when RWCU, C RHR, D RHR, or Condensate Transfer is required for Alternate Decay Heat removal. The conditions of the stem require Alternated Decay Heat removal methods.
Condition E action E.5 states Evaluate the following systems for alternate decay heat removal:
0 RWCU (Subsequent F) 0 C RHR (Attachment 1) 0 D RHR (Attachment 2) 0 Condensate Transfer (Subsequent G)
The stem does not provide core exposure history other than shutdown 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> ago.
The student could assume Beginning of Core Life, End of Life or anywhere in between.
Since the stem does not rule out RWCU or Condensate Transfer operation, RWCU can be assumed in service and can be used in conjunction with natural circulation once level has been raised to 80 - 90 inches. The heat removal means is natural circulation removing heat from the fuel bundles to the bulk reactor coolant then to RACS and the Main Condenser. Under normal operation with RWCU rejecting 69 gpm from CRD injection, with some heat removed through RACS and some removed by replacement water from CRD. Based on stem conditions, RWCU is required for Alternate Decay Heat Removal. RWCU is realigned IAW Subsequent Action step F which opens the cooling water supply valve ED-V035 full open and bypasses the Non Regenerative Heat Exchanger.
Heat removal capability in Alternate Decay Heat Removal Mode is approximately 15 to 16 Million BTUs per hour. ( per System Engineering B. Down)
The students are not required to know the value of BTUs per hour removal rate, or the BTU generation rated of the core at a particular time of core life. From a procedure user point of view, answer choice C is also correct when applied to Subsequent Action step E.
Accept answer choices A or C.
Given the following conditions:
- The plant is in Operational Condition 4 following a forced shutdown 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> ago.
- RHR Loop A operating in Shutdown Cooling.
- The 8 RHR pump is Cleared & Tagged for motor replacement.
- The A RHR pump develops a high vibration and trips on overcurrent.
- HC.OP-AB.RPV-0009, Shutdown Cooling, is entered.
Which of the following will be adequate to maintain Operational Condition 4?
Crosstie C or __
D RHR pump for heat removal.
Maximize
__ -. -..... RWCU bottom head drain flow. -..
Raise..............
level to +80 inches using
. -...... natural circulation for heat - removal.
Inject
- - __ with
-. - - Core Spray from the
a B
-1 Application F
l Hope Creek 1
02/24/20(
Emergency and Abnormal Plant Evolutions
(-1 3 -1 2
295021A104 295021 Loss of Shutdown Cooling qA-l.,- Ability to operate and/or monitor the following as they apply to LOSS OF SHUTDOWN COOLING:
~
AA1.04 Alternate heat removal methods 3.7 3.7 1-1 justification Crosstie C or D RHR pump for heat removal. correct - RPV-0009 subsequent action E, RHR Pumps C & 0 may be realigned to provide alternative decay heat removal.
Maximize RWCU bottom head drain flow. -incorrect-Maximizing bottom head drain flow does not provide heat removal adequate to maintain less than 200 degrees.
Raise level to +80 inches using natural circulation for heat removal. -Incorrect-per subsequent action E Natural Circulation does not provide heat removal, only circulation.
Inject with Core Spray from the CSTs to the RPV. -incorrect-This is not an approved method of Alternate DHR.
H C. 0 P-AB. R PV-0009
_-^_^-
ABRPV9E007
(
utdown (Material Required for Examination
..+\\:;, 1 None Question MO$lffcatioW~ethOd*
~ ; Significantly Modified Bank QID# (261332 Sig Mod Tuesday, March 25.2003 4:18:21 PM Page 15 of 58
Seating Chart LSRO Written Examination administration 4/7/2003 NTC Classroom 32/33 Front Scarpati >
Beltz Fitz-gibbon
Questions asked during: LSRO Examination administration.
0822 0830 0842 0905 1043 U
37 13 21 27 27 4/7/2003 Time Question #
kq=
Candidate Beltz Scarpati Fitzgibbon S carpati Fitzgibbon Fitzgibbon Question Is there a formula sheet provided?
Figure 3.1.5 missing from references Not sure what the question is asking.
What does isolated from injection mean? Does that mean isolated by procedure?
Is 10-5 complete in its entirety?
Are the bullets listed in order of occurrence?
Answer provided No.
Figure found in package out of sequence.
Answer to the best of your ability.
Yes, isolated by procedure.
Answer question based on conditions provided.
Yes.
ES-50 1 Post-Examination Check Sheet Form ES-501-1 (R8, SI)
Task Description I. Facility written exam comments or graded exams received and verified complete Facility written exam comments reviewed and incorporated and NRC grading completed, if necessary
- 2.
- 3.
Operating tests graded by NRC examiners
- 4.
NRC Chief examiner review of written exam and operating test grading completed
- 5.
Responsible supervisor review completed
- 6.
- 7.
- 8.
Facility notified of results
- 9.
Management (licensing official) review completed License and denial letters mailed Examination report issued (refer to NRC MC 0610)
I O.
Reference material returned after final resolution of any appeals Date Complete 4b103
ECG Section ii Pg 1 of 5 HOPE CREEK EVENT CLASSIFICATION GUIDE Glossary of Acronyms & Abbreviations Section ii AAAG AC ADS ALARA APRM ARI ARM ASAP ASM AS ATWS BKGD BKR BNE CACS CAS CCPM CEDE CDE CFR CIS CNTMT CP CPM CR CREF CRIDS CRD CSS DC DAPA CDE DEI DEMA DEP Accident Assessment Advisory Group (Delaware)
Alternating Current Automatic Depressurization System As Low As Reasonably Achievable Average Power Range Monitor Alternate Rod Insertion Area Radiation Monitor As Soon As Possible Administrative Support Manager Administrative Supervisor Anticipated Transient Without Scram
Background
Breaker (electrical circuit)
Bureau of Nuclear Engineering (NJDEPE)
Containment Atmosphere Control System Central Alarm Station Corrected Counts per Minute Committed Effective Dose Equivalent Committed Dose Equivalent Containment Isolation System Control Point Control Room Control Room Emergency Filter System Control Room Integrated Display System Control Rod Drive Core Spray System I
+
t Code of Federal Regulations c
Contain men t (Barrier)
L r
Counts Per Minute
- r 4 i t
, I L-t ' -.
c - - -
I
. 4 -
~
-c
/
Direct Current Drywell Atmosphere Post Accident (Radiation monitor)
Deep Dose Equivalent Dose Equivalent Iodine Delaware Emergency Management Agency Department of Environmental Protection (NJ)
HCGS Rev. 00
DID DLD DOE DOT DPCCIDCR DPM DRCF EACS EAL EAS EC ECCS ECG EDG ED0 EMRAD ENC ENS EOC EOF EOP EPA EPA EPC EPIP EPZ EQPT ERDS ERM ERO ES F Essx FC FFD FRVS FTS GE HCLL HCGS Direct Inward Dial (phone system)
Drywell Leak Detection Department of Energy Department of Transportation Discharge Prevention, Containment, & Countermeasures/
Discharge Cleanup & Removal Plan Disintegrations per Minute Dose Rate Conversion Factor ESF Equipment Area Cooling System Emergency Action Level Emergency Alert System (Broadcast)
Emergency Coordinator Emergency Core Cooling Systems Emergency Classification Guide Emergency Diesel Generator Emergency Duty Officer Emergency Radio (NJ)
Emergency News Center Emergency Notification System (NRC)
Emergency Operations Center (NJ & DE)
Emergency Operations Facility Emergency Operating Procedure Emergency Preparedness Advisor Environmental Protection Agency Emergency Preparedness Coordinator Emergency Plan Implementing Procedure Emergency Planning Zone Equipment Emergency Response Data System Emergency Response Manager Emergency Response Organization Engineered Safety Feature Electronic Switch System Exchange (Centrex)
Fuel Clad (Barrier)
Fitness For Duty Filtration, Recirculation, and Ventilation System Federal Telecommunications System (NRC)
General Emergency Heat Capacity Level Limit Hope Creek Generating Station ECG Section ii Pg2of5 HCGS Rev. 00
HCTL HEPA HPCI HTV HVAC HWCI HX IAW IC ICMF IDLH IRM I/S KI KV LAC LCO LDE LEL LLD LOCA LOP LPCI LPZ MCR MDA MEA MEES MET M.O.U.
MRO MSIV MSIVSS MSL NAWAS NCO NDAB NE0 NETS Heat Capacity Temperature Limit High Efficiency Particulate Absorbers High Pressure Coolant Injection Hardened Torus Vent Heating, Ventilation & Air Conditioning Hydrogen Water Chemical Injection Heat Exchanger In Accordance With Initiating Condition Initial Contact Message Form Immediately Dangerous to Life and Health Intermediate Range Monitor In Service Potassium Iodide Kilovolt Lower Alloways Creek Limiting Condition for Operation Lens Dose Equivalent Lower Explosive Limit Lowest Level Detectable Loss of Coolant Accident Loss of Offsite Power Low Pressure Coolant Injection Low Population Zone Main Control Room Minimum Detectable Amount Minimum Exclusion Area Major Equipment & Electrical Status (Form)
Meteorological Memorandum of Understanding Medical Review Officer Main Steam Isolation Valve Main Steam Isolation Valve Sealing System Main Steam Line National Attack Warning Alert System Nuclear Control Operator Nuclear Department Administration Building (TB2)
Nuclear Equipment Operator Nuclear Emergency Telecommunications System ECG Section i i Pg3 of5 HCGS Rev. 00
NFE NFPB NJSP NOAA NPV NRC NSS NSSSS NSTA NUMARC N W S OBE OCA ODCM OEM OHA OPCON OSB osc PAG PAR PASS PC PCIG PCIS PSIG RAD RAL RC RCA RCAM RCIC RCS RHR RM RMO RMS RPS RPS RPV RRCS Nuclear Fuels Engineer Normal Full Power Background New Jersey State Police National Oceanographic and Atmospheric Administration North Plant Vent Nuclear Regulatory Commission Nuclear Shift Supervisor Nuclear Steam Supply Shutoff System Nuclear Shift Technical Advisor Nuclear Management and Resources Council National Weather Service Operating Basis Earthquake Owner Controlled Area Offsite Dose Calculation Manual Office of Emergency Management (NJ)
Overhead Annunciator Operating Condition Operational Status Board (Form)
Operations Support Center Protective Action Guideline Protective Action Recommendation Post Accident Sample System Primary Containment (Barrier)
Primary Containment Instrument Gas System Primary Containment Isolation System Pounds Square Inch Gauge Radiation Reportable Action Level Reactor Coolant Radiologically Controlled Area Repair and Corrective Action Mission Reactor Core Isolation Cooling Reactor Coolant System (Barrier)
Residual Heat Removal (Containment Heat Removal)
Recovery Manager Recovery Management Organization Radiation Monitoring System Radiation Protection Supervisor Reactor Protection System Reactor Pressure Vessel Redundant Reactivity Control System ECG Section ii Pg4of5 HCGS Rev. 00
RSM RWCU SACS SAE SAM SAS SBO SCBA SCP SDE SDM SLC SJAE SNM SNSS sos SPDS SPV SRM SRPT SRV SSCL SSE ssws SSNM TAF TDR TED TIP TLV T/S TSC TSS TSTL TSTM UE UFSAR UHS USCG VDC WB Radiological Support Manager Reactor Water Cleanup (System)
Safety Auxiliaries Cooling System Site Area Emergency Severe Accident Managemen t Secondary Alarm Station (Security)
Station Blackout Self Contained Breathing Apparatus Security Contingency Procedure Shallow Dose Equivalent Shutdown Margin Standby Liquid Control Steam Jet Air Ejector Special Nuclear Material Senior Nuclear Shift Supervisor Systems Operations Supervisor (Security)
Safety Parameter Display System South Plant Vent Source Range Monitor Shift Radiation Protection Technician Safety Relief Valve Station Status Checklist Safe Shutdown Earthquake Station Service Water System Strategic Special Nuclear Material Top of Active Fuel Technical Document Room Total Effective Dose Equivalent Traversing Incore Probe Threshold Limit Value Technical Specifications Technical Support Center Technical Support Supervisor Technical Support Team Leader Technical Support Team Member Unusual Event Updated Final Safety Analysis Report Ultimate Heat Sink United States Coast Guard Volts Direct Current Whole Body ECG Section ii P g 5 o f 5 HCGS Rev. 00
I
..~
I a
Initiating Condition OQCON EAL # -
I I
M I
E l
E l
G I
N I
Y I
I I
A I
T I
I 0 1 N
I L
I V
I L
I s
- - I I
R I
E t
C I
C I
I I
I I
E t
Action Required
[
1,2,3,4,5 1
1.I.I
.a IF Reactor Coolant Sample Activity
> 4 pCi/gm Dose Equivalent 1-131 I
I 1.0 Fuel Clad Challenge 1.1 RCS Activity
_ _ c _ _ _
Fuel Clad Degradation
...... ~-
1,2,3,4
)
...../
1.I.l
.b IF Valid Offgas Pretreatment Radiation Monitor (9RX62 1 / 9RX622)
High Alarm Condition (1
2.2E+04 mRem/hr)
. I----
(
12,394 r
1.1.1.c IF -
Valid Main Steam Line Radiation Monitor High High Alarm Condition
( 2 3 times Normal Full Power Background) 1 NOTE:
Refer to Section 3.0, Fission Product Barrier Table prior to Event Classification
__ ~-
Refer to Attachment 1 UNUSUAL EVENT
(
1,2,3 1.1.2
-1 ANY SRV is determined AND I
to be Stuck Open 1
THEN I
I Refer to Attachment 2 ALERT
I 2.0 RCS Challenge 2.1 RCS Leakage HCGS EC(;
Rev. 00 Page I of I Initiating Condition RCS Leakage
~
('
1, 2, 3 OPCON EAL #
1 I
E l
E l
G I
N I
Y I
I I
A I
T I
0 1 I
I I
M I
I R
I I
c I
C I
I I
N I
L I
V I
L I
S I
. I 2.1.l
.d IF 2.1.l
.c IF Reactor Coolant System Identified Leakage
> 25 gpm averaged over any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period - -
2.1.1.b IF 2.1.1.a IF I
Reactor Coolant System Pressure Boundary Leakage
> 10 gpm (Using IO minute average)
Successful Isolation of a Reactor Recirc Pump Dual Seal Failure within 10 niiiiutes of recognition Reactor Coolant System Unidentified Leakage
> 10 gpm (Using 10 minute average)
I I
I
-.-r -.
.I THEN NOTE:
Refer to Section 3.0, Fission Product Barrier Table prior to Event Classification Action Required Refer to Attachment I UNUSUAL EVENT
1.;
1 have occurred which, in the judgment of the Emergency Coordinator, indicate a I
I'oteeti;il Degradation N
I of Plant Safety c
I 4.0 E C Discretion 4.1 Emergency Coordinator Discretion tiCGS ECO Rev. 00 Page I of I Increased monitoring of Safety Functions is warranted w
/' Oilier Conditions Exist Which.*
In the Judgment of the Emergeiicy Coordinator Warrant Dcxlaration ol'
,-- - '~-------------..
,' Other Conditions Esist Which hi the Judgment of the Emergency Warrant Declaration v
~ -...
Other Conditions Exist Which bi the Judgment oftlie 13mrrgency Coordinator Warrant other Coiiditions Exist Wliicli le tlir Jiidgment ol'the I;niergeiicy Coordinator Warrant I )ecliiriitioii ol'iiii Ilnusual Event. '
/
Initiating I
Condition
\\.
a~eiieral Emergency OPCON Ehl, #
I I
kl I
I R
l I
E l
E l
E l
G I
I All 1
4.1.l I
I A
I T
I I
0 1
C I
I I
N I
L I
V I
L I
S I
I I
I
-1 Action Required I
i THEN i
l
&laration ol' an AM a Site Area Enicrgency
/
A' 7
(.fl All
\\
i All have occurred which, in the judgment of the Emergency Coordinator, indicate an Actual or likely major failure of plant functions needed for protection of the public THEN 4.1.4 IF Events are in progress or have occurred which, in the judgment of the Emersency Coordinator, indicate an Actual or imminent substantial core degradation with the potential for loss of containment THEN I
i i
\\
Refcr to Attachment 4 GENERAL EMERGENCY Refer lo Attachment 3 SITE AREA EMERGENCY Refer to Attachment 1 UNUSUAL EVENT Refer to Attachment 2 ALERT 1
1.
Initiating Condition OPCON EAL ##
I M
1 An Automatic I
1 Reactor Scram R
I I
Condition exists E
l 5.0 Failure to Scram 5.1 ATWS
/ Failure of the Reactor Protection..
System (RPS) to Successfully Complete a Reactor Scram (Automatic and Manual)
Failure of the Reactor Protection System (RPS) to Successfully Complete a Reactor Scram I
ANY Manually Initiated Reactor Scram An Automatic Reactor Scram (RPS) IS NOT successful N
I Y
I I
I A
I T
I I
0 1
c I
C I
I N
I L
I V
I L
I S
I I
I I
I Action Required Control Room IS NOT successful (RPS) from the
[AND -
1
' I AND from the Control Room (RPS gnJ ARI) DID NOT REDUCE and MAINTAIN Reactor Power to 5 4%
J
- - I -
Refer to Attachment 2 ALERT
\\\\
and Reactor Power is above 4%
7-1,2-)
5.1.3 -
1
[
Reactor Scram attempts r'
IICGS ECG Rev. 00 Page I of 1
~
Failure of the Reactor Protection System (RPS)
'.\\
to Successfiilly Complete a Reactor Scram (Automatic and Manual) and there is indication of an
\\. -
Estreine Challenge to the Ability to Cool the Core
~
(---c---)
...~...
5.1.4 EITHER one of the following:
Reactor Water Level CANNOT BE MAINTAINED > -190" The combination of Suppression Pool Temperature and RPV Pressure CANNOTBE MAINTAINED below the HCTL Curve Refer to Attachment 4 GENERAL EMERGENCY
--I Refer to Attachment 3 SITE AREA EMERGENCY I-
6.0 Radiological ReleasedOccurrences 6.1 Gaseous Effluent Release tIC0S ECCi Rev. 00 Page 1 of 4
~
~
Any Un1)lanned Kclease of Gaseous Radioactivity lo Tcchnical Specifications for 60 riiinulcs or longer Initiating Condition Any Unplanned Rclease of Gaseous Radioactivity to the Environment that Esceeds 2 tinics the IOCFR20. Appcndis B linlrts for 60 niinutcs or longcr
( E~rvlronnient Escecds Times 'lie
'\\_.
OPCON EAL #
I I
E l
E l
G I
N I
Y I
I I
A I
T I
I 0 1 N
I L
I M
I I
R I
I I
C I
I I
I I
5 7 '
I L
I s
t
.I Action Required
(
All-
)
6.1.I
.a Dose IF issessment Dosc Asscssiiicnl ildiciitcs EITHER otic of tlic following ill Illc MEA or beyond
- IS ciilculi~tcd on IllC SSCL:
TEDE -I-D;iy DOSC
> 2.OE-01 mRem Thyroid-CDE Dosc
> 6.8E-01 mRem bascd on Plant Vcnt clllucni saniplc analysis and NOT on a dcbult Noblc Gas lo lodinc Riitio 6.1.1.b Field Measured IF
'os? Ra_!!L -
Dose Rak mcasurcd at tlrc Prolcckd Arca Boundav or bcyond EXCEEDS
.OS mRem/hr above nornial background All
)
6.1.l.c Sample IF
& ! e 5 Gascous cllltIcnt rclcasc saniplc analysis for ANY onc of tlic following indicatcs ii concentration o t FRVS:
2 1.13E-03 pCi/cc Total Noble Cas 3 2.7 1 E 4 7 pCi1c.c I-I3 I NPV:
2 2.43E-04 pCi/cc Total Nohle Cas SPV:
? 2.27E-OS pCi/cc Total Noble Gas 2 5.44E-O9 pCi/cc I-I3 I 5.81 E-08 pCi/cc 1-131
(
All 6.1.I
.d Alarm IF ndica!!?!?
Valid High Alarni rcccivcd from ANY otic of ilic following Plant Efflucnt RMS Ch;inncls:
FRVS Noble Cas (Grid 113; 9RX6KO)
NPV Noble Cas (Grid 1/3; 9RX590)
NPV Iodine (Grid 3; 9RX601)
SPV Nohle Cas (Grid 113; 9RXSXO)
HTV Nuhle Cas (Grid 3; 9RX5 16)
I AND Total Plant Vcnl rclcasc ratc EXCEEDS EITHER onc of tlic following liniils:
4.8OEi-03 pCi/sec Total Noble Cas 1.15E+00 pCi/scc I-l31(WV & SI'VOM,Y)
~ -.
1 Dose Asscssmcnt results NOT availablc I
-r---
6.0 Radiologica I ReleasedOccu rrences 6.1 Gaseous Effluent Release
.~
~
~~
Any Unplanned Release of Gaseous Radioactivity to the Environtnent that Escccds 200 times the IOCFR20, Appendix B limits for 30 niinutes or longer h
y Unplanned Rclease of Gaseous Radioactivity t&
( Enviroiiiiicnt that Exceeds 200 Times thc Radiological Tcchiiical
.... Spccifications
-. for..-.
I5. ___.
tniiiutes or longer Initiating Condition
.~
~
L..
OPCON EAL #
I I
M I
I H
I E
l E
l G
I N
I Y
I A
I T
I 0
1 I
I I
I c
I C
I I
I I
I N
l L
I V
I L
I s
j Action Required 6.1.2.a Dose IF
!s_sessment Dosc Asscssnicnt OW of tlic following at tlic MEA or bcyond 21s c;rlcttlatcd on I h C SSCL:
iridicillcs EITHER TEDE.I-DiIv DoSC
> 2.0E+01 mRem Thyroid-CDE Dosc
(,.8E+01 mRem c m w t s;rmple analysis aiid w o n a dcbult Noblc Gas to Iodine Ratio bilscd 011 PliltIt VCtIt
(
All )
6.1.2.b kERate- __ -
Dosc Rate medsurcd at tlic Protected Arca Boundary or beyond EXCEEDS 5 mRem1hr Field Measured IF I.
1 Release is ongoing for 2 15 minutes I
7-THEN
(/-Ai- )
6.1.2.c Sample IF L!!?!Ek.
Gascous cmucnt rclcase samplc analysis for one of tlic following indicatcs a concentration oT:
FRVS:
2 1.13E-01 pCi1cc Total Nohle Gas
> 2.71E-OS pCi1cc 1-131 NPV:
2 2.13E-02 pCilcc Total Nohlc Cas 25.81E-06 pCi1cc 1-131 SPV:
2 2.27E-03 pCi1cc Total Noble Gas 2 5.UE-07 pCi1cc 1-131 I
~,,-... A ~ - -.
1 6.1.2.d Alarm IF dications...-
Valid High Alarm rcccivcd froin ANY oiic or tlic following Plant Emucnt RMS Chaniicls:
FRVS Nohle Cas (Grid 1/3; 9RXOXO)
NPV Nohle Gas (Grid 113; 9RXS!Xr)
NPV Iodine (Grid 3; 9RX60I)
SPV Noble Gas (Grid 1/3; 9RXSXO)
HTV Nohle Cas (Grid 3; 9RX5 16) 1 AND Total Plant Vent release ralc EXCEEDS ElTHERonc of the following lioiirs:
J.80E+05 pCi/sec Total Noble Gas I
1.1SE+O2 pCi/sec 1-13 I (NPV & SPV ONLY)
-I--- -
.J I
Dose Assessment results NOT availabic LRelease is ongoing for 2 30 minutes I
. __ - _- -- 7 AND 7-
A THEN t
Refer to Attachment 2 ALERT
Initiating Condition OPCON EAI, ##
E M
E I<
G E
N c
Y A
C I
1 0
N L
E V
E L
S I
I I
I I
I I
I I
I I
I I
I I
I I
I I
I I
I I
I I
I I
I I
I I
I I
I Act ion Required 6.0 Radiological ReleasedOccurrences 6.1 Gaseous Effluent Release ticcis ECG Kcv. 00 Page 3 of4 I
~
Boundary Dose Resulting from an Actual or Imminent Release of Gaseous Radioactivity Exceeds 100 mrem Total Efkctlvc Dose Equivalcnt (TEDE) or 500 mRem Thyroid CDE Dose for the actual or projected duration of the release
(
All.)
6.1.3.a Dose IF ssessment - -
Dosc Assessment onc of tlic following 111 ilic MEA or bcyond
- IS CiilculiiIcd 011 thc SSCL.
11idlCiitCS EITHER TEDE -I-D:iy DOSC
> I.OEi-02 mRcm Thyroid-CDE Dose 2 S.OEt02 mRem bascd on Plant Vcnl etllucnt siiiiiplc analysis and U T on a dcfiiult Noblc Gas to lodiiic Ratio 6.1.3.b Field Measured IF s e Bat!- -
Dose ~
I C
nicasured at tlic Protected Arca Boundary or bcyond EXCEEDS 100 m Remlh r AND I
L I
Rclcase is espcctcd lo continue for 2 15 minutes L......
(
All 1
/
6.1.3.c
!?!!E!?
Analysis of field survey samplcs at tlie Protcctcd Arca Boundiity iiidicatcs EITHER onc of the following:
Field Survey IF 2 4.3GE+02 CCPM 3.85E-07 pCilcc 1-131 I
(
All
)
6.1.3.d Alarm IF Ilca!!ons--_
Valid High Alarm rcccived froni FRVS Noble Gas (Grid 1/3, 9RX680)
NPV Noble Gas (Grid ID, 9RX590)
SPV Noble Cas (Grid 1/3, 9RX580)
HTV Nohle Gas (Grid 3, Y R X S 16) onc of Ihc following Plant Effluent RMS Channcls r- -.... -..........
.L_
Total Plant Venl rcleasc rate EXCEEDS 7,6E+07 pCihw Total Noble Gas I
1 AND Refer to Attachment 3 SITE AREA EMERGENCY
... I Dose Asscssmcnt results NOT available I
.- -1 AND 1 - -. __.
Initiating Condition 4sgessrnent Dosc Asscssincnt I irdr ca~cs E ITHER OW of tlic following at tlic MEA or bcyorid a s calcul;itcd 011 rlrc SSCL:
TEDE -I-Day DOSC 2 I.OE+03 mRem Thyroid-CDE Dosc
> - S,OE+03 mRem biiscd on Plant Vcnl cmucnt s;rinplc analysis and 011
- I dcl;ruh Noblc G;rs 10 lodiirc Ralio OPCON EAL #
E M
E R c E
N c
Y A
C T
I 0
N L
E V
E L
S 1
I I
I I
I I
I I
I I
I I
I I
I I
I I
I I
I I
I I
I I
I I
I I
I I
I I
I
. I Action Required IICGS ECG Rev 00 6.0 Radiological ReleasedOccurrences Page 4 Of 4 6.1 Gaseous Effluent Release Boundary Dose Rcsulting from an Actual or Imminent Release of Gaseous Radioactivity Exceeds 1000 mrem Total Effective Dose Equivalcnt (TEDE) or 5000 mRem Thyroid CDE Dose for the actual or projected duration of the release
- 1..........
6.1.4.b Field Measured IF
---1 Dose Rate Dose Ratc measured at the Protcctcd Area Boundary or beyond EXCEEDS 1000 mRemlhr I AND r
..I I
Rclcase is espectcd to coritirrue for 2 15 minutes I
I
(
AI'
)
L.
6.1.4.c Field Survey IF nalysis Analysis of field Protcctcd Area Boundary iiidicatcs EITHER onc of rlic following:
survcy s;lrnplCs at the
>1.36E+03 CCPM
> 3.8SE-06 pCi/cc 1-131
('--AI;-)
6.1.4.d Alarm IF ndlcatlons Valid High Alarm rcceived from ANY onc of thc following Plmt Eflluent RMS Ch:inncls:
FRVS Noble Gas (Grid 1/3; 9RX680)
NPV Noble Gas (Grid 1/3; 9RX590)
SPV Noble Gas (Grid 1/3; 9RX580)
HTV Noble Gas (Grid 3; 9RX5 16)
L Total Plant Vent rclcase ratc EXCEEDS 7.6Ei-08 pCi/sec Total Noble Cas
~
. L A N D Dose Assessmcnt results NOT available.
~
I Release is ongoing for 2 15
.-T
.J THEN Refer to Attachment 4 GENERAL EMERGENCY
Initiating Condition OPCON EAL #
1 1
M I
I R
l E
l E
l G
I N
I c:
1' I
A I
'I' I
I 0
1 I
I I
c I
N I
L I
V I
L I
S I
I I
I
-. I Action Required 6.0 Radiological ReleasedOccurrences 6.2 Liquid Effluent Release 7---
~
Any Unplanned Relcasr of Liquid Radioactivity to the 13ivironmnt that exceeds 200 'I'imes Radiological lhluiicel Specifications for 15 minutes or longer Any Unplanned Rclwse of Liquid Radioactivity to the Ihviroiunent that Esceeds 2 Times the Radiological TLdinical Specilicalions lor 60 minutes or longer 6.2.1 6.2.2
- -1 IF Valid Cooling Tower Blowdown Emuent Radiation Monitor High Alarm Condition I
I+
1 I THEN I
1 AND
--- 1 Sample analysis of liquid effluent indicates concentration in excess of 2 times Technical Specification h i t s
.- -1 1 AND
-- --1
. I..
Release continues for 2 60 minutes after the alarm occurs AND 1-I I
1 I i
Tecliii ical Specification limits Sample analysis of liquid efluent indicates concentration in excess of 200 times I
I I AND 1
Release continues for 2 15 niiniites after the alarm occurs
- PN Refer to Attachment 2 ALERT
Initiating Condition OPCON C ' f Y
I I
I A
I C
t T
I I
I N
I L
I V
I L
I s
i 0 1 I
I I
.-I Action Required 6.0 Radiological ReleasedOccu rrences 6.3 In-Plant Radiation Occurrences 7
Unplanned Increase in Plant Radiation 6.3.1 IF
-7 Unplanned rise in radiation levels inside the Protected Area
> 1000 times nornial as indicated by EITHER one of the following:
Permanent or portable Area Radiation Monitors General Area Radiological Survey
-~
I THEN Refer to Attachment I 1
UNUSUAL EVENT EICGS b X G Rev. 00 Page I of I 7-Release of Radioactive Material or increases in Radiation Levels within thc facility that impedcs operation of systems required to mainlain
-\\-
safe operations or to establish
__I I____-I-or maintain Cold
____ Shutdown 6.3.2.a IF 6.3.2.b IF Unplanned Dose Rates
> 2000 mRem/hr in ANY area of the plant which requires ACCESS to maintain plant safety hnctions (EXCLUDING the Main Control Room and CAS) as indicated by EITHER one of the following:
Permanent or portable Area General Area Radiological Survey Radiation Monitors Unplanned Dose Rates
> 15 mRem/hr in EITHER one of the following:
Main Control Room Security Central Alarm Station (CAS)
I THEN
.L to Attachment 2 ALERT
Initiating Condition I
Visual Obsenation I
OPCON 1
EAI, #
1 M
I R
I I
G I
E l
I N
I C
I Y
I A
I I
I I
Action Required 6.0 Radiological Releases/Occurrences 6.4 Irradiated Fuel Event IWGS ECG Rev. 00 Page I of 2 I
Unplanned Increase in Plan1 Radiation
(
5 'j
~
6.4.1.a IF Uncontrolled water level drop in the Reactor Cavity as indicated by EITHER one of the following:
Visual Observation Reactor Water Level Shutdown Range Level Indicator I BBLI-R605 1..
I 7--
( __ __ All 1
6.4.1.b IF Uncoiltrolled water level drop in the Spent Fuel Pool as indicated by Valid Fuel Pool Low Level Alarm Condition I
Refer to Attachment 1 UNUSUAL EVENT
Initiating Condition OPCON EAL #
1 M
I E
/
R I
I G
I E
l I
N l
C I
Y I
A I
T I
0 1
I I
E l
I E
l I
I I
I c
I
I N
I L
I V
I L
I S
I J
Action Req u i red 6.0 Radiologicu. Releases/Occurrences 6.4 Irradiated Fuel Event Events that have or may result in uncovering Irradiated Fuel outside the Reactor Vessel Major Damage to
(
Irradiated Fuel 6.4.2.a IF 1
Major Damage to Irradiated Fuel has occurred I
1
] AND Valid High Alarm received from ANY one ofthe following RMS channels: -
RefUel Floor Exhaust Channel A (9RX627)
Refbel Floor Exhaust Channel B (9RX628)
Refbel Floor Exhaust Channel C (9RX629)
-I----
,-- i AI 6.4.2.b IF Unplanned rise on ANY one of the following Area Rad Monitors or by general area rad survey indicates
> 2000 mRemhr:
Spent Fuel Storage Pool Area (9RX707)
New Fuel Criticality Storage Channel A (9RX6 12)
New Fuel Criticality Storage Channel B (9RX6 13) 6.4.2.c IF Visual observation of Irradiated Fuel uncovered I-Refer to Attachment 2 ALERT HCGS ECG Rev. 00 Page 2 of 2
7.0 Electrical Power 7.1 Loss of AC Power Capabilities ltCGS ECG KCV. 00 Page I of2 THEN
~
,. Loss of o~sil~'po-w~e~an-d-'All-O~~~l~c-A-c
.. - _ _ _ _ - ~
AC power capability to Vila1 Buses reduced to a Single POW~?
( Source for greater than 15 minutes such that any additional single 1.0s~ of All OfTsIte Power tu Vilirl ikrscs for grciitzr )
{
Power to 4. I6 KV Vital Buses during either in Cold )
Initiating Condition
~liiui I5 ininutcs
- 4.
I6 KV Vital
~ Buys ' (.Shutdawn or
-. Refucling for greater
_ _ than 15
....... niinutq
_ _ ~. _ -
OPCON
(
All EAI, #
7.1.1 1
IF I
E 1
1 M
I I
lliiplaniied Loss of E
I j
Power from Station R
I Service Transformers G
I 1 1AX501 ANDIBX501 E
I
/
N I
j 4.16 KV Vital Buses Y
I I AND I
I to &L I
I c
I I
A I
T I
0 1 I
I I
C I
I I
N I
L I
V I
L I
S I
_1 Action Required
> 15 minutes have elapsed THEN
- i Refer to Attachment 1 UNUSUAL EVENT
('
I, 2, 3 7
7.1.2.a IF Loss of 4.16 KV Vital Bus Power Sources (Offsite and Onsite) which results in the availability of ONLY one 4.16 KV Vital Bus Power Source (Offsite or Onsite) 7
( 4, 5, Deheled )
7.1.2.b IF 4.16 KV Vital Buses are deenergized AND Refer to Attachment 2 ALERT J
lni tiating Condition OPCON EAL ##
I M
I I
R I
E l
G I
N I
Y I
A I
T I
I 0 1 I
I I
I E
t c
f C
I I
I I
N l
L I
V I
L I
Action Required 7.0 Electrical Power 7.1 Loss of AC Power Capabilities I IC(iS iC 15 minutes have elapsed THEN I
I Refer to Attachment 3 SJTE AREA EMERGENCY I
I---
Prolongcd Loss of All Offsiic and Onsitc AC Powcr lo All Vital AC Buscs i
2, 3
)
7.1.4.a AND I
I
-I 1
1, 2, 3 7.1.4.b AND Restoration of Power to at least one 4 16 KV Vital Bus within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is NOT likely Loss of any 2 Fission Product Barriers has occurred or is Imminent 1 THEN I THEN I
I
.~.
I 1
Refer to Attachment 4 1
GENERAL EMERGENCY
7.0 Electrical Power 7.2 Loss of DC Power Capabilities HCOS IICO Rev. 00 Page I of I
~
AND Unplanned Loss of All Vital 125 VDC Power during either Power Operation, Startup or Hot Shutdown for greater than _.
I5
- Minutes Initiating Condition Unplanned Loss of All Vital 125 VDC Power during either
( Cold Shutdown or Refbeling Mode for greater than 15 minutes OPCON EAL #
I T
I I
I L
I Action Required
( 4, 5, Defbeled )
7.2.1 IF I
1 IJnplanned degraded voltage condition for such that voltage is < 108 VDC Vital 125 VDC Buses,
-- 1 I AND
-1 I
I I
> 15 niiiiutes have elapsed
.. J 1
1 THEN 7.2.3 IF Unplanned degraded voltage condition for such that voltage is < I08 VDC Vital 125 VDC Buses,
> 15 minutes have elapsed I
-1 THEN I
1 I-I
~
Refer to Attachment 1 UNUSUAL EVENT 1
Refer to Attachment 3 SlTE AREA EMERGENCY
8.0 System Malfunctions 8.1 Loss of Heat Removal Capabililty HCGS ECG I
Rev 00 Page I of I Complete I.oss of Functions Needed to Achicve Cold Shutdown Conditions I Loss of Reactor Water Level hat has or will Uncover Fuel
.-l in the Reactor Vessel Inability to Maintain h e Plant in Cold Shutdown
-a' Initiating Condition OPCON EAL #
I I
E l
M I
I E
l R
I G
I E
I N
I c
Y I
I I
A I
T I
0 1
I I
C I
I I
N I
L I
V I
L I
I I
I S
I Action Required
(
4,5
)
8.1.2 IF Unplanned, Complete Loss of systems available to provide Decay Heat Removal functions Technical Specification required 1 AND
- - - I -
I AND 1
1 1
RCS Temperature has risen to
> 200°F (Excluding a <I5 nrinutes rise
>200°F with a heat removal function restored)
An UNCONTROLLED temperature rise is RAPIDLY approaching ZOOOF (with NO heat removal function restored) 1 I.
.-I. '
.I
- - ---\\
( -- 4, 5
)
8.1.3.a IF Reactor Water Level (Top of Active Fuel)
REACHES -161" I
8.1.3.b IF Loss of Main Condenser capabilities, as evidenced by I
an inability to remove Decay Heat from the Reactor I
1 AND I
1 Loss of Torus capabilities as evidenced by EITHER one of the following:
Entry into an Unsafe region of ANY one of the following curves:
- Heat Capacity Temperature Limit (HCTL) Curve
- Heat Capacity Level Limit (HCLL) Curve
- Pressure Suppression Pressure (PSP) Curve
I THEN I
I THEN I
- - 1 f - ---
Refer to Attachment 2 ALERT
I nit i a t i ng Condition OPCON EAL #
1 E
l E
l G
I N
I c
Y I
1 I
A I
I C
I T
I 1
1 0 1 N
I L
I V
I L
I s.,
Action Required M I I
R I
E I
I I
I I
15 minutes have elapsed since the loss of OHAs 8.0 System Malfunctions 8.2 Loss of Overhead Annunciators hpldnncd IAbS Of Mob1 Or dl llnplannrd Ius of Mosl or All Control Room (Annunciators dnd a significant Transient is in Progress or Compensatory \\
Atuiunciaiion or Indication in the Control 1
( 1,2,3 1 8.2.1 IF
' Unplanned loss of > 75%
1 of Main Control Room Overhead Annunciators
- -A
/TnEN-t 1 AND 15 minutes have elapsed since the loss ot OHAs 1
I 8.2.2.a I AND A ignificont transient""
1
- - I I
I is in progress
~ _ _ _.
THEN
_____.I-..
Refer to Attachment I UNUSIJAL EVENT 1
..A 1 AND BOTf I of the following.
CRIDS SPDS are NOT AVAILABLE I
I I
hiability to Monitor a Sigiiilicant 'I'raisieiit i n I%ogress Loss of > 75% of Main Control Room Overhead Annunciators I AND 1
I 1 A significant transient** is in progress I I I AND ROTt 1 of the following CRIDS SPDS are NOT AVAILABLE I AND Main Control Rooin liidications arc U T ii\\iiihblC to nioiiitor ANY one of the followlrlg
' RCS Status Reactivity Control ECCS Containment Parameters
-. THEN Refer to Attachment 3 SITE AREA EMERGENCY
- NOTE: A Significant Transient is based on EC judgment, but includes ias a minimum ANY one of the following:
RX SCRAM, LOAD REJECTION >25% POWER, ECCS INJECTION, THERMAL POWER OSCILLATION >lo%.
I nit iat ing Condition OPCON EAI, #
1 I
I E
l M I R
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Action Required 1
8.0 System Malfunctions 8.3 Loss of Communications Capabililty Unplanned Loss of All Onsite or Offsite Communications Capabilities
( - A l l - )
8.3.1.a IF Unplanned loss of ALL ONSITE comniunications as evidenced by the loss of systems:
Direct Inward Dial System of the following (DID)
(Gaitronics)
Station Page Systciii Station Radio System I
i 1 THEN
(
All
)
8.3.1.b IF Unplanned loss of ALL OFFSITE communications as evidenced by the loss of
&of the following systcms:
Direct Inward Dial Systciii Nuclear Eniergency Esss System (Centres)
(DID)
Tclcphone System (NETS) phone Y
Refer to Attachment 1 UNUSUAL EVENT
Initiating Condition OPCON Main Control Room Evacuation has been Initiated Action Required 8.0 System Malfunctions 8.4 Control Room Evacuation HCGS ECG Rev. 01 Page I of I cK3 0.4.2 IF Main Control Room Evacuation has been initiated I
and Plant Control cannot be established 8.4.3 1
Refer to Attachment 2 ALERT AND Control of the plant CANNOT be established from outside the Main Control Room within 15 minutes SITE AREA EMERGENCY
I Initiating Condition OPCON I
I A
I T
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I 0 1 N
I L
I V
I L
I I
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S I
Action Required 8.0 System Malfunctions 8.5 Technical Specifications
~
~
~
?
Inability to Reach Required Operational Condition within Technical Specification Limits
(
1,2,3
)
8.5.1 IF Plant is NOT brought to the REQUIRED Operational Condition within the Technical Specification required time limit 7
THEN Refer to Attachment I UNUSUAL EVENT 11ctis 1:co Rev. 00 Page I of I
9.0 Hazards - InternaVExternal I
9.1 Security Threats I
~ - - - _ _ _ --__
Security Event in a Plant Vital Area Security Event in a Plant Protected Area Confirined Security Event Which Indicates a Potential Degradation in I_. the
- - Level of Safety of the Plant
/
Initiating Condition OPCON EAL #
1 I
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0 1
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l M
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E 1
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I V
I L
I I
1 I
Action Required C A I 1
- ~-
.- -)
9.1.I IF -
Confirmed security threat directed toward the station as evidenced by ANY one of the fo 1 lowing :
Credible threat of malicious acts or destructive device within the Protected Area resulting in SCP-5 implementation Credible intrusion or assault threat to the Protected Area resulting in SCP-5 implementation Attempted intrusion or assault to the Protected Area resulting in SCP-7 or SCP-11 implementation Malicious acts attempted or discovered within the Protected Area resulting in SCP-10 implementation normal plant operations resulting in SCP-8 implementation Destructive Device discovered within the Protected Area resulting in SCP-IO implementation Hostage/Extortion situation that threatens I " N Refer to Attachment 24 UNUSUAL EVENT (Common Site)
-.\\
.. /'
9.1.2 Con firmed hostile intrusion or inalicious acts as evidenced by ANY one of the following:
Discovery of an intruder(s), armed and violent, within the Protected Area resulting in SCP-6 implementation Hostage held on-site in a non-vital area resulting in SCP-I implementation Malicious acts attempted or discovered within a Vita! Area (which is in the Protected Area) resulting in SCP-10 implementation that does not meet criteria of 9. I.3 THEN Cii) 9.1.3 -
~
IF __
Confirmed hostile intrusion or malicious acts in Plant Vital Areas that involve actual or likely major failures of plant functions needed for the protection of the public as evidenced by any one of the following:
Discovery of an intruder(s), armed and violent, within the Vital Area resulting in SCP-6 implementation Destructive device discovered in a Vital Area resulting in SCP-10 implementation HCGS ECG Rev. 01 Page I of 1 Security Event Resulting in Loss o Ability to Reach and Maintain Cold Shutdown 9.1.4 IF Security event resulting in the actual loss of physical control of EI'I'IIER one of the following:
Control Room Remote Shutdown Panel Malicious act(s) attempted or discovered in a Vital Area resulting in SCP-10 implementation AND the malicious act(s) is of a magnitude that it SPECIFICALLY results in Damage to ANY one of the following:
TWO OR MORE Trains of a Safety System MORE THAN ONE Safety System Any Plant Vital Structure which renders the structure incapable of performing its Design Function Damaged Safety System(s) or Plant Vital Structure is required for the present Operational Condition AND THEN
I ni t iat ing Condition OPCON EAI, ##
Action Required Any Plant Vital Structure which renders thc structure incapable of performing 11s Design Function E
M E
R c;
E N c Y
A c 1'
I 0
N L
E V
E L
S I
I I
I I
1 I
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I I
I I
I 1
9.0 Hazards - InternaUExternal Fire within the Protected Area Boundary Not Estinguished wilhin 15 minutes of Detection I
All
[
All 1
9.2.1 IF 9.2.1 IF I
1 Rcpon of a fire from personnel at the scciic II
...~
Valid Fire Alarni is received iii llic Main Control Room 1_
AND I '
~-..... -
Fire is within ANY onc of the following Plaiit Slnsturcs (EXCLUDING snliill fires that have NO potential to affect Safety Systems or Protected Area Pcriiianenl Plant Structures) 9.2 Reactor Building j
Turbine Building Conrrol/Aus Building Scnicc/R;id WiiSIC Building Low Lcvcl Radwastc Interim Storage Facility
' I I
- i
. 1 AND'-. _.
i Fire IS NOT cstinguished within 15 minutes of I, 8
I I
Receipt of a Valid Fire Alarm Repon of a fire froai llrc scciic:
ElTHER one of the following:
i e
+ THEN Refer to Attachment 1 1
IJNUSIJAI, EVENT I
Fire 1
I I
i 1
Fire Affecting the Operability of Plant Safety Systems Rcquircd to Establish or Mainlain Safe Shutdown
(
All -
9.2.2 IF Fire w~tliin ANY OIIC of tlic following Pl;in~ VII;II SIructurcs Rcactor Building ConlroVAus Building Scnice Walcr Intake Structure ScniccRid WiistC Building
I Rcactor Building Coiitrol/Aus Building Service Water Intake Slructure Scnicemd Waste Building Initiating Condition The Explosion is of a niagnitudc 11i;it it SPECIFICALLY rcsults in Damage to ANY one of tlic following:
I OPCON EAI, ##
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E c:
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I N
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I I
I Action Required 9.0 Hazards - InternaUExternal 9.3 Explosion Esplosis Affecting the Operability of Plant Safety System Required to Establish or Maiiitain Safe Shutdown i
Natural and Destructwe Phenomena i
Affecting the Protected Area 9.3.1 IF Confirmed Explosion within the Protccted Area I
I 1 AND I
~
i I
I Rcpon of visiblc daniagc to P l m equipnicnt or Protected Area Pcrniancnt Plant Structurcs
! THEN 1
t.
Refer to Attachment 1 UNUSUAL EVENT I
I nit i at i n g Condition OPCON EAI, #
I I
E l
h-I I
I E
l H
I I
E:
N I
1' I
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0 1 I
I I
I N
l L
I V
I 13 I
I I
I E
E s
i Action Required 9.0 Hazards - InternaUExternal 9.4 Toxic/Flammable Cases I
Release of Toxic or Flammable Gases Deemed Detrimental to Safe Operation of the Plant
[
All I
9.4.1.a IF I
1 Notification by Local, County, or i State Officials for the potential need j
to EVACUATE i
non-essential personnel I
due to an Off'site Toxic Gas release I
j AND I
SNSS deems evacuation j
of non-essential personnel I
is required I
I I
I I THEN I
Refer to Attachment 24 UNUSUAL EVENT (Common Site)
I I
Aii 1
9.4.1.b IF Uncontrolled Toxic Cas release within the Protected Area in ANY area which does not norinally require an atmospheric survey or Respiratoiy Protection for entry 9.4.1.c IF Uncontrolled Flammable Gas 1
Protected Area I
~
i i
release within the that RESILTS in Flanimable Gas concentrations 25% of the LEL j
I I
E xc EE r) i NG I AND Routine Plant Operations are IMPEDED based on EITHER one of the following:
Access restrictions caused by the uncontrolled release Personnel injuries have occurred as a result of the release I THEN Refer to Attachment I UNUSUAL EVENT
Initiating cr? n:i i t io n OPCON EAI, ##
I I
E l
kl I
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(;
I E
l N
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I
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I V
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I 1
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Action Required of the following Plant Structures Reactor Building 9.0 Hazards - InternaVExternal 9.4 ToxidFlammable Gases I IC( ;s ECG Kzv 00 Iagz 2 of 2 Release of Toxic or Flammable Gases Within a Facility Structure Which Jeopardizes Operation of Systems Required to Maintain Safe Operations or to Establish or Maintain Cold Shutdown Conditions i
Ail 9.4.2.a IF IJncontrolled Toxic Cas release within ANY one of the following Plant Structures Reactor Building Turbine Building Control/Aux Building Service Water Intake Structure Service/Rad Waste Building
[ AND Toxic Gas concentrations result in ANY one of the following.
An IDLt1 atiiiosphere Plant personnel report severe adverse health reactions,
[
The Threshold Limit Value (TLV) being EXCEEDED including burning eyes, nose, throat, dizziness
- -r
- 1.
~..~ -.
1 AND Plant personnel are unable to perform actions necessary to complete a Safe Shutdown of the plant without appropriate personnel protection equipment 1
/, THEN Refer to Attachment 2 ALERT
Initiating Condition OPCON EAL #
I I
E l
M I
I E
l R
I G
I E
l N
l c
I I Y
I I
I A
I I
I c
'r I
I N
l L
I V
I 12 I
I 0 1 I
I I
S I
Action Required 9.0 Hazards - InternaVExternal 9.5 Seismic Event Natural and Destructive Phenomena Affecting the Protected Area i An-',
9.5.1.a IF i Seismic Event fell I
by personnel within the I Protected Area I
(
AI1
)
9.5.1.b IF Valid Actuation of the Seismic Trigger (> 0.01g) has occurred as verified by the SMA-3 Event Indicator (flag) being WHITE on Panel IO-C-673 in the Upper Relay Room v
Refer to Attachnient 24 I
i I UNUSUAL EVENT (Coninion Site) j I
I
)
Natural and Destructive Phenomena Att'ecting the Plant Vital Area i
(
All 1
1 AND 9.5.2 Valid Actuation ofthe Seismic Switch (> 0.lg) has occurred as veritied by Ell1 IER one of the following:
Valid actuation of Main Control Room Overhead Annunciator C6-C4 Power Supply Drawer is lit on Panel IO-C-673 in the Upper Relay Room AMBER Alarm light on the Seismic Switch 1 THEN j
Refer to Attachment 2 A L E R I
Initiating Condition OPCON Ac t io n Required 9.0 Hazards - Internal/External 9.6 High Winds i
(
Natural and Destructive Phenomena Affecting the PI otected Area All
)
(
All I
i Natural and Destructive Phenomena Affecting the Plant Vital Area
(
All 1
9.6.1.a 9.6.1.b IF IF 9.6.2 I
I
- 1 I
Report of a Tornado IOUCI 1ING DOWN within the Protected Area Sustained wind speeds
> 75 M PH for 15 minutes, measured at ANY elevation of the Met Tower i I THEN I
- j j
1 Refer to Attachment 24
! UNUSUAL EVENI (Cooinioe Site)
The Wind Speed is of a magnitude that it SPECIFICAL1.Y results in Damage to ANY one of the following.
Rendering ANY of the following structures incapable of pertbriiiing TWO OR MORE subsystems of a Safety System MORE THAN ONE Safety Systeni its Design Function
- Reactor Building
- Control/Aux Building
- Service Water Intake Structure
- Service/Radwaste Building I
1 AND Damaged Safety System(s) or Plant Vital Structure is required for the present Operational Condition THEN Ketkr to Attachment 2 A I,E IW
9.0 Hazards - InternaVExternal 9.7 Flooding I nit iat ing Condition Internal Flooding in Excess of Sump Handling Capability Affecting Safety Related Areas of the Plant Internal Flooding Aftecting the Operabilityof Plant Safety Systems Required to Establish or Maintain Safe Shutdown
(
I OPCON
-All
)
9.7.1 EAL #
I IF I
M I
I E
l R
1 1
I 1
Reactor Building Floor Levels above the Maximum E
l Viwil Observation of Uncontrolled Flooding that confirms ANY one of the following c
I N
I y
Norrnal Floor Level (> I ' I ) referenced in EOP 103, Secondary Containment Control E
Receipt ofa SSWS Pump Room Flooded Alarm Greater than 2" of water in ANY area that contains a I
I A
I
'I' I
0 1 c
1 I
1 1
N l
L I
V I
L I
I I
I S
I Action Required Siifety System(s), not included above i
I t I
Refer to Attachment I I I UNUSUAL EVEN'I' All -
9.7.2 IF Visual Observation of Flooding within ANY one of the following Plant Vital Structures I
Reactor Building ControVAux Building Service Water Intake Structure I
Service/Kad Waste Building I AND The Flooding is ofa magnitude that it SPECIFICALLY results in Damage to ANY one of the following TWO OK MORE subsystenis of a Safety Systetii MORE I H A N ONE Safety System Any of the above listed Plant Vital Structures which renders the structure incapable of performing its Design Function I AND i
~
i j 1
I I
I I
i I
Ilaniaged Safety System(s) or Plant Vital Structure is required for the present Operational Condition I
THEN t
ALEK'I' 1 liefer to Attachment 2
Initiating Condition OPCON EAL #
I I
E l
kl I
E l
R I
G I
E l
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I 0 1 I
I I
c I
I I
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V I
1, I
I I
I I
E t
s t
Action Required 9.0 Hazards - InternaYExternal 9.8 Turbine Failure / Vehicle Crash / Missile Impact IlCGS F,CG Rev. 00 Page I of 1 Natural and Destructive Phenomena Affecting Certain Structures Within the Protected Area Natural and Destructive Phenomena
( Affecting Certain Structures Within the Plant Vital Area
( - -r-1 9.8.1.a IF Catastrophic damage to the Main Turbine as evidenced by EITHER one of the following:
Main Turbine casing Main Turbine/Generator penetration Ihrnage potentially releasing Lube Oil or Hydrogen Gas IO the Turbine Building
( - Ail-
)
9.8.1.b IF I
Vehicle Crash /
Missile Impact with or within ANY one of the following Plant Structures:
Reactor Building Turbine Building ControVAux Building Service Water Intake Service/Rad Waste Building S t ructure I
1 THEN 1
Refer to Attachment 1 UNUSUAL EVENT
(
All -
9.8.2 IF Vehicle Crash / Missile Impact with or within ANY one of the following Plant Vital Structures Reactor Building ControVAux Building Service Water Intake Structure Service/Rad Waste Building I AND I
The Vehicle Crash / Missile Impact is of a magnitude that it SPECIFICALLY results in Damage to &Y one of the following 1
~
I I
TWO OR M O B subsystems of a Safety Systeni MORE THAN ONE Safety Systeni Any of the above Plant Vital Structures which renders the structure incapable of performing its Design Function I
I AND Damaged Safety System(s) or Plant Vital Structure I
1 is required for the present Operational Condition ATHEN Refer to Attachment 2 ALERT
Initiating Condition OPCON EAL # -
1 I
M I
I E
l R
I G
I E
l N
I C
I Y
I A
I T
I 0
1 I
I I
I I
c I
f N
I I
I L
l E
l V
I E
l L
I I
I 1
Action Required 9.0 Hazards - InternaVExternal 9.9 River Level Natural and Destructive Phenomena Meeting the Protected Area
. -~
(
All j 9.9.1.a IF River Level > 99.5' 1- -
I-.
(\\ -AN - - )
9.9.1.b IF River Level < 80.0' THEN I
Refer to Attachment 24 UNUSUAL EVENT (Common Site)
11.0 Reportable Action Levels 11.1 Technical Specifications rc...................
INI'IMIWN 01: ANY IINI'I' SI l\\l'll)OWN
~ l 1 l ~ ~ U l ~ l ~ l )
I3Y 'l'lll: 'I'IX'I INICAI. SI'IXII:ICAI'IONS
\\ -
1 IOCFRS0.72(b)(2)(i) 11.I.I
.a IF Unit shutdown is IN l'llA'~3l to comply with
'I'cclinical Specilications r.-
THEN 1
Refer to Attachment 14 4 I lour Rcnort
... 1 I!XCEI3)ING ANY 'I'IXI INICAI, SI'I:CII~ICAI'ION S A I W Y I.IMII
[ IOCFR50.36(c)( I). 'IS 6.7. I.a]
\\. \\
~
( 1,2,3,4,5
_. (as applicable in T / S D 1l.l.l.b IF Exceeding ANY one of the following
'Technical Specification Safety Limits:
T/S 2.1. I, 'I'IIERMAL POWER, Low Pressure or Low Flow
' 'I/S 2.1.2,THERMAL POWER, High Pressure and High Flow T/S 2.1.3,REACTORCOOLANT SYSTEM PRESSURE T/S 2.1.4, REACTOR VESSEL WATER LEVEL I ICGS ECG Rev. 02 Page I of 2 ANY I)I~VIA'fION FROM T/S 011 I.ICENSI: CONI>I'I'ION I'URSUAN'I '1'0 lOCl~.RS0.54(~)
I IOCFR50.72(h)(l I]
c
___ - - 1 L
11.I.I
.c IF Action required because no action consistent with Technical Specifications or license can provide adequate or equivalent protection in an emergency (See NC.NA-AP.ZZ-0005 (Q) for guidance on deviation from written procedures)
___i----
THEN THEN
.. ~. ~
I Refer to Attachment 12 1 I lour Report
11.0 Reporta Action Levels Iagc 2 of 2 11.1 Technical Specifications VIOLATION OF THE REQUIREMENTS \\
C O N T A I N E D IN THE OPERATING LICENSE ) ( -
ANY EVENT REQUIRING AN ENGINEERING EVALUATION BY TECIINICAL SPECIFICATIONS OR COMMITMENT Initiating Condition
,~
\\
... [IICGS Operating License, Sections 2.F]
....--/ / \\
~
[T/S 3.4.6.1,3.4.4, 3.7.51
~
~
.--,/
I I.I
.3.a IF Violation of ANY one of the rcquircnicnts contained in Section 2.c o11hc ()pcrating I.iccnsc l<X~*I-~Il
- IS otlimvise provided in the I cchn ical S pcc i I ica t ions o r Ilii\\~ironiiicnr~il Irolcclion 11;iii I
1 THEN I I.I
.3.b IF Any of the
-r/s 1.cos for I< < S Iressure/
Ienipcrat ure (T/S 3.4.6.1) are exceeded thereby requiring an Engineering Evaluation.
I I
Refer to Attachment 20 7 A IJ,-...-
D,-,..A I
All 11.I
.3.c IF The conductivity, chloride concentration or pl I in the RCS is in excess of its specified limits per TIS 3.4.4 Action Statement C. 1 thereby requiring an Engineering Evaluation 11.I
.3.d IF One or more snubbers are found to be IN O 113 I1A 13 I,F and have been replaced or restored to an 0 I 13 R A 13 LE stat us, an Engineering Evaluation shall be performed per TIS 4.7.54 THEN Refer to Attachment 22 c\\rr-i I 17I) I>..-.......
I
. L
I 11 i I i ;i t i 11 g C '()id i t ioii I
I A
I T
I 0 1 C
I I
I I
N l
I Action I>,.I...:..n,1 11.0 Reportable A tion L els I ICGS ECG liev. 01 Page I of I 11.2 Degraded or Unanalyzed Condition 11.2.1 IF A s judged by thc OS/l~l)O, an event or condition that results in ANY one of the ti,llo\\ving:
I he condilion oftlie plant. including its principal safcty barricrs, bcing scriously degraded;
'I'he plant being in an unanalyzed condition that significantly degrades plan1 safety.
I THEN TI IAr COULD I IAVE 1'I~I:VENl'ED Cl:lCl'AlN SAFETY FUNC I'IONS
[ IOCFR50.72
-- (b)(3)(v)]
(-----Ai --I I I
.2.2.b Any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to pcrforni ANY one of the following:
- 4. Shutdown the reactor and maintain it in a safe shutdown condition B. Remove residual heat.
C. Control the release of radioactive material D. Mitigate the consequences of an accident
. I THEN I
I r.
I I
Refer to Attachment 26
11.0 Reportable Action Levels 1 I.3 System Actuations Actuation is NOT part of a pre-planned sequence during testing or reactor operation.
..~
RAl, #
It E
I' 0
H T
I N
c; A
C T
I 0
N L
E V
E L
S Action I
Refer to Attachment 26 8 Ilour Renort I
I I
I I
I I
I I
I I
I I
I I
I I
I I
I I
I I
I I
I I
I I
I I
I I
I I
I I
Initiating C' o n d it io 11
( 1 IT0 N Required (1
All 1
-../
11.3.1 IF Valid ECCS Actuation, Manual or Automatic, has or should have occurred 1 AND ECCS results, or should have resulted, in discharge to the vessel
~
ACI'IJATION 01: KCX"OK
(
\\
I I OCI~'RS0.72(b)(2)(iv)(l3)~
Wl lliN CIU'I'ICAI, I~NCliI'I' lll<lil'l.ANNl~l)
IF Any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical except when the actuation
- esults from and is part of a pre-plannec sequence during testing or reactor operation.
f Refer to Attachment 14 4 Iiour ReDort THEN I ICGS ECG Rev. 02 I'agc I of I
,~.............
-\\
I VAI II) AC I IJA TION 01 I IS I I I) SY5 I I'hl I XCI I) I 1'1U PI ANN1 I) [ IO('1 It50 72(b)(3)(iv)(A)l
(-
All
')
11.3.3 IF Any event or condition that results in valid actuation of any system listed in Technical Basis 1 1.3.3 except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.
L THEN
11.O Reportable Action Levels 11.4 Personnel Safety / Overexposure Refer to Attachment 17
- _____ 1 c -.A!!..
OI'CON 1tAL ##
1 I.4.1 I
IF I
I I I
I A
i T
I 0 1 I
I 1
N l
L i
E l
V I
L l
s I
I 1
I E
. -..I Action Renilired B
for overcsposure as indicated by
--- ANY one of the follrnving:
W I ) K exposure > - 25 Rem I,DK exposure > - 75 Hem Sl)K exposure > - 250 Item l<clcase of radioactive material nside or outside of a Restricted Area so that had an individual been present for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> the individual could have received > 5 times the occupational ALI (Annual Limit of Intake) which would usually equate to 25 Rem CEDE. I'his Does NOT apply to areas where personnel are NOT normally stationed during routine operations.
Refer to Attachment 12 1 Iiour Kenort 1 I
.4.2.a IF PERSONNEL OVEREXPOSURE or potential for
.)verexposure within a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period, as indicated by ANY one of the following:
TEDE exposure > 5 Rem I,DE exposure > 15 Rem SI)E exposure > SO Item Release of radioactive material inside or outside of a Restricted Area so that had an individual been present for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> the individual could have received > 1 times the occupational ALI (Annual Limit of Intake) which would usually equate to 5 Rem CEDE.
'I'his Does NO'[ apply to areas where personnel are N O 1 normally stationed during routine operations.
- - -1 THEN Refer to Attachment 14 4 lour Report L
A
. - _- -I I I.4.2.b IF Any fatality has occurred within the Owner Controlled Area (OCA) 1...........
I THEN 1
1 1.O Reportable Action Levels 11.4 Personnel Safety / Overexposure
'IO A l ~ M l N l S l ' l ~ A l l V ~ ~
ERROR (I3L.IND TfS'l 13Y I A 1 3 )
I IOC'I~'1126, AI'I'. A, 2.8(~)(5)1 1'
- -.~.
SIGNIFICANT FITNESS FOR EVENTS [ I OCFR26.731 Y C'ON'I'AhllNA'llil) I'IXSON ~ l l l A N S l ' O l C l ~ l ~ ~ ~
'ti '1'0 AN Ol~'l~Sl~l'l~
hll~l)lCAl, l~AC'll,l'l'Y I X ) R l'R~~:AI'hlliN'l' I I OC'):R50.72(b)(3)(sii)j
,) I (,,
Action Refer to Attachment 27
.\\..
Q lIn....
D.-...,-...t
( -
3 I I.4.3.a IF Any event that is determined to be rcportahlc by the Mcdical Review Officer (MRO) or designee IAW I'SEG Nuclear's Fitness for Duty Program (NC.NA-AP.ZZ-O042(Q))
AND
- A The reportable details of the event are made available to the OS by the MRO or designee THEN I Refer to Attachment 19 1
? A 1 1 -..- n----4 1
11.4.3.b IF The occurrence of a false positive error on a blind lab performance test specimen under I OCFR26 as determined by the Medical Review Officer (MRO) IAW PSEG Nuclear's Fitness for duty Program (NC.NA-AP.ZZ-0042( Q) 1 AND 1...............................
r---
The reportable details of the event are made available to the OS by the MRO or designee THEN I Refer to Attachment 19
- A I I...... n
?.
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11.0 Reportable Action Levels 11.5 Environmental / State Notifications Initiating
('ondition 1
Action Reauired I
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I 11.5.2.a IF Spillldischarge of an industrial chemical or petroleum product outside of a plant structure within the Owner Controlled Area (OCA) that results in EITIIER one of the fol lowing :
Spill /discharge that has passed through the engineered till and into the ground water as confirmed by licensing Spill / discharge that CANNOT be cleaned up within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and no contact with groundwater is suspected r--------.
THEN
(- All j I 1 5 2. b IF EITIIER one of the following events occur:
Observation of a spiWdischarge of an industrial chemical or petroleum product from on-site into the Delaware River or into a storm drain Observation of an oil slick on the Delaware River from any source Note:
This event MAY require IMMEDIATE
( I 5 minute) notifications. DO -
NOT delay implementation of Attachment 16.
I--
_____ J THEN Refer to Attachment 16 I
SpilVDischarge ReDortinrz I
Refer to Attachment 16 1
Snill/Discharrze Renorting
. i
\\.-
I ICGS f3CG I<cv. 01 I'ilge 1 of 2
(- All
'1 11.5.2.c IF L.
i s judged by the OS/EDO, ANY one of he following events has occurred:
Unusually large fish kill Protected aquatic species impinge on Circulating or Service Water intake screens (eg.; sea turtle, sturgeon) as reported by Site personnel Any occurrence of an unusual or important event that indicates or could result in significant environmentill impact casually related to plant operation; such as the following:
+
Onsite plant or aniinal disease outbreaks Mortality or unusual occurrence o any species protected by the Endangered Species Act of I973 Increase in nuisance organisms or conditions Excessive bird impactation
+
NJPDES Permit violations
+
Excessive Opacity (smoke)
-._ -. -_ __ lTHEN Refer to Attachment 15 I
Environmental Protection Plan
Initiating C. o nd i t io 11 OICON RAL #
1 I
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Action Iteq u ired 11.0 Reportable Action Levels 11.5 Environmental / State Notifications I ICGS t i u i Rcv. 01 Iagc 2 of 2 c All ]
1 1.5.3 IF EITHER one of the following events occur:
Personal injury due to an occurrence to a boiler or pressure vessel A boiler or pressure vessel explosion Q
THEN
-1 Refer to Attachment 28 B&PV Reporting
Initiating Condition MODE ItAL #
---I I
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11.0 Reportable Action Levels 11.6 After-the-Fact EMERGENCY CONDITIONS
~
AFTER-Ti {E-FACT
(-xi-->
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.6.1 IF Discovery of events or conditions that had previously occurred (event was NOT ongoing at the time of discovery) which EXCEEDED an Emergency Action Level (EAL) and was NOT declared as an emergency More than ONE HOUR has elapsed since the condition occurred I
AND There are currently NO adverse consequences in progress as a result of the event THEN Action Rea u ired Refer to Attachment 12 1 Ilour RePort
11.0 Reportable Action Levels 11.7 Security / Emergency Response Capabilities I I I
I I
1 IlCGS ECG Rev. 04 Page I of 1 Any Non-Emergency safeguards event that is reportable in accordance with 10CFR73.71 as determined by Security (SCP-I 5)
MAJOR LOSS OF EMERGENCY ASSESSMENT CAPABILITY, OFFSITE RESPONSE CAPABILITY, OR COMMUNICATIONS CAPABILITY [ 10CFRSO.72(b)(3)(xiii)]
FEGUARDS EVENTS TIiAT ARE DETE T O BE NON-EMERGENCIES, BUT ARE REP II1E NRC WIIItlN ONE HOUR [ IOCFR7 OPCON RAL ##
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P 0
R T
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1 I.7.1.a IF
(- All )
11.7.1.b IF I
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Action Required I 1.7.1.c IF
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OS/EC determines that an event(s) (excluding a scheduled test or preplanned maintenance activity) has occurred that would impair the ability to deal with an accident or emergency as indicated by the Loss of ANY one of the following:
THEN Refer to Attachment I 1 1 Hour Report (Common Site)
Nuclear Emergency Telecommunications System (NETS) for > 1 hr ENS for > 1 hr in the Control Room, TSC, and EOF
@/A if reported by the NRC).
More than 17 Offsite Sirens for >
1 hr Use of the EOF for > 8 hrs All Meteorological data (Hope Creek AND Salem) for one parameter for > 8 hrs Site access due to Acts of Nature (snow, flood, etc.)
I Refer to Attachment 25 8 Hour Report (Common Site)
Use of the TSC for > 8 hrs SPV, NPV, or FRVS vent radiation effluent monitors with no alternate method of monitoring for > 72 hrs SPDS OR CRIDS for > 8 hrs More than > 75% OHA's Concurrent multiple accident or emergency condition indicators which in the judgement of the OS significantly impairs assessment capabilities Refer lo Technical Basis I I.7.1.E for ERD!
THEN f
Refer to Attachment 26 8 Hour Report
Act ion 1)',,...;..,.,I A news release is planned
~
11.O Reportable Action Levels 11.8 Public Interest I ICGS ECG IW. 0 I I'age 1 of I
/GNUSUAL CONDi-riONS DIRECTLY AFFECTING I
( AILOWAYS CREEK TOWNSIIIP (LACT) [LAC -MOU]
tJNtJSCJAI, CONI11 I IONS WAKRAN I I N 6 A NEWS I<LI,IIASII O K N O I'IFICA IION OF GOVERNMEN I' AGl.:NCIES [ I OCFK50 72(b)(2)(ui)l I
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I AND Notifications to a Local, State or Federal agency has been or will be made I
I 4 1 1 n
I I
Refer to Attachment 14
/
-~...
(
._ *I'
__ J 11.8.2.b IF As judged by the OS/EDO, events which are the responsibility of PSEG Nuclear which have or may result in EITHER one of the following:
Anticipated unusual movement of equipment or personnel which may significantly affect local traffic patterns Onsite events which involve alarms, sirens or other noise which may be heard off-site THEN I
Refer to Attachment 21
-. -~-.
- I I
111 it iat i ng Condition ANY one of the following events occur involving Special Nuclear Material (SNM) or Spent Fuel:
Loss, other than normal operating loss, of SNM Actual or attempted theft or unlawful diversion of SNM Shipment of SNM of low strategic significance or Spent Fuel that is lost or unaccounted for after the estimated time of arrival A lost or unaccounted for shipment of SNM of low strategic significance or Spent Fuel has OI'CON I
Action It e (1 u i red Il('(iS I,:('(;
Itev 02 Page I o f 2 11.0 Reportable Action Levels 11.9 Accidental Criticality / Special Nuclear Material / Rad Material Shipments - Releases 11.9.1.a IF Any unplanned or accidental criticality THEN f
Refer to Attachment 12 I Ilour Report I
[
been recovered or accounted for THEN
' l ~ l ~ l ~ l ~ l ~
OK LOSS OF I.I CI:N S 13) M A'I'ER I A 1, I lOCFK20.2201(a)(
I)(i)l 11.9.1.c IF Lost, stolen or missing licensed material > 1000 times the quantity specified in 10CFR20 Appendix C in such circumstances that an exposure could result to persons in Unrestricted Areas. - - - - -.
I THEN Refer to Attachment 1 1 1 Ilour Report (Coninion Site)
I nit ialing
('ondi tion OI'CON ItAL #
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. I Action Required
_ _ _ _ - ~ -
Accidents during the transportation of radioactive material which are reported to PSE&G I IC(IS ECG Rev 02 11.0 Reportable Action Levels 11.9 Accidental Criticality / Special Nuclear Material / Rad Material Shipments - Releases I'llgC 2 Of 2 J
/ /('()NTAMINA.I.ION OUISII~I; 01: TI 11:
1IA1)101,OGlCAl.I.Y CON'I'IIOI.I,l<l) -)
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.- AREA ~lOCI'lI50.73(b)[2)(si))
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--j I:XCIISSI v IJ CONTAM INAIWN' ANINOR IL4I)IA'I'ION I.I'VI3.S O N A I'ACKAGI< I IOCI'I120.1906(d)],,
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( _... All
....- 3 11.9.1.e IF THEN Receipt survey indicates that pac kagc contani inatiodradiat ion lcvels equal or exceeds ANY one of the following:
2200 dpm/IOO cm2 200 mWhr on contact 10 mWhr at 3 feet Refer to Attachment 10 I Hour Report I -
c 1
11.9.2.b IF Discovery of a Contaminated Area OUTSIDE of the RCA with removable activity due to licensed material as the shipper that involve (or potentially involve) damage to the cargo THEN I
.-_.__.--_I_
Refer to Attachment 18 4 Hour Report Location of Contaminated Area is OUTSIDE of Plant Structures 1.-
. _. -. I Location of Contaminated Area is such that a contaminated person or material may Size of Contaminated Area is LARGE
(> I00 FT2) have left the Protected Area 1
ITHEN J
Refer to Attachment 13 4 IIour Report
Initiating Condition Of'C'ON Action 11.0 Reportable Action Levels 11.10 Voluntary Notifications
~ _ _ _ - -
Events/conditions warrant voluntary/courtesy NRC notification [ 1 OCFR50.72 - Voluntary Report]
(7ii
\\. 7
/
11.10.2 IF In thc judgement of the OS, i
notification to the NKC is warranted I AND
- 1 NO dthcr 13ALs or RALs appear to be applicable THEN r
I I ICGS ECG Kcv. 01 Page I of 1
~~~
UNITED STATES F!L'CLEAR REGULATORY COMMISSION WASHINGTON. D.C. 205550001
\\'
PSEG NUCLEAR LLC DOCKET NO. 50-354 HOPE CREEK GENERATING STATION FACILITY OPERATING LICENSE License No. NPF-57 1: The Nuclear Regulatory Commission (the Commission or the NRC) has found that:
A.
The application for a license filed by the Public Service Electric &
Gas Company, acting on behalf of itself and Atlantic City Electric Company (the licensees), complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Ccmmission's regulations set forth in 10 CFR Chapter I, and all required notifications to other agencies or bodies have been duly inaae ;
Csnsmxtion of the Hope Crsek Generating Station (the facility) has Seen subs:antially completed in conformity with Construction Pernit Nc. C??R-120 and the application, as amended, the provisions of the Acz and the regulations of the Commission;
-. -?.e facility will operate 12 conformity with the application, as mmded. the provislcns ~ l f Eke Act, and the regulations of the C~lx.rss;cn (except a3 exezF:ed frcm compliance in Seczion 2.D. below);
3'-
,..ere is reascnable asszrbxs: ( 2 ) that the acrivities authorized by L
i
- 2sera:ing license car. be fzr.ducted withouz endangering the health
--.-i.,--
--..,,--ed in ccmpliance xi:k the Cmmission's regulations set forth ;n aT.2 safety s: the publ-c, ar.5 (ii) thac such activities will be I 2 CFF. Chapter : ( e x c e ~ ~
as exempced frcm compliance in Sectic;! 2.D.
=e,2.**, ;
I
?SES Xclear LLC is te=b::=a:ly qxalified t ~ ?
engage in the accivi:ies aq::.L.critea by this ::=.esse ir. acccrdance with :be Cmmiss:oz's reTdlacions set fcrth ir. 12 CTi? Chapter I; 1
"ke l:censee has sazisfied :he applicable provisions of 10 CFR Part
---, 'Finaxial Protec::x ReTLirements and Indemnity Agreements,
" cf the Commission's regulat'cx;
- A..
-. -ne issuance of this licezse w i l l not be inimical to the common defense and security o r z c the health and safety of the public; PSE&G CONTROL I
COPY # c,/a Amendment No.
135 I
ti.
After weighing the environmental, economic, technical, and cther benefits of the facility against environmentai a d other coscs and considering available alternatives, the issuance of Facility Operating License No. NPF-57, subject to ?he conditions.for proteczion of the environment set -forzh in the Environmental Protection P l a ~
attached as Appendix B, is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been sacisfied; and I. The receipt, possession, and us2 of source, byproduct and special nuclear ma:eriai as authorized by this license will be in accordance with the Commission's regulations in 10 CFR Parts 30, 40 and 70.
2. Based on the foregoing findir.gs and approval by the Nuclear Regulatory commission at a meeting on July 21, 1986, the License for Fuel Loading act!
Low Power Testing, License No. NPF-SO, issued on April 11, 1980, is superseded by Facility Operating License NPF-57 hereby issued to PSEG Nuclear LLC (the licensee), =o read as follows:
A.
This license applies t 3 the Hope Creek Generating Station, a boiling xater nuclear reactor, a?.d associated equipment (the facility) owned by PSEG Nuclear L i z. The facility is located on the licensee's site on the east bank of the Gelaware River in Lower Alloways Creek Township, Salem County, New Jersey. The facility is located approximately eight xles southwest cf Saleir.. New Jersey and is described in the PSEG NLclear LLC Final Safe:jr Analysis Report, as supplemented and amended, a d. i;1 the Environmerzal.l,e?ort, as supplemented and amended.
I I
(1) ?SEG Nczlear LLZ. ;:rs:ar.:
- a section I33 of the Act and 10 CFR
?ar:
S C,
Z C posress. s e and opera:e the facility at the absve desig-azed l:car:z:.
- r. Eaiem Cxmry, New jersey, ir. accordance
.,-zt;
- he przze5zres ar.5 ::r.:za:ions set fcrzh in t h i s license;
( 3 ) 2SEG Nuclear L X. ;-Ts;E?T.:
- 3 the Act and 10 CFR Part 7C. tc receive, pcssess ar.2 -se a:
any time special nuclear material as reactor fuel, ir. arzcri&?ze wrth the limitations for storage an:!
amounts reqdired f::
Teaczar operation, as described in :he Final Safety Analysis hegcrt. a s supalemented and amended; Amendment No. %?-5k 13 5 I
3 PSEG Nuclear LLC, pursuant to the Act and 10 CFR Parts 30, 4 0 and 70, to receive, possess, and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; PSEG Nuclear LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction.
to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components ; and PSEG Nuclear LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.
C. This license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
Maximum Power Level PSEG Nuclear LLC is authorized to operate the facility at reactor core power levels not in excess of 3339 megawatts thermal (100 percent rated power1 in accordance with the conditions specified herein.
I Technical Specifications and Environmental Pro.tection Plan I
The Technical Specifications contained in Appendix A, as revised through Amendment No.
, and the Environmental Protection Plan contained in Appendix B. are hereby incorporated into the license.
PSEG Nuclear LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
Inservice Tescinq of Pumps and Valves (Section 3.9.6, SSER No. 4 ) '
This License Condition was satisfied as documented in the letter from W. R. Butler ( " 2 C ) to C. A. McNeill, Jr. (PSE&G) dated December 7, 1987. Accordingly, this condition has been deleted.
- The parenthetical notation following the title of many license conditions denotes the section of the Safety Evaluation Report and/or its supplements wherein the license condition is discussed.
Amendment No. 131 I
L-v
( 4 ) Inservice Inspection (Section 6.6, SER; Sections 5.2.4.3 and 6.6.3, SSER No. 5)
- a.
PSE&G shall submit an inservice inspection program in accordance with 10 CFR 50.55a(g)(4) for staff review by October 11, 1986.
- b.
Pursuant to 10 CFR 50.55a(a)(3) and for the reasons set forth in Sections 5.2.4.3 and 6.6.3 of SSER No. 5, the relief identified in the PSE&G submittal dated November 18, 1985, as revised by the submittal dated January 20, 1986, requesting relief from certain requirements of 10 CFR 50.55a(g) for the preservice inspection program, is granted (5) Solid State Loqic Modules PSEG Nuclear LLC shall continue, for the life of the plant, a I
reliability program to monitor the performance of the Bailey 862 SSLMs installed at Hope Creek Generating Station. This program should obtain reliability data, failure characteristics, and root cause of failure of both safety-related and non-safety-related Bailey 862 SSLMs. The results of the reliability program shall be maintained on-site and made available to the NRC upon request.
(6) Fuel Storaqe and Handlincr (Section 9.1, SSER No. 5)
- a. No more than a total of three ( 3 ) fuel assemblies shall be out of approved shipping containers or fuel assembly storage racks or the reactor at any one time.
- b. The above three (3) fuel assemblies as a group shall maintain a minimum edge-to-edge spacing of twelve (12) inches from the shipping container array an:: the storage rack array.
- c. Fresh Fuel assemblies, when stored in their shipping con-tainers, shall be stacked no more than three (3) containers high.
(7) Fire Protection (Section 9.5.1.8, SSER NO. 5; Section 9.5.1, SSER No. 6 1 I
PSEG Nuclear LLC,na!l implement and maintain in effect all provisions of the approved fire prot.ection program as described in the FiMl Safety Analysis Report f o r the facility through Amendment No. 15 and as described in its submittal dated May 13, 1986, and as approved in the SER dated October 1984 (and Supplements 1 through 6) subject to the following provision:
PSEG Nuclear LLC may make changes to the approved fire I
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.
Amendment NO. H, 129 I
(8) Solid Waste Process Control Proqram (Section 11.4.2, SER; Section 11.4, sSER NO. 41 PSEG Nuclear shall obtain NRC approval of the Class B and C solid waste process control program prior to processing Class B and C solid wastes.
I (9) Eheruency Planninq (Section 13.3, SSER No. 5 )
In the event that the NRC finds that the lack of progress in completion of the procedures in the Federal Emergency Management Agency's final rule, 44 CFR Part 350, is an indication that a major substantive problem exists in achieving or maintaining an adequate state of emergency preparedness, the provisions of 10 CFR Section 50.54(s) ( 2 ) will apply.
(10) Initial Startup Test Proqram (Section 14, SSER No. 5)
Any changes to the Initial Startup Test Program described in Section 14 of the FSAR made in accordance with the provisions of 10 CFR 50.59 shall be reported in accordance with 50.59(b) within one month of such change.
(11) Partial Feedwater Heating (Section 15.1, SER; Section 15.1, SSER No. 5; Section 15.1, SSER No. 6 )
The facility shall not be operated with reduced feedwater temperature for the purpose of extending the normal fuel cycle.
After the first operating cycle, the facility shall not be operated with a feedwater heating capacity that would result in a rated power feedwater temperature less than 400°F unless analyses supporting such operation are submitted by the licensee and approved by the staff.
( 1 2 ) Detailed Control Room Desiqn Review (Section 18.1, SSER No. 5)
- a. PSE&G shall submit for staff review Detailed Control Room Design Review Summary Reports I1 and I11 on a schedule consistent with, and with contents as specified in, its letter of January 9, 1986.
- b. Prior to exceeding five perzent power, PSE&G shall provide temporary zone markings on safety-related instruments in the control room.
Amendment No.f 29 (13) safety Parameter DiSDlaY Svstem (Section i 8. 2, SSER NO. 5 )
Prior to the earlier of 91) days after restart frcm the first refueling outage or July 12, 1988, ?SE&G shall add the following parameters to the SPDS and have them operational:
- a. Primary containment radia:ion
- b. Primary containment isolation status
- c. Combustible gas concentration ir, primary ccncainment
- d. Source range neutron flux
( 1 4 )
Additional Conditions The Additional Conditions contained in Appendix C, as revised through Amendment No. 135,.are hereby incorporated into this license.
accordance with the Additional Conditions.
1 PSEG Nuclear LLC shall operate the facility in (15) PSE&G to PSEG Nuclear LLC License Transfer Conditions
- a. PSEG Nuclear LLC shall take all necessary steps to ensure that the decommissioning trust is maintained in accordance with the applicarion, the requirements of the Order Approving Trarsfer of License and Conforming Amendment, dated February 16, 2000, and the related Safety Evaluation dated February 16, 2 D O G.
The deccm-nisslcz:ng :
s t
agreement shall provide thac:
I )
The !JSP :f asfe:s LT! bat!? the qualified and am-qcaLi2iei Z x 5 s shal; be linited tc expenses related tc, decc?z.:sri=n:zT cf :Re u n i ~
as defined by the NRC in i t s reqA:afls?.s and issuancss, and as provided in the uniz's ::censs a d any anon&nencs thereto. Siowever, z ~ o n cc.r.;:efi:z c f 2eczrr.~ssi;~ing, as defined above,
- he ass2:s ray be usec! f o r any purpose au:horized by
- ax.
- nves'"s-'-
.- -i
--.e ssc~r::~es 2,'
3~h2r sbligaticns sf 2 )
PSE&G c r affillazas chereof, or their successors c r assigzs. sr.a::
ba 2rohibited. In addition, excep: fcr inves:meczs
- Lei f3 narket indexes cr o:her non-nuclear sector :.c:xa:
C ; L ? ~ S,
investments in any entity owning one C Y mcre r;:z,ear power p1an:s shall be prohibited.
3 ) No disbursemenfs or payments from the trust shall be made by zke trxszee uti: the trustee has first given
- he XRC 3 C days riotice of the payment. In addition, nc dishrsemenrs cr payments from the trust shall be made if :he trx:ee receives pricr written notice of objecticz from the Director, Office of Nuclear Reactcr Regulatioc.
I Amendment No. I:,
2 7, IS;,
1 2 k 135 41 The trust agreement shall not be modified in any material respect without prior written notification to the Director, Office of Nuclear Reactor Regulation.
- 5)
The trustee, investment advisor, or anyone else directing the investments made in the trust shall adhere to a prudent investor standard, as specified in 18 CFR 35.32(3) of the Federal Energy Regulatory Commissions regulations.
- c. PSEG Nuclear LLC shall not take any action that would cause PSEG Power LLC or its parent companies to void, cancel, or diminish the commitment to fund an extended plant shutdown as represented in the application for approval of the transfer of this license from PSE&G to PSEG Nuclear LLC.
D. The facility requires exemptions from certain requirements of 10 CFR Part 50 and 10 CFR Part 70. An exemption from the criticality alarm requirements of 10 CFR 70.24 was granted in Special Nuclear Material License No. 1953, dated August 21, 1985. This exemption is described in Section 9.1 of Supplement No. 5 to the SER. This previously granted exemption is continued in this operating license.
exemption from certain requirements of Appendix A to 10 CFR Part 50, is described in Supplement No. 5 to the SER.
schedular exemption to the requirements of General Design Criterion 64, permitting delaying functionality of the Turbine Building Circulating Water System-Radiation Monitoring System until 5 percent power for local indicatlon, and until 120-days after fuel load or control room indication (Appendix R of SSER 5). Exemptions from certain requirements of Appendix J to 10 CFR Part 50, are described in Supplement No. 5 to the SER.
requirement of Appendix J, exempting main steam isolation valve leak-race testing at 1.10 Pa (Section 6.2.6 of SSER 5 ) ; an exemption from Appendix J, exempting Type C testing on traversing incore probe system shear valves (Section 6.2.6 of SSER 5 ) ; an exemption from Appendix J, exempting Type C testing for instrument lines and lines containing excess flow check valves (Section 6.2.6 of SSER 5 ) ; and an exemption from Appendix J, exempting Type C testing of thermal relief valves (Section 6.2.6 of SSER 5 ). These exemptions are authorized by law, will not present an undue risk to tha public health and safety, and are consistent with the camon defense and security. These exemptions are hereby granted. The special circumstances regarding each exemption are identified in the referencad section of the safety evaluation report and the supplements thereto.
granted pursuant to 10 CFR 50.12.
will operate, to the extent authorized herein, in conformity with the application, as amended, the provisions of the Act, and the rules and regulations of the Commission.
An This exemption is a Those include an exemption from the These exemptions are With these exemptions, the facility Amendment NO. W, 129 I
.U' 8
E. The licensee shall fully implement and maintain in effec: all provisims of the CsmLission-apprwed physical security, gcarcl training and qualification, and safeguards contingency plans including amendments made pursuat ts provisions of =he MistelLar.e:u Amendments and search Requirements revisions to 10 CFR 73.55 ( 5 i CR 27817 and 27822) and ta the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The plans, which contain Safeguards Infarmation prcrecced under 10 CFR 73.21, are entitled: "Salem-Hope Creek Nuclear Generating Station Security Plan," with revisions submitted thrsugh Security Training and Qualification Plan,' with revisions submitted through December 17, 2001; and 'Salem-Hope Creek Nuclear Generacizg Station Security Contingency Plan," with revisions submitted thrzugk June 2, 1998. Changes made in accordance with 10 CFR 7 3. 5 5 shall be implemented in acccrclance with the schedule set forth therein.
December 17, 2001; 'Salem-Hope Creek Nuclear Generating Staticn
- t. Except as otherwise prsvided in the Technical Specificacisns 3r Environmental Protecclsn Tlan. PSEG Nuclear LLC shall report any viciations of the reqdirements contained in Section 2. C cf this within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> t 3 the XRC Cperations Center via the kergency Kccification System xifh written followup within thirty days in atzzrciance with the przcedures described in 10 CFR 50.73(b), (c), ard (e).
r,,ense 7 i r
in the foll3wing manner: initial notification shall be made.
s. The lLcensees shall kave &?d maintain financial prctection of such
=:.?e and in such &mz'-'.zs as :;?e Cxrzissia= shall require i?.
accsrclance with Sect::r.
- ?:
zf the Azomic Energy Act = f 1954, as
&Tended, to ccver p ~ t l r z ::&ility claims.
I NUCLEAR REGU WTSR Y
- zrlginal signed by H.R: 3enton -
Harcld 8. Zenton, Directar Cffrce 3: Nuclear Reactor Regulacion fz=::sxres:
Acpendix A - Technlca: S;e::f:ra::ons (NUREG-1202)
.;p?endix B - Em?ircnner.za: ?r;:ezzicn Plan Amendment No. 5!9,138 I
3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 SHUTDOWN MARGIN LIMITING CONDITION FOR OPERATION 3.1.1 The SHUTDOWN MARGIN shall be equal to o r greater than:
- a.
0.38% delta k/k with the highest worth rod analytically determined, o r
- b.
0.28% delta Wk with the highest worth rod determined by test.
APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3, 4 and 5.
ACTION:
With the SHUTDOWN MARGIN less than specified:
- a.
In OPERATIONAL CONDITION 1 or 2, reestablish the required SHUTDOWN MARGIN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or be in at least HOT SHUTOOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
- b.
In OPERATIONAL CONDITION 3 or 4, inmediately verify all insertable control rods to be inserted and suspend all activities that could reduce the SHUTDOWN MARGIN. In OPERATIONAL CONDITION 4, establ ish SECONDARY CONTAINHENT INTEGRITY within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
- c.
In OPERATIONAL CONDITION 5, suspend CORE ALTERATIONS and other activities that could reduce the SHUTOOWN MARGIN and insert all insertable control rods within I hour.
Establish SECONDARY CONTAIN-MENT INTEGRITY within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
SURVEILLANCE REQUIREHENTS 4.1.1 The SHUTDOWN MARGIN shall be detemined to be equal to or greater than specified at any time during the fuel cycle:
- a.
- b.
By measurement, prior to or during the first startup after each refueling.
By measurement, within 500 MUD/T prior to the core average exposure at which the predicted SHUTOOWN MARGIN, including uncertainties and calculation biases, is equal to the specified limit,
- c.
Within I2 hours after detection of a withdrawn control rod that is hmovable, as a result of excessive friction or mechanical inter-ference, or is untrippable, except that the above required SHUTDOWN MARGIN shall be verified acceptable with an increased allowance f o r the withdrawn worth of the immovable or untrippable control rod.
HOPE CREEK 3/4 1-1
REACTIVITY CONTROL SYSTEMS 3/4.1.2 REACTIVITY ANOMALIES LIMITXNG CONDITION FOR OPERATION 3.1.2 The reactivity equivalence of the difference between the actual ROD DENSITY and the predicted ROD DENSITY shall not exceed 1% delta k/k.
APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.
With the reactivity equivalence difference exceeding 1% delta k/k:
- a.
Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> perform an analysis to determine and explain the cause of the reactivity difference; operation may continue if the difference is explained and corrected.
- b.
Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.1.2 The reactivity equivalence of the difference between the actual ROD DENSITY and the predicted ROD DENSITY shall be verified to be less than or equal to 1% delta k/k:
- a.
During the first startup following CORE ALTERATIONS, and
- b.
At least once per 31 effective full power days during POWER OPERATION.
HOPE CREEK 3/4 1-2
REACTIVITY CONTROL SYSTEMS 3/4.1.3 CONTROL RODS CONTROL ROO OPERABILITY LIMITING CONDITION FOR OPERATION v
3.1.3.1 A l l control rods shall be OPERABLE.
APPLICABILITY:
OPERATIONAL CONDITIONS 1 and 2.
ACTION:
- a.
With one control rod inoperable due t o baing inrmovable, as a result of excessive f r i c t f o n or mechanical interference, or known to be untrippable:
- 1.
Wfthfn one hour:
a)
Verify that the inoperable control rod, i f withdrawn, i s separated from a l l other inoperable control rods by a t least two control cells i n a l l directions.
Disarn the associated directional control valves** hydraulically by closing the drive water and exhaust water isolation valves.
b)
Otherwise, be in a t least HOT SHUTOMJN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
Restore the inoperable control rod, i f withdrawn, to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be f n a t least HOT SHUTDOVN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
- 2.
- b.
With one or nore control rods ttippable but inoperable for causes other than addressed i n ACTION a, above:
- 1.
If the inoprrable control rod(s) i s withdram, within one hour:
a)
V e r i f y that the inoptrablt withdrawn control rod(s) i s separated from a11 other inoperable withdrawn control rods by a t l e a s t two control cells in a l l directions, and Dtcwnrtrate the insertion capability of the Inoperable withdrawn control rod(s) by inserting the control rod(s) a t least one notch by drive water pressure within the n o m 1 operating range*
b)
Othcrwf sc, i nrert the i nopcrabl e wf thdraun control rod( s) and d i Sam the associated directional control valves** either:
a)
Electrically, or b)
Hydraulically by closing the drive water and exhaust water isolation valves.
- The inoperable control rod may then be withdrawn t o a position no further withdrawn than I t s positlon when found t o be inoperable.
- May be rearmed intermittently, under administratlvc control, t o permit testing associated with restoring the control rod t o OPERABLE status.
L HOPE CREEK 3/4 1-3 c
(Continued)
Othmrwise, k in a t l08.t BOT gEttrrodwN w i t h i n thm nuct 12 bour8.
- 3.
The provimLon8 of Spmclfic8tion 3.0.4 arm not applfcablo.
C.
- d.
0.
With more thur 8 control coda i a o p r r a b l m, bo i n at leut fwrr SEUTDWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
With ono 8er.m dimchug8 rolum8 raat v8ltn and/or one scrm d s c h u p.
oolum8 drain valoo inoperable mad opmn, rostora thm iaoprr8bl8 valv~[r) to OPE-mtatus within 24 burn or b. i n 8t h8.t HOT SIIDTOOWW within th. next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
With any scram dischuge rolunm vent valve(8) and/or m y.cram dfachuge velum0 drain valvo(m) oth.rviso inoporabl8, ro8torm the inoperable valvo(s) t o OP-status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or k i a at least liOT SHUTDOW wit-tha amxt 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
4.1.3.1.1 demorutrat86 OPERULX by:
th. scram discharge volum8 &.in and vmnt valve..hall be
- b.
A t h a r t oace par 24 hour8 wmrifying mach velio t o b. op.n,* 8nd
- b.
A t loaat onco p r 31 day. cyclfng mach vdvm through 8t hut one cauplote cycle of f u l l +ravel.
4.1.3.1.2 wlthdr8un control rotla not rmqulrod to ha*, m o l t dlroctional control o l l v r r df8uacld Whin above the lou pouu 8epoint of tho RWX, all I
- Thoma valor@ may bo cloaod intermittently for tmrttng Und8r rdmlnistretiw control#.
- m y ba toannod intermlttmntly, u n d u a&ninirtratlve'eontrol, to p o d t t o s t h g r a ~ o c l r t o d with tortoring the control rod to OPERLBIZ mt8tu..
\\.-
EOPE CREKK 314 1-4 madmen+ I O ~.
105
SURVEXLLANCE R E Q U I ~ S (Continued) electrically or hydraulically shall be demonatr8ted OPERABLE by moving eacn control rod at leust one notch:
- a.
A t leaat once per 7 days, a d
- b.
I within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when any control rod is imnovable a8 a result of utcesrive friction or mechanical interference.
4.1.3.1.3 Surveillance Repuircmcnta 1.1.3.2, 4.X.3.4.
4.1.3.5, 4. L. 3. 6 axad 4.1.3.7.
9.1.3.1.4 demonstrating:
All control rods shall be demdnstratad OPERABLE by performance of The scram discharge volume shall be determined OOERABLB by
- a.
The scram discharge volume drain 8ad vaat valve.
ORERABLZ at learnt once per 18 month., by varifying that the & a b axad vent valvea:
- 1.
Close wfthfn 30 seconb after receipt of a a i m 1 for control rod. to scram, m d
- 2.
open when the scram signal is reset.
1 ROPE CREEK 3/4 1-5 Amendment No.811
REACTIVITY CONTROL SYSTEMS CONTROL ROD MAXfHUM SCRAM INSERTION TIMES LIMITING CONDITION FOR OPERATION 3.1.3.2 The maximum scram insertion time o f each control rod from the fully withdrawn position to notch position 5, based on de-energiration of the scrim pilot valve solenoids as time zero, shall not exceed 7.0 seconds.
APPLICABILITY:
OPERATIONAL CONDITIONS 1 and 2.
ACTION: -
- a.
With the maximum scram insertion time o f one or more control rods exceedf ng 7.0 seconds:
- 1.
Declare the control rod(s) with the slow insertion time Inoperable, and
- 2.
Perform the Surveillance Requirements o f Specification 4.1.3.2.c at least once per 60 days when oprratlon i s continued with three or more control rods with naudrnum scram insertion times in excess o f 7.0 seconds.
Otherwise, be i n at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
The provisfons o f Specffication 3.0.4 are not applicable.
- b.
SURVEIL LANCE REQUIREMENTS 4.1.3.2 The rnaxfmtnn scram insertion time o f the control rods shall be dernon-strated through measurement with reactor coolant pressure greater than or equal to 950 psig and, durfng single control rod scram time tests, the control rod drive pumps Isolated from t h e accumulators:
- a.
For all control rods prior to THEW1 POWER exceeding 40% of RATED THERHAL P O W ? following CORE ALTERATIONS or after a reactor shutdown that I s greater than 120 days.
- b.
For specifically affected fndividual control rods following malntenancc on or modificrtfon to the control rad or control rod drive ryrtea which could affect the scram Inscrtfon tine of those Specific control rods, and
- c.
For at least 10% of the control rods, on a rotating basis, at least once per 120 days of POWER OPERATION.
v HOPE CREEK 3/1 1-6
REACTIVITY CONTROL SYSTEMS CONTROL ROD AVERAGE SCRAM INSERTION TIMES LIMITING CONDITION FOR OPERATION 3.1.3.3 The average scram insertion time o f all OPERABLE control rods from
~
the fully withdrawn position, based on de-energization of the scram pilot.6alve solenoids as time zero, shall not exceed any o f the following:
Position Inserted From Ful ly Withdrawn Average Scram Inser-tion Time (Seconds) 45 39 25 05 0.43 0.86 1.93 3.49 APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.
ACT ION:
With the average scram insertion time exceeding any o f the above limits, be i n at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.1.3.3 All control rods shall be demonstrated OPERABLE by scram time testing from the fully withdrawn position as required by Surveillance Requirement 4.1.3.2..
HOPE CREEK 3/4 1-7
REACTIVITY CONTROL SYSTEMS FOUR CONTROL ROD GROUP SCRAM INSERTION TIMES LIMITING CONDITION FOR OPERATION 3.1.3.4 The average scram insertion time, from the fully withdrawn position, for the three fastest control rods in each group of four control rods arranged in a two-by-two array, based on deenergization of the scram pilot valve solenoids as time zero, shall not exceed any of the following:
Position Inserted From Ful ly Withdrawn Average Scram Inser-tion Time (Seconds) 45 39 25 05 APPLICABILITY:
OPERATIONAL CONDITIONS 1 and 2.
A C T 1 ON:
0.45 0.92 2.05 3.70
- a.
With the average scram insertion times o f control rods exceeding the above limits:
- 1.
Declare the control rods with the slower than average scram insertion times inoperable until an analysis is performed to determine that required scram reactivity remains for the slow four control rod group, and
- 2.
Perform the Surveillance Requirements o f Specification 4.1.3.2. c at least once per 60 days when operation i s continued with an average scram insertion time(s) in excess o f the average scram insertion time limit.
Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
The provisions o f Specification 3.0.4 are not applicable.
- b.
SURVEILLANCE REQUIREMENTS 4.1.3.4 from the fully withdrawn position as required by Surveillance Requirement 4.1.3.2.
All control rods shall be demonstrated OPERABLE by scram time testing HOPE CREEK 3/4 1-8
REACTIVITY CONTROL SYSTEMS CONTROL ROD SCRAM ACCUMULATORS LIMITING CONDITION FOR OPERATION zx 3.1.3.5 Each c o n t r o l rod scram accumulator s h a l l be OPERABLE.
I APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 5*.
a.
I n OPERATIONAL CONDITIONS 1 o r 2 :
I
- 1.
With one c o n t r o l rod scram accumulator inoperable and r e a c t o r pressure 2 900 psig, w i t h i n 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, V : e
- c s s t tr.e acccmulaccr ass3z:a:ez w i t h each withdrawn c o n t r o l 3. Y. i i. Z.
a m -.
, - l i c a b l e if a l l inoperable control rod scram a::::
- : a t z r s a r e associa:ed
.n*lt.", f z l i y i n s e r t e d contra1 rods.
..I.. _ _
I. - -
aspiicable cs concrz:
r x s :moved p e r S p e c i f i c a t i o n 3. 9. 1 0. 1 o r
~-
a )
Restore t h e inoperable accumulator t o OPERABLE s t a t u s,
o r b )
I n s e r t t h e associated c o n t r o l rod, d e c l a r e t h e a s s o c i a t e d c o n t r o l rod inoperable and disarm t h e associated control valves e i t h e r e l e c t r i c a l l y or h y d r a u l i c a l l y by c l o s i n g t h e d r i v e water and exhaust water i s o l a t i o n valves.
O t h e r w i s e, be i n a t least HOT SHUTDOWN with t h e next 1 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
L.
3 With two o r y o r e c o n t r o l rod scram accumulators inoperable and r e a c t o r pressure 2 900 p s i g,
a )
W i t h i n 20 r i n u t e s of discovery of t h i s condition concurrent with charging water pressure < 940 psig, r e s t o r e charging water header pressure t o 2 940 p s i g otherwise p l a c e t h e mode switch i n t h e shutdown p o s i t i o n *
- l and b)
W i t h i n one h o u r insert t h e a s s o c i a t e d c o n t r o l radsl o e c l a r e :.?e associated c o n t r o l rods inoperable azd d-sarrr tr.e a s s s c i a t e d c o n t r c l valves e i r h e r elec:r:cal:y 2: h y d r a u l i c a l l y b y c l o s i n g t h e c r l v e wa:er ar.3 e x h a ~ s : water i s o l a t i o n valves.
O t h e r w i s e, be I::
d f l e a s t HOT SHUTDOWN within :he n e x t 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
3 / 4 1-9 Amendment N0.98
REACTIVITY CONTROL SYSTEMS LIMITING CONDITION FOR OPERATION (continued)
ACTION (Continued)
- 3.
With one o r more c o n t r o l rod scram accumulators i n o p e r a b l e and r e a c t o r p r e s s u r e < 990 p s i g,
a )
Immediately upon discovery of charging water header p r e s s u r e < 940 p s i g, v e r i f y a l l c o n t r o l rods a s s o c i a t e d with inoperable accumulators are f u l l y i n s e r t e d otherwise p l a c e t h e mode switch i n t h e shutdown position**, and b )
Within one hour i n s e r t t h e a s s o c i a t e d c o n t r o l r o d ( s ),
d e c l a r e t h e a s s o c i a t e d c o n t r o l r o d ( s ) i n o p e r a b l e and disarm t h e a s s o c i a t e d c o n t r o l valves e i t h e r e l e c t r i c a l l y o r h y d r a u l i c a l l y by c l o s i n g t h e d r i v e water and exhaust water i s o l a t i o n valves.
Otherwise, be i n a t least HOT SHUTDOWN within t h e next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
- b.
I n 0PZ:RATIONAL CONDITION 5*:
- 1.
With one or more withdrawn c o n t r o l rods inoperable, upon a i sco~'.: r y immediately i n i t i a t e a c t i o n t o f u l l y i n s e r t i n o p e r a b l e withdrawn c o n t r o l rods.
C.
The provisions of S p e c i f i c a t i o n 3. 0. 4 a r e n o t a p p l i c a b l e.
v 2. l. 3. 5 Each c o n t r o l rod scram accumulator s h a l l be determined OPERABLE:
a.
A:
leas:
-rice p e r 7 days by v e r i f y i n g t h a t t h e i n d i c a t e d p r e s s u r e 1s qreazer than o r e q s a l t o 940 p s i g u n l e s s t h e c o n t r o l rod i s
- r.sezted and disarmed o r scrammed.
I L - - L.:z:
- .-.E' d=f'J"
...- l a t o r ass3c;a:ez x r z h each withdrawn c o n t r o l
.._.. I _ C
.._- *;;::=able t o c o n t r o l r 3 3 s rer.3ved p e r S p e c i f i c a t i o n 3.9.10.1 o r
~g;iicaSle if a l l inoperable c s z t r s l rod scram U : Z. - :, ~ ~ : C : S a r e a s s z c i a t e d w i t h f u l l y l n s e r t e d c o n t r o l rods.
3 / 4 1-10 Amendment No. 98
3.1.3.6 All control rods shall be coupled to their drive mechanisms.
APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 5 f ACTION:
- a. In OPERATIONAL CONDITION 1 and 2 with one control rod not coupled to its associated drive mechanism, within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s:
I
- 1.
If permitted by the RWM,, insert the control rod to accomplish recoupling and verify recoupling by withdrawing the control rod, and:
a)
Observing any indicated response of the nuclear b)
Demonstrating that the control rod will not go to the instrumentation, and overtravel position.
2.
If recoupling is not accomplished on the first attempt or, if I
not permitted by the RWM, then until permitted by the RWM, declare the control rod inoperable, insert the control rod and disarm the associated directional control valves** either:
a)
Electrically, or b)
Hydraulically by closing the drive water and exhaust water isolation valves.
Otherwise, be in at least HOT SHVTWWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
- b. In OPERATIONAL CONDITION S* with a withdrawn control rod not coupled to its associated drive mechanism, within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> either:
- 1.
Insert the control rod to accomplish recoupling and verify recoupling by withdrawing the control rod and demonstrating that the control rod will not go to the overtravel position, or 4..
If recoupling is not accomplished, insert the control rod and disarm the associated directional control valves** either:
a)
Electrically, or b)
Hydraulically by closing the drive water and exhaust water isolation valvee.
- c. The provisions of Specification 3. 0. 4 are notapplicable.
.kt least each withdrawn control rod.
Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.
- May be rearmed intermittently, under administrative control, to permit testing associated with restoring the control rod to OPERABLE status.
HOPE CREEK 314 1-11 Amendment NO. 105
REACTIVITY CONTROL SYSTEMS SUQVEILLANCE REQUIREMENTS 4.1.3.6 Each affected control rod shall be demonstrated t o be coupled to its drive mechanism by observing any indicated response o f the nuclear instrumen-tation while withdrawing the control rod to the fully withdrawn position and then verifying that the control rod drive does not go to the overtravel position:
- a.
P r i o r to reactor criticality after completing CORE ALTERATIONS t h a t could have affected the control rod drive coupling integrity,
- b.
Anytime the control rod is withdrawn t o the "Full out" position i n subsequent operation, and
- c.
Following maintenance on or modification to the control rod o r control rod drive system which could have affected the control rod drive coupling integrity.
HOPE CREEK 3/4 1-12..
PACTIVITY CO NTROL SYSTEMS CONTROL R OD POSITION 1M)ICATION LIMITING CONDITION FOR OPERATION l
l l
I l
l I
I 1
I I
l l
I I
=
=
I I
=
~
~
~
~
~
=
=
~
~
=
~
~
~
~
=
=
~
~
m
=
=
~
~
~
~
~
~
3.1.3.7 The control rod position indication system shall be OPERABLE.
BPPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 5'.
&XION:
- a.
In OPERATIONAL CONDITION 1 or 2 with one or more control rod position indicators inoperable, within 1 hour:
- 1.
2.
3.
Determine the poeition of the control rod by using an alternate method, or:
A ) Moving the control rod, by single notch movement, to A b) Returning the control rod, by single notch movement, to its c) Verifying no control rod drift alarm at leart once per 12 position with an OPERABLE position indicator, original position, And houra, or Hove the control rod to a position with an OPERABLE position indicator, or When THERMAL POWER is:
A ) Within the preset power level of the RWH, declare the control rod inoperable.
b ) Greater than the preset power level. of the RWH, declare the control rod inoperable, insert the control rod and disarm the aseociated directional control valves** either:
- 1)
Electrically, O r
- 2)
Hydraulically by closing the drive water and exhaust water isolation valves.
Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
- b.
In OPERATIONAL CONDITION 5
- with a withdrawn control rod poeition indicator inoperablo, move tho control rod to A position with an OPEFLABLE position indicator or insert the control rod.
- c.
The provisionr of Specification 3.0.4 are not applicable.
- A t least each withdrawn control rod.
Not applicable to Control rods removed per Specification 3.9.10.1 or 3.9.10.2.
- May be rearmed intermittently, under administrative control, to permit testing aseociated with restoring the control rod to OPERABLE statu..
HOPE CREEK 314 1-13 Amendment NO. 105
'L REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS 4.1.3.7 The control rod position indication system shall be determined OPERABLE by verifying:
- a.
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> that the pos indicated, tion o f each control rod i s b.
That the indicated control rod position changes during the movement o f the control rod drive when performing Surveillance Requirement 4.1.3.1.2, and
- c.
That the control rod position indicator corresponds t o the control rod position indicated by the "Full Out" position indicator when performing Survei 1 lance Requirement 4.1.3.6. b.
1-HOPE CREEK 314 1-14
REACTTVIfV C ~ N T R O L SYSTEMS COtdTROL ROO O R I V E n0USTkG SUPPORT 3..1.3.8 The c o n t r o l rod drive housing support shall be in place.
APPLICA8ILfTV:
ACT 103:
With the ctantro SHUTDOW v i t h i n OPERATIOMAL CONOIflONS 1, 2 and 3.
rod drfve housing support n o t i n place, be In a t 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and I n CULD SHUfDCYN u'lthfn the ;'oI'l%ing SURVEILLANCE REQUIREMWtS 4.1.3.8 The control rod U t f v r houslng support shall be verfffrd t o be i n p l a c e by a vfsurl fnsp8ctlon prfor fo Startup 8 y tlw I t has been dlsars.crbled or when wfntrlnrncr has been performed I n UH control rod dtlva houslng s u p p o r t area.
HOPE CREEK 3/4 1-19
-I
REACTIVITY CONTROL SYSTEMS u4.1.4 CONTROL ROD PROGRAM CONTROLS
'V W
ROD WORTH MINIMIZER 3.1.4.1 The Rod worth minimizer (RWM) shall be OPERABLE.
APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2**, when THERMAL POWER is less than or equal to 10% of RATED TRERMAL POWER, minimum allowable low power setpoint.
I ACTION:
I
- a.
With the RWM inoperable after the first 12 control rods are fully withdrawn, operation may continue provided that control rod movement and compliance with the prescribed control rod pattern are verified by a second licensed operator or other technically qualified member of the unit technical staff who is present at the reactor control conaole.
- b.
with the RWM inoperable before the first twelve (12) control rods are fully withdrawn, one startup per calendar year may be performed provided that the control rod movement and compliance with the prescribed control rod pattern are verified by a second licensed operator or other technically qualified member of the unit technical staff who is preecnt at the reactor control console.
C.
Otherwise, control rod movement may be only by actuating the manual acram or placing the reactor mode ewitch in the Shutdown position.
4.1.4.1 The RIM ehall be demon8trated OPERABLE:
- a.
In OPERATIONAL CONDITION 2 within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> prior to withdrawal of control rode for the purpoee of making the reactor critical, and in OPERATIONAL CONDITION 1 within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> prior to RWM automatic initiation when reducing T"U POUER, by verifying proper indication of the selection error of at least one out-of-eequence control rod.
Entry into OPERATIONAL CONDITION 2 and withdrawal of eelected control rode is permitted for the purpoee of determining the OPERABILITY of the RWM prior to withdrawal of control rods for the purpose of bringing the reactor to criticality.
- See Special Test Exception 3.10.2.
HOPE CREEK 3/4 1-16 Amendment No. 1Oj
REACTIVITY CC"RoL SYSTEMS 3/4.1.4 CONTROL ROD PROGWlM CONTROLS ROD WORTH MINIMIZER
- b.
In OPERATIONU CONDITION 2 within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> prior to withdrawal of control rods for the purpose of making the reactor critical, verifying the rod block function by demonstrating inability to withdraw an out-of-sequence control rod.
by C.
d.
In OPERATIONAL CONDITION 1 within one hour after R W automatic initiation when reducing THERMAL POWER, by verifying the rod block function by demonstrating inability to withdraw an out-of-sequence concrol rod.
By verifying that the control rod patterns and sequence input to the RWM computer are correctly loaded following any loading of the program into the computer.
HOPE CREEK 314 1-16a I
Amendment No. 105
THIS P.1GE [STE\\TIOV.4LLI. LEFT BL\\.\\K
7-C ROD SEOUENCE CONTROL SYSTEM The material originally contained in Section 3 / 4. 1. 4. 2 was deleted with the issuance of Amendment No.
. However, to maintain numerical continuity between the succeeding sections and existing station procedural references to those Technical Specification sections, 3/4.1.4.2 has been intentionally left
- blank, HOPE CREEK 3 / 4 1-17 Amendment NO. 105
'v R E A C T I V I T Y CONTROL SYSTEMS ROD BLOCK MONITOR L I M I T I N G CONDITION FOR OPERATION 3.1.4.3 Both rod block monitor (RBM) channels shall be OPERABLE.
APPLICABILITY:
equal t o 30% of RATED THERMAL POWER.
OPERATIONAL CONDITION 1, when THERMAL POWER i s greater than o r ACTION:
- a.
With one RBM channel inoperable:
- 1.
V e r i f y t h a t the reactor i s not operating on a LIMITING CONTROL ROD PATTERN, and
- 2.
Restore the inoperable RBM channel t o OPERABLE status w i t h i n 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Otherwise, place the inoperable rod block monitor channel i n the tripped condition w i t h i n the next hour.
- b.
With both RBM channels inoperable, place a t least one inoperable rod block monitor channel i n the tripped condition w i t h i n one hour.
SURVEILLANCE REQUIREMENTS 4. 1. 4. 3 Each of the above required RBM channels shall be demonstrated OPERABLE by performance o f a:
- a.
CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION a t the frequencies and for the OPERATIONAL CONDITIONS specified in Table 4.3.6-1.
b.
CHANNEL FUNCTIONAL TEST p r i o r t o control rod withdrawal when the reactor i s operating on a LIMITING CONTROL ROD PATTERN.
HOPE CREEK 3/4 1-18
REACTIVITY CONTROL SYSTEMS 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM LIHITING CONDITION FOR OPERATION 3.1.5 The standby l i q u i d control system consists of two redundant subsystems and shall be OPERABLE.
APPLICABILITY:
OPERATIONAL CONDITIONS 1, 2, and S*
ACTION: -
- a.
Xn OPERATIONAL CONDITION 1 or 2:
- 1.
With one system subsystea inoperable, restore the subsystem t o O P E W L E status within 7 days or be i n a t least HOT SHUTOOW within the next l2 hours.
With both system subsystems inoperable, k s t o r t a t least one subsystem t o OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> o r be i n a t least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
With one system subsystem inoperable, restore subsystem t o OPERABLE status within 30 days o r insert a l l insertable control rods within the next hour.
With both standby l i q u i d control system subsystems inoperable, Insert a l l insertable contra? rods within one hour.
- 2.
- b.
I n OPERATIONAL CONDITION 5*:
- 1.
- 2.
SURVEILLANCE REQUIREMENTS 4.1.5 Tho standby l i q u i d control system shall be demonstrated OPERABLE:
- 8.
A t l e r s t once-pr). 2 1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> by verififng that;
- 1.
The t m e r r t u n of the s o d i u pentaborate solution i n the storage tank i s greater than o r equal to 7OOF.
- 2.
- 3.
The available volume o f s o d i u pentaborate solution i s within the l i m l t s o f Figure 3.1.5-1.
The M a t tracing c i r c u i t f r OPERABLE by d8temining the teaperrture o f tho pmp suction piping t o bo greater than o r equal t o 70OF.
'*with any control rod withdrawn.
Not appltcablo to control rods r o v e d pet Specification 3.9.10.1 or 3.9.10.2.
HOPE CREEK 3/4 1-19
C' REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
, :.... I. i..
.;..*.i
- b.
At least once per 31 days by:
- 1.
Verifying the continuity o f the explosive charge.
1
- 2.
Determining that the available weight of sodium pentaborate is greater than or equal to 5,776 lbs and the concentration of boron i n solution is within the limits of Figure 3.1.5-1 by chemical analysis.*
- 3.
Verifying that each valve (manual, power operated or automatic) in the flow path that i s not locked, sealed, or otherwise secured in position, is in its correct position.
- c.
Demonstrating that, when tested pursuant to Specification 4.0.5, the minimum flow requirement of 41.2 gpm, per pump, at a pressure of greater than or equal to 1255 psig is met.
- d.
At least once per 18 months during shutdown by:
- 1.
Initiating one of the standby liquid control system subsystem, including an explosive valve, and verifying that a flow path from the pumps to the reactor pressure vessel is available by pumping demineralized water into the reactor vessel and verifying thatathe relief valve does not actuate.
f o r the explosive valve shall be from the same manufactured batch as the one fired or from another batch which has been certified by having one of that batch successfully fired.
injection subsystems shall be tested in 36 months.
The replacement charge Both
- 2.
- Demonstrating that all heat traced piping between the storage tank and the injection pumps is unblocked and then draining and flushing the piping with demineralized water.
- 3.
Demonstrating that the storage tank heaters are OPERABLE by verifying the expected temperature rise o f the sodium penta-
,borate solution in the starage tank after the heiters are ent rg i zed.
"This test shall also be performed anytime water or boron is added to the solution or when the solution temperature drops below 70°F.
been found to be inoperable and may be performed by any series o f sequential overlapping or total flow path steps such that the entire flow path i s included-
- This test shall 8lS0 be performed whenever both heat tracing circuits have HOPE CREEK 314 1-20 Amendment NO. 11 I
Eld, 3 a
4 1
- =
t E-t 3-t I
I I
1 I
I I
8 a
2 W
i a
a 314 1-a
r--_.
i
? -.
I.
... l i
....,. I...
j
- 2::-
- mse, 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1 As a minimum, the reactor protection system instrumentation channels shown in Table 3.3.1-1 shall be OPERABLE.
APPLICABILITY: As shown in Table 3.3.1-1.
ACTION :
- a. With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip System requirement for one trip system, place the inoperable channel(s) and/or that trip system in the tripped condi-tion* within twelve hours. The provisions of Specification 3.0.4 are not applicable.
- b. With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip System requirement for both trip systems, place at least one trip system** in the tripped condition within one hour and take the ACTION required by Table 3.3.1-1.
SURVEILLANCE REQUIREMENTS 4.3.1.1 Each reactor protection system instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.1.1-1.
4.3.1.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 months.
4.3.1.3 The REACTOR PROTECTION SYSTEM RESPONSE TIME of each reactor trip functional unit shall be demonstrated to be within its limit at least once per 18 months. Neutron detectors are exempt from response time testing. For the Reactor Vessel Steam Dome Pressure - High Functional Unit and the Reactor Vessel Water Level - Low, Level 3 Functional Unit, the sensor is eliminated from response time testing for RPS circuits. Each test shall include at least one channel per trip system such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific reactor trip system.
4.3.1.4 The provisions of Specification 4.0.4 are not applicable for entry into OPERATIONAL CONDITION 2 or 3 from OPERATIONAL CONDITION 1 for the Inter-mediate Range Monitors.
- An inoperable channel need not be placed in the tripped condition where this would cause the Trip Function to occur. In these cases, the inoperable channel shall be restored to OPERABLE status within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or the ACTION required by Table 3.3.1-1 for that Trip Function shall be taken.
- If more channels are inoperable in one trip system than in the other, place the trip system with more inoperable channels in the tripped condition, except when this would cause the Trip Function to occur.
HOPE CREEK 3/4 3-1 Amendment No. 101
TABLE 3.3.1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION MINIMUM OPERABLE CHANNELS PER TRIP SYSTEM (a)
APPLICABLE 0
PERAT I ONAL CONDITIONS ACTION 1
2 3
1 2
3 FUNCTIONAL UNIT (b).
Range Monitors Flux - High
- 1. Intermediate
- a.
Neutron 3
2 (d) 2 3, 4 5 (c) 3 2
(d) 2 3, 4 5
- b.
Inoperative (e),
2.
Average Power Range Monitor Neutron Flux - Upscale, Setdown 2
2 (d) 1 2
3 4
4 1
2 3
2 3, 4 5 (c)
- a.
- b.
C.
- d.
Flow Biased Simulated Thermal Power - Upscale 2
1 2
1 Fixed Neutron Flux - Upscale 2
2 2 (d)
Inoperative
- 3. Reactor Vessel Steam Dome Pressure - High 1
2
- 4. Reactor Vessel Water Level - Low, Level 3 1
2
- 5.
Main Steam Line Isolation Valve -
Closure 4
4 1
Amendment No. 39 3/4 3-2 HOPE CREEK
TABLE 3.3.1-1 (Continued)
REACTOR PROTECTION SYSTEM INSTRUMENTATION APPLICABLE MINIMUM OPERATIONAL OPERABLE CHANNELS CONDITIONS PER TRIP SYSTEM (a)
ACTION FUNCTIONAL UNIT
- 6.
This item intentionally blank
- 7.
Dr ywe 11 Pressure - High 1, 2(h) 2 1
- 8.
Scram Discharge Volume Water Level - High 2
2 1
3
- a. Float Switch
- b. Level Transmitter/Trip Unit 2
2 1
3
- 9.
Turbine Stop Valve - Closure 6
- 10.
Turbine Control Valve Fast Closure, Valve Trip System Oil Pressure - Low 6
- 11.
Reactor Mode Switch Shutdown Position 1, 2 3, 4 5
2 2
2 1
7 3
- 12.
Manual Scram 1, 2 3, 4 5
2 2
2 1
8 9
HOPE CREEK 3/4 3-3 Amendment No. 53
ACTION 1 ACTION 2 ACTION 3 ACTION 4 ACTION 5 ACTION 6 ACTION 7 ACTION 8 ACTION 9 TABLE 3.3.1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION ACTION Be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
Verify all insertable control rods to be inserted in the core and lock the reactor mode switch in the Shutdown position within one hour.
Suspend all operations involving CORE ALTERATIONS* and insert all insertable control rods within one hour.
Be in at least STARTUP within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
This ACTION is deleted Initiate a reduction in THERMAL POWER within 15 minutes and reduce turbine first stage pressure to less than the automatic bypass setpoint within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
Verify all insertable control rods to be inserted within one hour.
Lock the reactor mode switch in the Shutdown position within one hour.
Suspend all operations involving CORE ALTERATIONS*,
and insert all insertable control rods and lock the reactor mode switch in the SHUTDOWN position within one hour.
- Except replacement of LPRM strings provided SRM instrumentation is OPERABLE per Specification 3.9.2.
HOPE CREEK 3/4 3-4 Amendment No. 53
TABLE 3.3.1-1 (Continued)
REACTOR PROTECTION SYSTEM INSTRUMENTATION TABLE NOTATIONS A channel may be placed in an inoperable status for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for required surveillance without placing the trip system in the tripped condition provided at least one OPERABLE channel in the same trip system is monitoring that parameter.
This function shall be automatically bypassed when the reactor mode switch is in the Run position.
Unless adequate shutdown margin has been demonstrated per Specification 3.1.1, the "shorting links" shall be removed from the RPS circuitry prior to and during the time any control rod is withdrawn*.
The non-coincident NMS reactor trip function logic is such that all channels go to both trip systems. Therefore, when the "shorting links" are removed, the Minimum OPERABLE Channels Fer the Trip System are 4 APRMS, 6 IRMS and 2 SRMS.
This function is not required to be OPERABLE when the reactor pressure vessel head is removed per Specification 3.10.1.
This function shall be automatically bypassed when the reactor mode switch is not in the Run position.
This function is not required to be OPERABLE when PRIMARY CONTAINMENT INTEGRITY is not required.
With any control rod withdrawn. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.
I I
This function shall be automatically bypassed when turbine first stage pressure is 5 159.7 psig equivalent to THERMAL POWER less than 30% of RATED THERMAL POWER. To allow for instrument accuracy, calibration, and drift, a setpoint of 5 135.7 psig is used.
Also actuates the EOC-RPT system.
- Not required for control rods removed per Specification 3.9.10.1 or 3.9.10.2.
HOPE CREEK 3/4 3-5 Amendment No. 35
Q Z S ELECTRICAL POWER SYSTEMS DISTRIBUTION - SHUTDOWN LIMITING CONDITION FOR OPERATION
~
3.8.3.2 As a minimum, 2 of the 4 channels, one of which shall be channel A o r channel. B, o f the power distribution system shall be energized with:
- a.
A. C. power distribution:
- 1.
Channel A, consisting o f :
a) 4160 volt A.C. switchgear bus b) 480 volt A.C. load centers c) 480 volt A.C. MCCs d )
208/120 volt A.C. distr e) 120 volt A.C. distribut bution panels on panels 2.
Channel B, consisting o f :
a) 4160 volt A.C. switchgear bus b) 480 volt A.C. load centers c) 480 volt A. C. MCCs d) 208/120 volt A.C. distribution panels e) 120 volt A.C. distribution panels 3.
Channel C, consisting o f :
a) 4160 volt A.C. switchgear bus b )
480 volt A.C. load centers c )
480 volt A.C. MCCs d) 208/120 volt A.C. distribution panels 10A401 108410 108450 108212 108411 108451 108553 10Y401( source: 108411) 10Y411(source:10B451) 10Y501(source:10B553) 1AJ481 1YF401(source:1AJ481) 1AJ482 10A402 108420 108460 108222 108421 io8461 108563 10Y402(source:108421) 10Y412( source: 106461) 10Y502( source: 108563) 1BJ481 1YF402(source:1BJ481) 1BJ482 10A403 108430 108470 106232 108431 108471 108573 10Y403(source:10B431) 10Y413(source:10B471) 10Y503(source:108573) 1 t
HOPE CREEK 3/4 8-21
ELECTRICAL POWER SYSTEMS t
- u,:--
LIMITING CONDITION FOR OPERATION (Continued) v e) 120 volt A. C. distribution panels 1CJ481 1CJ482 1YF403(source:1CJ481)
- 4.
Channel D, consisting o f :
a) 4160 volt A. C. switchgear bus 10A404 b) 480 volt A. C. load centers 106440 108480 c) 480 volt A. C. MCCs 108242 108441 108481 106583 d) 208/120 volt A. C. distribution panels 10Y404(source:10B441) 10Y414(source:108481) 1OY504(source:108583) e)
120 volt A.C. distribution panels 1DJA?1 lYF404(source:lDJ481) 1DJ482
- b.
D. C. power distribution:
- 1.
- 2.
- 3.
4.
Channel A, consisting o f :
a) 125 volt D.C. switchgear b) 125 volt D.C. fuse box c) 125 volt D.C. distribution panel Channel 8, consisting of:
a) 125 volt D.C. switchgear b) 125 volt D.C. fuse box c) 125 volt D.C. distribution panel Channel C, consisting o f :
a) 125 volt D.C. switchgear b) 125 volt D.C. fuse boxes c) 125 volt D.C. distribution panel Channel 0, consisting o f :
a) 125 volt D.C. switchgear b) 125 volt D.C. fuse box c) 125 volt D.C. distribution panel 100410 1AD412 1AD417 10D420 1BD412 1BD417 10D430 10D436 1CD412 1CD448 1CD417 10D440 100446 1DD412 1DD448 1DD417 HOPE CREEK 3/4 8-22
4 ELECTRICAL POWER SYSTEMS i,
LIMITING CONDITION FOR OPERATION (Continued)
APPLICABILITY:
OPERATIONAL CONDITIONS 4, 5 and *,
ACTION:
- a.
With less than two channels of the above required A.C. distribution system energized, suspend CORE ALTERATIONS, handling o f irradiated fuel in the secondary containment and operations with a potential for draining the reactor vessel.
With less than two channels o f the above required D.C. distribution system energized, suspend CORE ALTERATIONS, handling o f irradiated fuel in the secondary containment and operations with a potential for draining the reactor vessel.
The provisions o f Specification 3.0.3 are not applicable.
- b.
- c.
SURVEILLANCE REQUIREMENTS 4.8.3.2 At least the above required power distribution system channels shall be determined energized at least once per 7 days by verifying correct breaker/switch alignment and voltage on the busses/MCCs/panels.
L' "When handling irradiated fuel in the secondary containment.
HOPE CREEK
3 9 2 A t least 2 source range monitor. (SRH) channels s h a l l be OPERABLE a*:
- nser:cd to tnr normal operating level with: #u
- a.
Annunciation and continuous visual indication i n the control roop,
- b.
One of the required SRM detectors located in the quadrant where CORE ALTERATIONS are being perfofmod and the other requlrea SRM aeteCr3r Iocated in an adjacent quadrant, and Unless adequate shutdown aargin has been demonstrated per Specifica-tion 3.1.1, the "shorting links" removed from the RPS circuitry prior to and during the time any control rod i s withdrawn.
- c.
u.
During a SPIRAL UNLOAD, the count rate may drop below 3 CDS when the number of assemblies r m a i n i n q in the core drops to sixteen or less.
During a SPIRAL RELOAD, up to four fuel assrmblles nay Be loaded in the four bundle locations (mediately surrounding each o f the four S R h p r i o r to obtaining 3 cos.
loaded, the 3 C
~
S count rite is not required.
j e.
Until these assemblies have Been U i t h the requirements of the above spec4fication not SJtlSfitd, f m e d i a t r l y suspend all operations involving CORE ALTERATlONS and insert all intertaOle control roar.
4.9.2 Each of the above required SRM channels s h a l l be demonstrated OPERABLE By:
- a.
At least once per l.2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s:
- 1.
P e r f o m r n c r of J C M N N E L CHECK, "The use o f specyal movable detectors during CORE ALTERATIONS i n place of the
,are connected to the n o w 1 SRH circultr.
norm1 SRN nuclear detrctors 4s permiss4ble as long as there special detectors "Not required tor control rods rOmV8d per Speclflcatfon 3.9.10.1 and 3.9.10.2.,
fhrra SRN channels shall be OPERA&
for crlticrl shufdovn m r g l n dnonstra-tions. An SRn detlCf0r say be retracted provided a channel indication o f I t least 100 CDS 1 s u i n u i n e a.
nOPE CREEK 3/4 9-3 w n - n t NO. 14 1
i-REFUEL I NG I) P E RAT IONS
?
';a.
SURVEILLANCE REQUIREMENTS (Continued)
- 2.
V e r i f y i n g the detectors are inserted t o the normal operating l e v e l, and 3.
Ouring CORE ALTERATIONS, v e r i f y i n g t h a t the detector of an OPERABLE SRM channel i s located i n the core quadrant where CORE ALTERATIONS are being performed and another i s located i n an adjacent quadrant.
- b.
Performance o f a CHANNEL FUNCTIONAL TEST:
- 1.
- 2.
V e r i f y i n g t h a t the channel count r a t e i s a t l e a s t 3 cps.
- 1.
P r i o r t o contr:l rmt withdrawal, Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior t o the s t a r t o f CORE ALTERATIONS, and A t l e a s t once per 7 days.
I
- c.
- 2.
Prior t o and a t least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during CORE ALTERATIONS***,
and I
- 3.
A t l e a s t once per 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />s***.
Unless adequate shutdown nargin has been demonstrated per Spect f i c a t i o n 3. 1. 1, v e r i f y i n g t h a t the RPS c i r c u i t r y "shorting l i n k s " have been removed, w i t h i n 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> prior t o and a t l e a s t once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during the time any control rod i s withdrawn.**
d.
4 I
""Not required f o r control rods removed per S p e c i f i c a t i o n 3.9.10.1 or 3.9.1:
Except as noted i n Specifications 3.9.2.d and 3.9.2.e.
t *
- HOPE CREEK 3/4 9-4 Amendment No.31 c
REFUELING OPERATIONS L
3/4.9.3 CONTROL ROD POSITION LIMITING CONDITION FOR OPERATION 3.9.3 A l l control rods shall be inserted.*
APPLICABILITY:
OPERATIONAL CONDITION 5, during CORE ALTERATIONS.**
ACT ION:
With a l l control rods not inserted, suspend a l l other CORE ALTERATIONS, except t h a t one control rod may be withdrawn under control o f the reactor mode switch Refuel p o s i t i o n one-rod-out interlock.
SURVEILLANCE REQUIREMENTS 4.9.3 A l l control rods shall be v e r i f i e d t o be inserted, except as above speci f i ed:
- a.
Within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> p r i o r to:
- 1.
The s t a r t o f CORE ALTERATIONS.
- 2.
The withdrawal o f one control rod under the control of the reactor mode switch Refuel p o s i t i o n one-rod-out interlock.
- b.
A t l e a s t once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
Except control rods removed per Speci f i c a t i o n 3.9.10.1 o r 3.9.10.2.
- See Special l e s t Exception 3.10.3.
HOPE CREEK 3/4 9-5
r -
v L-3 / 4 9 - 7 henciment s o. 137
2 3 3-3 t
THIS PAGE INTENTIONALLY DELETED.
HOPE CREEK 314 9-9 Amendment No. 31
I
REFUEL I NG 0 PE RAT IONS 3/4.9.8 WATER LEVEL - REACTOR VESSEL 3.9.8 At least 22 feet 2 inches of water shall be maintained over the top of the reactor pressure vessel flange.
APPLICABILITY:
During handling of fuel assemblies or control rods within the being handled are irradiated or the fuel assemblies seated within the reactor vessel are irradiated.
.reactor pressure vessel while in OPERATIONAL CONDITION 5 when the fuel assemblies ACTION:
With the requirements o f the above specification not satisfied, suspend all operations involving handling of fuel assemblies or control rods within the reactor pressure vessel after placing all fuel assemblies and control rods in a safe condition.
4. 9. 8 The reactor vessel water level shall be determined to be at least its minimum required depth within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> prior to the start o f and at least once p e r 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> during handling of fuel assemblies or control rods within the reactor pressure vessel.
HOPE CREEK 3/4 9-11
REFUELiNG OPERATIONS 3/4.9.9 WATER LEVEL - SPENT FUEL STORAGE POOL LIMITING CONDITION FOR OPERATICjN 3. 9. 9 At least 23 feet of water shall be maintained over the top of irradiated fuel assemblies seated in the spent fuel storage pool racks.
APPLICABILITY: Whenever irradiated fuel assemblies are in the spent fuel storage pool.
ACTION:
With the requirements of the above specification not satisfied, suspend all movement of fuel assemblies and crane operations with loads in the spent fuel storage pool area after placing the fuel assemblies and crane load in a safe condition. The provisions of Specification 3. 0. 3 are not applicable.
SURVEILLANCE REQUIREMENTS 4.9.9 The water level in the spent fuel storage pool shall be determined to be at least at its minimum required depth at least once per 7 days.
c-HOPE CREEK 3/4 9-12
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REFUELING OPERATIONS 3/4.9.10 CONTROL ROD REMOVAL V
SINGLE CONTROL ROD REMOVAL LIMITING CONDITION FOR OPERATION 3.9.10.1 One control rod and/or the associated control rod drive mechanism may be removed from the core and/or reactor pressure vessel provided that a t least the following requirements are satisfied until a control rod and associ-ated control rod drive mechanism are reinstalled and the control rod is fully inserted in the core.
- a.
The reactor mode switch is OPERABLE and locked in the Shutdown posit o r in the Refuel position per Table 1.2 and Specification 3. 9. 1.
on
- b.
- c.
The source range monitors (SRM) are OPERABLE per Specification 3. 9. 2 The SHUTDOWN MARGIN requirements of Specification 3.1.1 are satisfied, except that the control rod selected to be removed;
- 1.
May be assumed to be the highest worth control rod required to be assumed to be fully withdrawn by the SHUTDOWN MARGIN test, and
- 2.
Need not be assumed to be immovable or untrippable.
d.
A l l other control rods in a five-by-five array centered on the control rod being removed are inserted and electrically or hydraulically disarmed or the four fuel assemblies surrounding the control rod or control rod drive mechanism t o be removed from the core and/or reactor vessel are removed from the core cell.
e.
A l l other control rods are inserted.
f A l l fue.1 loading operations shall be suspended.
APPLICABIllTY:
OPERATIONAL CONDITIONS 4 and 5.
AC: I C h.
W i t h the requirements o f the above specification not satisfied, suspend removal o c the control rod and/or associated control rod drive mechanism from the core and/or reactor pressure vessel and initiate action to satisfy the above requ i reme n t s.
HOPE CREEK 3/4 9-13
REFUELING OPERATIONS SURVEILLANCE REQUIREMENTS 4.9.10.1 Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> p r i o r t o the s t a r t o f removal o f a c o n t r o l rod and/or t h e associated c o n t r o l rod d r i v e mechanism from t h e core and/or r e a c t o r pressure vessel and a t l e a s t once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> t h e r e a f t e r u n t i l a c o n t r o l r o d and associ-ated c o n t r o l r o d d r i v e mechanism are r e i n s t a l l e d and the c o n t r o l r o d i s i n s e r t e d i n t h e core, v e r i f y t h a t :
- a.
- b.
C.
- d.
- e.
- f.
The r e a c t o r mode switch i s OPERABLE per Survei 1 lance Requirement 4.3.1.1 o r 4.9.1.2, as applicable, and locked i n the Shutdown p o s i t i o n o r i n t h e Refuel p o s i t i o n w i t h t h e "one rod out" Refuel p o s i t i o n i n t e r l o c k OPERABLE per S p e c i f i c a t i o n 3.9.1.
The SRM channels are OPERABLE p e r S p e c i f i c a t i o n 3.9.2.
The SHUTDOWN MARGIN requirements o f S p e c i f i c a t i o n 3.1.1 are s a t i s f i e d per S p e c i f i c a t i o n 3.9.10.1. c.
A l l other c o n t r o l rods i n a f i v e - b y - f i v e a r r a y centered on t h e c o n t r o l r o d being removed are i n s e r t e d and e l e c t r i c a l l y o r h y d r a u l i c a l l y disarmed o r t h e f o u r f u e l assemblies surrounding t h e c o n t r o l r o d o r c o n t r o l r o d d r i v e mechanism t o be removed from the core and/or r e a c t o r vessel are removed f r o m t h e core c e l l.
A l l other c o n t r o l rods are inserted.
A l l f u e l loading operations are suspended.
HOPE CREEK 3/4 9-14
REFUELING OPERATIONS MULTIPLE CONTROL ROD REMOVAL LIMITING CONDITION FOR OPERATION
'v 3.9.10.2 Any number of control rods and/or control rod drive mechanisms may be removed from the core and/or reactor pressure vessel provided that at least the following requirements are satisfied until all control rods and control rod drive mechanisms are reinstalled and all control rods are inserted in the core.
- a.
- b.
C.
- d.
- e.
- f.
The reactor mode switch is OPERABLE and locked i n the Shutdown position or in the Refuel position per Specification 3.9.1, except that the Refuel position "one-rod-out interlock may be bypassed, as required, for those control rods and/or control rod drive mechanisms to be removed, after the fuel assemblies have been removed as specified below.
The source range monitors SRM are OPERABLE per Specification 3.9.2.
The SHUTDOWN MARGIN requirements of Specification 3.1.1 are satisfied.
All other control rods are either inserted or have the surrounding four fuel assemblies removed from the core cell.
The four fuel assemblies surrounding each control rod or control rod drive mechanism to be removed from the core and/or reactor vessel are removed from the core cell.
All fuel loading operations shall be suspended.
APPLICABILITY: OPERATIONAL CONDITION 5.
ACTION:
With the requirements o f the above specification not satisfied, suspend removal o f control rods and/or control rod drive mechanisms from the core and/or reactor pressure vessel and initiate action to satisfy the above requirements.
HOPE CREEK 3/4 9-15
REFUELING OPERATIONS SURVEILLANCE REQUIREMENTS X
d 4.9.10.2.1 Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior to the start of removal of control rods and/or control rod drive mechanisms from the core and/or reactor pressure vessel and at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter until all control rods and control rod drive mechanisms are reinstalled and all control rods are inserted in the core, verify that:
- a.
The reactor mode switch is OPERABLE per Surveillance Requ or 4.9.1.2, as applicable, and locked in the Shutdown pos the Refuel position per Specification 3.9.1.
The SRM channels are OPERABLE per Specification 3.9.2.
- b.
rement 4.3.1.1 tion or in
- c.
The SHUTDOWN MARGIN requirements of Specification 3.1.1 are satisfied.
- d.
All other control rods are either inserted or have the surrounding four fuel assemblies removed from the core cell.
- e.
The four fuel assemblies surrounding each control rod and/or control rod drive mechanism to be removed from the core and/or reactor vessel are removed from the core cell.
- f.
All fuel loading operations are suspended.
.-/~'
4.9.10.2.2 Following replacement o f all control rods and/or control rod drive mechanisms removed in accordance with this specification, perform a functional test of the "one-rod-out" Refuel position interlock, if this function had been bypas sed.
HOPE CREEK 3/4 9-16
a;..--
REFUELING OPERATIONS L
3/4.9.if RESIDUAL PEAT REWVAL AND COOLANT CIRCULATICN cfIGH WATER LEVEL LIKITING CONDITION FOR OPERkTI3N 3.9.11.1 A t least one shutdown cooling aode loop of the residual heat removal (RHR) systea shall be OPERABLE and i n operation* with:
- a.
- b.
One OPERABLE RYR heat exchanger.
APPLICABILITY:
vessel and t h e water level is greater than or equal to 22 feet 2 inches abgve the top o f the nactor pressurt veSS8l flange and heat losses to ambientnx are not sufficient to maintain OPERATIONAL CONDITION 5.
OPERATIONAL CONDITION 5, when irradiated fuel is i n the r e a c t o r ACTION: -
- a.
With no RHR shutdown cooling -de loop OPERABLE, within one hour and a t least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter, demonstrate tho operability of a t least one alternate method capable o f decay hoat renoval.
Otherwise, suspend all operations involving an increase in the reactor decay heat load and establish SECONDARY CONTAImENT INTEGRITY w i t h i n 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
- b.
W i t h no RHR rhutdom cooling reode establish reactor coolant cfrculatfon by an alternate method and monitor reactor coolant w r r r t u r r a t least once per hour.
f n oprrrtion, w i t h i n one hour S U R V E I L W E REQUIRMENTS 4.9.11.1 A t least one shutdown cooling mode loop o f the residual heat removal s y r m or alkrnata method shall be verified to bo i n opwrtion and circulating reactor coolant a t least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
The shutdown coollng p~plp may be removed fma oporrtlon for up t o 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> per 8-hour period.
ature w i 11 occur (even though REFUELING condi tlons are being mi ntai ncd).
- Ambient losses airst be such t h a t no increase i n reactor vessel water temper-HOPE CREEK 314 9-17
REFUEL I NG 0 PE RAT I ON 5 i..
LOU WATER LEVEL LIMITING CONCTTInN
FOR OPERATT3N 3.9.11.2 system shall be CPERABLE and at least one loop shall be in operation,* with each loop consisting o f :
Two shutdown cooling mode loops o f the residual heat removal (RHR) a.
One OPERARLE 2HR pump, and
- b.
One OPERAELE RHR heat exchanger.
APPLICABILITY: OPERATIONAL CCNDITION 5, when irradiated fuel i s in the reactor vessel and the water level is less tnan 22 feet 2 inches above the top of the reactor pressure vessel flange and heat losses to ambient** are not sufficient to maintain OPERATIONAL CONDITION 5.
ACTION:
- a.
With less than the above required shutdown cooling mode loops of the RHR system OPERABLE, within one hour and at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> there-after, demonstrate the OPERABILITY of at least one alternate method capable o f decay heat removal for each inoperable RHR shutdown cooling mode loop.
With no RHR shutdown cooling mode loop in operation, within one hour establish reactor coolant circulation by an alternate method and monitor reactor coolant temperature at least once per hou'r.
- b.
I I
SURVEiLLANCE REQUIREMENTS 4.9.11.2 system or alternate method shall be verified to be in operation and circulating reacror coolant at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
A t least one shutdown cooling mode loop of the residual heat removal
'The shutdown cooling pump may be removed from operation f o r up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> per 8-nour period.
Ambient losses must be such that no increase i n reactor vessel water temper-ature w i 11 occur (even though REFUELING condi ti ons are being maintained).
\\
- I.
HOPE CREEK 314 9-i8 P,mendment No. - 19 _.
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ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 20030210 PSEG Internal Use Only Page 1 of 1 PSEG NUCLEAR FIRE DEPARTMENT NC.FP-AP.P-0025 (a) - Rev. 0 PRECAUTIONS AGAINST FIRE Sponsor Organlzatlon: Emergency Sewices REVISION
SUMMARY
Biennial Review performed YBS No -
NIA -
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- 2.
- 3.
- 4.
- 5.
This is a new procedure. This change is in support of revision number 6 to administrative procedure NC.NA-AP.ZZ-O025(Q), Operational Fire Protection Program. (80035483).
NAAP-25 is being revised to focus on high l6Vd process concepts, resulting in relocation of some process description detail and the associated forms, tables, and attachments to other lower tier documents.
This revision represents a significant editorial incorporation of the NAAP-25 Hot Work Program into a new stand alone procedure. The NAAP-25 sections utilized to create this procedure were primarlly:
0 5.8, Ignitable Metals 5.5, Control of Ignition Sources - Hot Work The associated NAAP-25 Hot Work forms, tables, and attachments have been relocated to this procedure.
Added term seismic joints to step 5.3.2.G to better define penetration in a fire barrier (70012801)
Added new section 5.1 1 Aerosol Products that deals with National Fire Protection Association (NFPA) Code 308, IMPLEMENTATION REQUIREMENTS:
Effective Date:
APPROVED:
Manager - Emergency Services
SECTION I. 0 2.0 3.0 4.0 5.0 ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 20030210 PSEG Internal Use Only 2
6.0 7.0 8.0 TABLES TABLE I TABLE 2 FORMS FORM -l Common PRECAUTIONS AGAINST FIRE TABLE OF CONTENTS TITLE PAGE PURPOSE.....................................................................................................
2 SCOPE.........................................................................................................
2 RESPONSIBILITIES...................................................................................... 2 PROCESS DESCRIPTION............................................................................
3 PROCEDURE................................................................................................. 4 5.4 5.2 5.3 5.4 5.5 5.6 5.7 5.8 5.9 5.10 5.11 Response to Fire Emergencies............................................................ 4 Fire Prevention Requirements.............................................................. 5 Fire Protection Impairment Program.................................................... 8 Control of Ignition Sources - Hot Work................................................. 8 Combustible Material Control - General I' All Areas.............................. 9 Transient Combustibles........................................................................ 11 Transient Combustible Fire Load Limits................................................ 11 Fire Department Transient Combustibles Inspections.......................... 12 Control of Flammable and Combustible Liquids and Gases................. 13 Aerosol Products.................................................................................. 15 Combustible Material Control - Safety Related Areas.......................... IO RECORDS......................................................................................................
15 DEFINITIONS................................................................................................ 16 REFERENCES..............................................................................................
17 Limited Transient Combustible Areas -Salem Station....................................
20 Common Transient Combustible/Estimated Heat Content.............................. 21 Transient Combustible Permit........................................................................
22 Page 1 of 22 Rev. 0
ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 20030210 1.o 1.I 1.2 2.0 3.0 3.1 3.2 NC.FP-AP.ZZ-0025 (Q)
PURPOSE This procedure is established to secure a reasonable level of safety to life and property from fire hazards incident to the occupancy and maintenance of PSEG Nuclear structures and facilities.
This procedure provides MINIMUM fire prevention measures and directs the user to procedures, which govern other fire protection processes in greater detail.
SCOPE This procedure applies to all PSEG Nuclear buildings and facilities located on the owner controlled property.
RESPONSIBILITIES All Directors and Mananers Ensure their organizations implement the applicable fire prevention requirements within facilities they maintain and in which they conduct work.
Manarrer - Emeraency Services Coordinate and implement the Operational Fire Protection Program requirements for protecting plant, systems, components, and materials.
Ensure General Employee Training (GET) discusses required employee actions in the event of fire.
Ensure walk downs are performed to monitor control of transient com bustibles.
Ensure an inspection program is implemented to verify compliance with applicable regulations.
Ensure Hot Work, Transient Combustible, and Fire Protection Impairment Permits (FPtPs) are properly prepared, authorized, and implemented.
Maintain daily communication with the Operations Superintendent andlor Control Room Supervisors on the status of Salem and Hope Creek Station FPIPs. [CD-434E].
Ensure ignition source control and monitoring for fire prevention.
Interface with fire insurance inspectors, state and local enforcement agencies Interface with offsite fire departments.
Define non-FD responsibility for required testing and inspection of fire protection equipment.
Ensure timely and effective preventive and corrective maintenance on fire protection systems and components assigned to the FD.
Common Page 2 of 22 Rev. 0
4LL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 200302 10 3.3 3.4 4.0 4.1 4.2 4.3 NC.FP-AP.U-0025 {Q)
Job Supervisors - Salem and Hope Creek Stations 0
Comply with fire prevention requirements of the Fire Protection Im painnent, Hot Work, and Transient Com bustible Control processes Ensure work practices support control of ignition sources, flammable liquids and gases, and ignitable metals.
All Personnel 0
Comply with the requirements of this procedure.
0 Ensure appropriate response to fires and fire alarms for the affected area.
0 Ensure fire prevention requirements are observed.
PROCESS This procedure identifies elements of the Operational Fire Protection Program categorized as Fire Prevention and Readiness to Detect and Suppress Fires.
This procedure is implemented in full detail by the following supplemental Fire Department administrative procedures.
SC.FP.AP.2Z-O003(Q), Actions for Inoperable Fire Protection - Salem HC.FP-AP.ZZ-O004(Q), Actions For Inoperable Fire Protection-Hope Creek NC.FP-AP.ZZ-O005(Q), Fire Protection Surveillance I Periodic Test Program 0
NC.FP-AP.ZZ-0009Q), Fire Protection Training Program NC.FP-AP.Z-001 O(Q), Fire Protection Impairment Program NC.FP-AP.ZZ-0012(Q), Safe Hot Work Practices NC. FP-AP.ZZ-O020(Q), Compensatory Measure Fire Watch Program Users of this procedure are advised to refer to the above departmental procedures for the full detaii and administrative controls which these procedures provide.
Common Page 3 of 22 Rev. 0
4LL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 200302 10 5.0 5.1 5.1.1 5.1.2 5.1.3 51,4 5.1.5 5.1.6 NC.FP-AP.ZZ-0025 (Q)
PROCEDURE Response to Fire EmerSencies In the event a fire alarm sounds within an office facility, (e.g. Material Center, Nuclear Operations Services Facility (NOSF), PSEG Nuclear Administration Building (TB-2), Hope Creek Administration 8uilding, Processing Center, or "8" Clean Facilities Building) evacuate the building in accordance with the established routes.
A. Accountability should be performed to ensure all personnel are out of the building and at a specified gathering place.
B. Evacuees should not crowd the entrances and should await further instructions from either the control room or FD personnel.
C. NEVER use the elevator to leave the building.
D. Always ensure fire boundary doors are closed.
For fire alarms within the Salem and Hope Creek Stations, listen to the instructions via the page system provided.
Upon discovering a fire, report it using the telephone via 3333, a manual pull box, or page system A. Use page line 1 for Salem U/1, line 2 for Salem U/2 B. Use page line 5 at Hope Creek When reporting an emergency, callers should provide information such as nature of the emergency, the location, and any equipment involved.
After reporting a fire, employees should fight the fire if its within their training and capabilities.
A. If not within the training and capabilities of the employee(s), to fight the fire, then the employee(s) should alert others in the area, evacuate to a safe distance and standby to direct FD personnel to the scene.
When the Control Room receives a report of a fire they should dispatch appropriate personnel in accordance with the applicable procedures.
Common Page 4 of 22 Rev. 0
NC. FP-AP.ZZ-0026 (Q) 5.2 Fire Prevention Requirements 5.2. I General The following conditions are prohibited and should be corrected immediately IAW N. J.A.C. 570-2-10 (New Jersey State Fire Code). These provisions should be applied in all buildings and facilities:
A. Dangerous conditions which are liable to cause or contribute to the spread B. Conditions that would interfere with the use of any fir0 protection C. Obstruction of egress pathways such as stairwells, exit doors and D. Accumulation of dust or waste in HVAC systems or grease in kitchen or E. Accumulations of grease on kitchen cooking equipment, or oil, grease or F. Accumulation of rubbish, waste, paper, boxes, or other combustible of fire.
equipment.
corridors.
other exhaust ducts.
dirt upon, under or around any mechanical equipment.
materials (trash) or excessive storage of any combustible materials in an area not designed for that purpose.
installed electrical wiring, equipment or appliances.
equipment for handling or use of combustible, explosive or othewise hazardous materials.
\\. Dangerous amounts of corn bustible, explosive or otherwise hazardous materials.
J. All equipment, materials, processes or operations that are in violation of the provisions and intent of this procedure, the MJ State Fire Code and any procedures for safe plant operation.
the FD (Ext. 2800/2803) for evaluation G. Hazardous conditions arising from defective or improperly used or H. Hazardous conditions arising from defective or improperly installed K. Fire Protection deficiencies observed should be immediately reported to L. Smoking is prohibited in all buildings and areas unless specifically posted.
5.2.2 Common Storage A. All storage areas should be maintained in a neat and orderly condition.
B. Storage of material should not affect access to or obstruct fire protection C. Thirty inches (30") of clearance, free of combustible material, shall be D. Materials should not be placed, stored or kept in any portion of an exit, systems or fire fighting equipment.
maintained around energized electrical panels.
elevator or at the bottom of a stairway or other means of escape.
Page 5 of 22 Rev. 0
NC. FP-AP.ZZ-0025 (Q) 5.2.3 Fire Barriers A. Fire Doors shall be maintained closed at all times (unless provided with an B. Blocking, propping or leaving a fire door open without proper authorization approved hold-open device).
for any time period is a violation of our fire protection program and a serious fire safety hazard. [CD-174B, CD1753X, CDJ64EI C. It is imperative that ALL PERSONNEL make sure fire doors close behind them. For doors with hold open devices, the door opening and door swing area need to be free of obstructions.
etc,) shall be maintained, repaired, restored or replaced when damaged, altered, or penetrated D. All required fire resistance rated assemblies (ceiling tiles, walls, doors, 5.2.4 E'lectrlcal wiring.
not be affixed to structures; extended through walls, ceilings or floors, or under doors or floor coverings; nor shall such cords be subject to environmental damage or physical impact.
other device not complying with NFPA 70, shall be prohibited.
A. Installed firefighting equipment, with the exception of portable fire A. Extension cords and flexible cords shall not be a substitute for permanent B. Extension cords and flexible cords shall be a minimum of 12/3 gauge and C. Multi-plug adaptors, such as cube adaptors, unfused plug strips or any 5.2.5 Fire Protection Systems extinguishers, may only be utilized for firefighting at the direction of the NFPS.
- 6. Use of fire protection water for non-fire protection purposes {not within the scope of a T-Mod), is prohibited without permission from the OSCRS of the affected station and the NFPS, along with issuance of a permit for tracking.
C. Actuated detection or suppression systems should not be shut off until authorized by the NFPS or affected station OSKRS. The FD should be notified immediately whenever a portable fire extinguisher is discharged.
D. Fire extinguishers mounted in the stations are for emergency us0 only.
E. Fire extinguisher location changes at Hope Creek are governed by the configuration control program and cannot be made until approved and documented via the design change process. In addition, use. of certain type ABC dry chemical extinguishers are prohibited at Hope Creek station.
Contact the FD for specZfic information. [CR990312177]
Common Page 6 of 22 Rev. 0
NC-FP-AP.22-OOPS (Q) 5.2.6 Portable Space Heaters A. Any use of portable space heaters, whether in a safety related or non-safety related application, will require a review and approval by Fire Protection. Additionally, the following provislons also apply to the use of portabte space heaters.
B. Portable space heaters use in office areas ur trailers are prohibited.
C. Portable space heaters used to maintain operability of Safety Related structures, systems, or components (SSCs) are installed In accordance with NC.DE-AP.ZZ-O030(Q), Control of Temporary Modifications.[C10403].
D. In non-safety related areas to support personnel comfort or equipment operation, use of a portable space heaters, will require a review by fire protection with the following restrictions:
- 1. The department placing the heater is responsible for placing it in the plant, inspecting it to ensure the unit is In serviceable condition and all safety features are operational.
- 2.
The responsible department shall monitor the operation of the heater at all times, E. Portable space heaters should be electrically powered where possible.
F. A clear combustible zone of 10 around the heater should be maintained at all times.
G. Portable heaters should not be placed in high traffic areas or other areas where personnel could come in contact with them.
H. Fuel fired portable heaters should not be used in a building unless all other alternatives are exhausted.
- 1.
Fuel fired portable heaters may not be used without adequate ventilation J. Fuel fired portable heaters may not be used in confined spaces.
K. Fuel fired portable heaters may not be fueled while in operation L. A portable fire extinguisher must be present while a fuel fired portable heater is in operation.
Common Page 7 of 22 Rev. 0
TlVE ON-THE 10
-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE NC-FP-AP.224025 (Q) 5.3 Fire Protection lmpalrment Program 5.3.1 The Fire Department is the fire protection impairment authority, responsible for implementing the Fire Protection Impairment Program in accordance with administrative procedure: ND.FP-APZ-001 O(Q), Fire Protection Impairment Program. [CD434E]
Work activities affecting the following fire protection systems, equipment and barriers shall be controlled through the Fire Protection Impairment Program 5.3.2
[CD-I74B, CD438A, CD-531A, CD-643A, CD-753X, CD-784E, CD-807X, CD-821q:
A. Fire water system (Le. pumps, hydrants, piping, hose stations, deluge and B. Carbon dioxide (CARDOX), Halon, and Foam systems.
C. Smoke and thermal detection systems.
D. Fire alarm and associated circuitry, E. Manual pull box alarm stations.
F. Fire doors and fire dampers.
G. Penetrations, including seismic joints, in fire barriers. 17001 28O11.
H. Marinite boards and structural steel fireproofing.
- 1.
Fire Wrap on cable tray, conduits, cables, and vent ducts.
J. Radiant energy shields.
K. Appendix "R" 8-hour battery powered emergency lights.
L. Transient corn bustibles in non-safety related areaslrooms.
A Fire Protection Impairment Permit (FPIP) is also required when equipment taken out of service affects operability of a fire protection system (Le. tagging a breaker that powers the detection system).
sprinkler systems).
5.3.3 5.4 5.4.1 5.4.2 5.4.3 5.4.4 Control of tnnition Sources - Hot Work The FD authorizes Hot Work Permits in accordance with procedure NC.FP-AP.Z24012(Q),
Safe Performance of Hot Work.
Work involving ignition sources, such as welding, cutting, burning, grinding or open flame soldering, is considered a fire watch required activity.
A Hot Work Permit (HWP) - Form I of this procedure, shall control all these processes. [CD=220C, CI)-30OX, CDS17Y, CD-754X, CD-821 X]
Work involving wire wheeling, needle gun use, or hot iron soldering is NOT considered hot work and a Hot Work Permit is not required.
A. Precautions identified in this procedure for hot work activities should be applied to ensure a fire safe work area.
B. At the discretion of the NFPS, the use of certain equipment for pre-heating such as Cooper Heat (trade mark) resistance heaters, or similar devices, may be considered Hot Work.
Common Page 8 of 22 Rev. 0
5.5 5.5.1 5.5.2 5.5.3 5.5.4 5.5.5 5.5.6 5.5.7 5.58 5 5 9 NC.FP-AP.U-0025 (Q)
Combustible Material Control - General I All Areas Commonly encountered transient combustibles and their estimated heat content are identified in Table 2.
Flammable or combustible liquids introduced into any aredroom within the stations shall be limited to daily usage. [CD1755X3 Use and storage of flammable liquids for decon within the stations shall be limited to five gallons. [CD-791X]
Items incidentat to Radiation Protection such as catch-bags, step-off-pads, roping, and stanchions are acceptable in limited quantities as determined by the FD. PC and trash drums are evaluated as transient combustibles.
Accumulations of waste, debris, scrap, rags, and other combustibles resulting from work activities shafl be removed from the work area immediately following job completion or at the end of a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period, whichever comes first.
For work in the RCA, combustibles shall be removed given proper consideration for A U R A or contained in a sealed container until removed.
[CD-428X]
Corn bustible structures, (including trailers, sheds, etc.) are prohibited from being placed inside of or within 30 feet of permanent buildings, unless a Fire Protection Impairment Permit (FPIP) or TCP as appropriate is issued authorizing placement.
Introduction of combustibles into Combustible Control Zones (CCZ) at Salem is prohibited without a Transient Combustible Permit (TCP), see Form 1, and appropriate compensatory measures. CCZs are used to limit the potential for exposure of sensitive equipment to fire and/or combustion products, to limit the potential for fire propagation through plant equipment hatches, and to reduce overall fire risk. Fixed combustible material may be installed in CCZs with approval from the fire protection designated design authority or the fire department. Transient combustibles are combustible materials that are not part of permanent plant equipment, components or installations. Transient combustibles are intended to pass through or reside in the plant for a brief duration usually associated with, but not limited to, a maintenance or design change work activity (e.g., combustible liquids, wood
& plastic products, waste materials, scrap, rags, trash bags, electrical extension cords, clothing hoop rings, or other corn bustible materials resulting from a work activity.)
Equipment or supplies (such as new fuel) shipped in untreated combustible containers can be unpacked (in preparation for immediate use) in safety-related areas if required for valid operating reasons. However, the combustible shipping materials shall be removed from the area immediately after unpacking, unless stored in metal containers with tight-fitting, self-closing metal covers or equivalent, Such material unless stored in the approved containers, shall NOT be left unattended at any time. [CD426X]
Rev. 0 Common Page 9 of 22
NC.FP-AP.ZZ-0025 (Q) 5.5.
I O All wood used for maintenance activities, refueling or modification operations, including wood carts and foreign material exclusion (FME) covers, etc. shall be fire retardant (performance rated Exterior Type per National Fire Protection Association Standard 703).
Combustible material for decoration in any building is prohibited, unless the materials have been made flame retardant with an approved flame retardant material or process and are approved by the NFPS prior to the installation, using the planned FPlP process.
Temporary buildings, enclosures, or fixtures within or on buildings should be constructed of noncombustible material. Use of combustible materials may affect compliance with Nuclear Electric Insurance Limited (NEIL) Property Loss Prevention Standards. Permanent items are prohibited unless evaluated and approved in accordance with NC.NA-AP.ZZ-0008 (Q) Configuration Control Program.
Storing combustible or flammable material in stairwells is prohibited.
Combustible or flammable materials in use shall not be stored on or against stations' fire rated barriers, where it will come in contact with such items as:
A. Metal floor hatches, pipes, conduits, cable trays, cable, and HVAC ducts that penetrate fire barriers.
- 6. Penetration seals installed in barriers or embedded sleeves used to form openings in fire barriers.
Combustible Materia! Control - Safety Related Areas Storage of combustibles is NOT permitted in Salem and Hope Creek Station's safety related areas/rooms except when approved by Engineering.
jCD4317YI When approved, storage of combustibles in safety related areas/rooms is permitted only in metal containers with tight-fitting, selfclosing metal covers or equivalent.
Storage for Salem's RCA safety related areas is limited to the engineering approved locations and/or the "A" Controlled Facilities Building, while Hope Creek's RCA safety related areas is limited to the engineering approved locations and/or sprinklered areas of the Turbine Building.
Use of non-fire retardant wood (such as large dunnage, wooden carts, or FME) in any safety-related aredroom, regardless of the total British Thermal Units (BTU) content, requires an evatuation by engineering.
Daily usage of combustible or flammable materials in safety related areas of Salem and Hope Creek Stations is considered transient combustibles and requires Section 5.7 of this procedure to be completed.
55.1 1 5.5.12 5.5.13 5.5.14 5.6 5.6.1 5.6.2 5.6.3 5.6.4 5.6.5 Common Page 10 of 22 Rev. 0
NC.FP-AP.ZZ-0025 (a) 5.7 5.7.1 5.7.2 5.7.3 5.7.4 5.7.5 5.8 5.8.1 5.8.2 5.8.3 5.8.4 5.8.5 Translent Combustibles All work activities requiring the introduction of transient combustible material into SAFETY RELATED AREAS or ROOMS shall be identified and administratively controlled. [CD=226C, CDSOOX, CD463Y, CD-807X, Job Supervisors or designees, for work activities introducing combustible items into safety-related areashooms, are responsible for performing a calculation of the transient combustibles BTU content.
This calculation should be completed during the walkdown phase of the workweek process.
Use Form 1 "Transient Combustible Permit" (TCP) for performing the calculation.
The FD will issue an approved TCP for SAFETY RELATED AREAS or ROOMS and specify compensatory measures based on the following:
A. Total transient combustibles being utilized in a fire aredroom.
B. Pre-existing quantities of transient combustibles found during roving C. Valid hot work activities.
D. Sound Fire Protection Practices.
E. Presence/operability of installed detection or suppressian equipment in Transient Combustible Fire Load Limits Transient combustible fire load limits have been established for safety related areas/roorns within the Salem and Hope Creek Stations and are specified in the Fire Hazards Analysis of each station. [CDSOOX, CD830F)
If it is determined the use of combustibles WILL EXCEED established limits, an engineering assessment is required to determine threshold limits and any additional controls BEFORE the combustibles are brought into the safety related areaslrooms.
Regardless of whether the transient combustibles do or do not exceed the limits, the FD will issue a TCP and specify compensatory measures.
Departments introducing the transient combustibles are responsible for the implementation, control and release of the fire watch compensatory measure.
Salem Station Transient Combustible Load Limit A. 400,000 BTU per fire area. [84220]
B. Table 1 contains the areas in both Salem Stations, which have restrictions CD-8214 patrols or the current TCP logs.
- areahom, to the established limits with regards to transient combustibles, Common Page 11 of 22 Rev. 0
NC.FP-AP.22-0025 (a) 5.8.6 Hope Creek Station Transient Combustible Load Limit A. 4,480,000 BTU per room. [CD839F]
B. Hope Creek rooms have restrictions with regards to transient combustibte load limits: pD-463rJ I. Room 4326 (CRD Removal and Repair Area) and Room 4333 (CRD Storage Area): Transient combustibles are prohibited with the exception of CRD cleaning and rebuitding equipment.
- 2. Vestibules 5101,5201 and Electricai Raceways 5216 and 5336:
Transient combustibles are prohibited.
3, Main Control Room and Control Complex Peripheral Rooms:
Transient combustibles are limited to Class "A" materials and small quantities of cleaning supplies incidental to Control Room Complex functions. This area shall not be used for any storage, as storage shall be limited to the peripheral rooms only. This does not include small quantities of items such as a ream of paper or stationary supplies like pens and notepads.
5.9 Fire Department (FD) Transient Combustibles Inspections 5,g.f 5.9.2 5.9.3 5.9.4 The FD performs a survey of transient combustibles in safety-related areas/rooms as part of their daily housekeeping inspections.
The FD performs a survey of transient combustibles in non-safety-relatad areadrooms as part of their weekly housekeeping inspections.
Non-compliances discovered will be documented using the notification process. v8105]
For Non-Safety Related Areas Only:
A. No combustible or flammable material should be stored in any non-safety related area of either Salem or Hope Creek Stations, unless the area contains a fire suppression system, due to insurance restrictions (NEIL Property Loss Prevention Standards). Exceptions to this requirement are based on a FD evaluation with implementation of compensatory measures.
(performance rated Exterior Type per National Fire Protection Association
[NFPA] Standard 703).
C. Introduction of combustibles considered to be excessive will be tracked by a Fire Protection impairment Permit.
B. All wood used for maintenance activities should be fire retardant wood Common Page 22 of 22 Rev. 0
NC.FP-AP.22-0025 (Q) 5.10 5.10.1 Control of Flammable and Combustible Liquids and Gases Storage or dispensing of flammable or combustible liquids shall be in accordance with NFPA Standard 30, the "Flammable and Combustible Liquids Code", unless an authority having jurisdiction (such as the Nuclear Regulatory Commission (NRC) or the Nuclear Electric Insurance Limited
{NEIL) grants exceptions or variances. All questions regarding compliance with NFPA 30 should be directed to the FD. [CDSOOX, CD-755Q Flammable and combustible liquids shall be handled carefully using the following guidelines: [CD-755x1 A. Kept in Factory Mutual (FM) or Undewriters Laboratory (UL) approved 5.10.2 safety cans when in use, handled and/or dispensed. The only exception is when liquids remain in their original containers to maintain purity.
- 6. Flammable and combustible liquids when NOT in use are stored in FM or UL approved flammable liquid storage cabinets regardless of container.
C. Proper electrical grounding and bonding are provided when dispensing.
A. Portable fuel tanks, or fuel tanks used for storage, require the following 5.q0.3 Portable Fuel Tanks measures:
I, A capacity > 660 gaflons requires a permit by the State of N.J.
- 2. Contained within a dike capabte of being drained frequently to remove accumulated rainwater or spills.
- 3. Located and arranged that access for fire fighting is not restricted.
- 4. Equipped with a lockable, automatic shutoff nozzle.
- 5. Portable ABC type fire extinguisher provided.
impairment perm it is required along with possible compensatory measures.
- 6. If any of the above fire prevention measures arc3 not met, a fire protection 5.10.4 Engineering is to assess and approve all permanent or temporary placements of flammable liquid storage cabinets and gas cylinders in safety-related areas of either Salem or Hope Creek. [CD=755X, CD-830)(1 A. Requests for permanent placement of a cabinet in a SAFETY RELATED area made via NI Notification for engineering evaluation and approval.
- 8. Requests for temporary placement of a cabinet in a SAFETY RELATED area are made via the Transient Combustible Permit process through the Fire Department.
Common Page 13 of 22 Rev. 0
5.10.5 5.10.6 5.10.7 5.1 0.8 5.10.9 5.1 0.
I O Common NC.FP-AP.ZZ-0025 (Q)
Placement of flammable liquid cabinets in non-safety related areas is in accordance with the following:
A. No more than 3 cabinets should be located in any one fire area.
B. In large plant areas (ie Turbine Building) more than 3 cabinets may be installed provided each group of 3 cabinets is separated by at least 100.
C. If the area is fully sprinklered, up to 6 cabinets may be installed in a single area without 100' separation.
Flammable liquid cabinets should be maintained in a neat orderly condition with all latches, hardware and hinges properly functioning.
The maximum allowable quantity of Class I, II, and IllA liquids should not exceed 120 gallons per cabinet.
NO storage of materials other than Class I, II and MA liquid is permitted in a Flammable liquid cabinet.
Flammable Gas Cylinders A. Flammable gas cylinders in any safety-related area or room, regardless of the amount, shall NOT be left unattended for > 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. [78111]
B. Flammable gas cylinders in any safety-related area or room, regardless of the amount requires a TCP.
1, An exception for the TCP would be when a Hot Work Permit ( H W )
is issued for the gas cylinders on a daily basis.
- 2. The requirement to monitor flammable gas cytinders left for longer than one hour in a safety related area or room of the stations remains the same. [CD-SSOX, TS990325186]
Compressed Gas Cylinders A. Personnel working with compressed gas cylinders shall comply with the precautions as specified in desk guide: NC. PM-DGZ-0001 (Z) Storage and Handling of Compressed Gas Cylinders: [CD-754X, CD-830a B. Compressed Gas Cylinder Control Tags must be affixed to cylinder@).
These tags should be maintained to match cylinder level (EMPTY, IN-USE, and FULL).
C. Compressed gas cylinders used for hot work with a valid daily hot work permit issued, are considered "in use". Those not in use will be treated as "storage" and shall MQT be left unattended for longer than one hour in safety related areas of the stations. [781 I I]
D. Cylinders should contain a job information tag attached for identifying the owners. vS990SO3190]
E. Cylinders should be kept in an upright position, secured in specially manufactured holding devices, or to a cart or permanent structure by chain or double wrapped wire (minimum No. 9 gauge), to prevent them from falling or being knocked over.
Page 14 of 22 Rev, 0
NC. F P-AP.22-0025 (Q) 5.1 1 5.1 1,l 5.1 1.2 5.1 1.3 511.4 5.11.5 5.1 1.6 6.0 6,1 6.2 F. Open cylinder valves slowly and close valves when work is complete or when personnel leave the area.
G. Keep cylinders capped whenever regulator is removed, while moving a cylinder, or when cylinder is in storage.
H. Promptly return cylinders to the appropriate gas bottle storage areas upon completion of the work.
- 1.
All storage and issuance of compressed gas cylinders in the gas bottle storage areas should be In accordance with Compressed Gas Association Pamphlet P-1.
J. Warning signs are required in areas where the possibility of hydrogen leakage exists (Re-combiners, Waste Gas Compressors, Decay Tanks, and Turbine Generators) to prevent open flames or other sources of ignition within 35 feet of the hazard. [CD=173B]
Aerosol Products An aerosol is a product that is dispensed from an aerosol container by a propellant.
The definition of aerosol does not apply to the storage or display of containers whose contents are composed entirely of liquid petroleum (LP) gas products.
A. Examples include MAP (methane, acetylene, propane) gas cylinders, B. Uses for these LP-gases might include small hand torches for brazing and An aerosol container is a metal can up to a maximum dze of 33.8 fluid ounces (1 000 mi) or a glass or plastic bottle up to a maximum sire of 4 fluid ounces (1 18 mi).
Storage and display of aerosols shall be in accordance with NFPA standard 308, Manufacture and Storage of Aerosol Products Code.
Aerosols introduced into any safety related areidroom within the stations shall be limited to daily usage.
Daily usage of aerosols in safety related areas of Salem and Hope Creek Stations is considered transient combustibles and is required to be controlled accordingly.
Records Transient Combustible Permits shall be retained with the work package in accordance with NC.NA-APZ-0011 (Q), Records Management Program The FD, in accordance with NAP-11, shall retain transient combustible permits and logs pertaining to hot work permits and fire protection impairment permits. [CP434E]
propane gas cylinders, and butane gas cylinders.
soldering.
(NAP-11). ICD-434E.1 Common Page t 5 of 22 Rev. 0
ON -THE -S POT 7.0 7.1 7.2 7.3 7.4 7.5 7.6 7.7 7.8 7.9 7.10 7.1 I 7.12 7.13 7.14 7.15 Common CHANGES MUST BE ATTACHED FOR FIELD USE NC. F P-AP.22-0025 (Q)
DEFINITIONS Aerosol - A product that is dispensed from an aersol container by a propellant. (N FPA 308).
Combustible Control Zone ICCZ) -An area of the plant in which transient corn bustible materia@) is prohibited without a Temporary Combustible Permit (TCP) and appropriate compensatory measures.
Combustible Liquid - A liquid having a Rash point at or above 100' F (37.8e C).
Combustible Material - Material, which in the form it is used and under the conditions anticipated, WtLL ignite, burn, support com bustion, or release flammable vapors when subjected to fir0 and heat.
Fines - very small particles in a mixture of various sizes.
Fire Barrier - Those construction elements (walls, floors, and their supports),
including beams, coIumns, penetration seats or closures, fire doors and dampers, that are rated by approving laboratories in hours of resistance to fire and used to prevent the spread of fire.
Fire Retardant Wood - Wood that has been treated with fire retardant chemicals and is performance rated Exterior Type per National Fire Protection Association Standard 703).
Fire Watch - An individual who has satisfactorily completed fire watch training and is designated by the FD for compensatory measure, hot work, or both fire watch type duties.
Flammable Lisuld - A liquid with a flash point below 100" F and a vapor pressure not exceeding 40 pounds per square inch absolute at 100" F. (Also known as a Class I liquid.)
Hot Work -Work that involves ignition sources.
lanition Source - Heat or flame from any source capable of igniting combustible or flammable material.
Impairment - Any condition affecting the intended operation of a fire protection system or barrier or posing as a fire hazard within the stations.
Liquified Petroleum Gas (LP-Gasl-Material having a vapor pressure not exceeding that allowed for commercial propane composed predominantly of the following hydrocarbons, either by themselves or as mixtures: propane, propylene, butane {normal butane or isobutene), butylenes (NFPA 58).
Nqncombustlble Material - Material which, in the form it is used and under the conditions anticipated, will NOT ignite, burn, support combustion, or release flammable vapors when subjected to fire and heat.
Permanent - A condition with an indefinite period of time.
Page 16 of 22 Rev. 0
A L L ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 20030210 NC.FP-AP.22-0028 (Q) 7.1 6 7.17 7.18 7.1 9 7.20 7.21 7.22 7.23 8.0 8.1 8.3 0.4 8.5 8.6 8.7 8.8 8.9 8.1 0 8.2 Safety Related Area - An area that contains systems and components required to shut down the reactor, mitigate the consequences of postulated accidents or maintain the reactor in a safe cold shutdown condition.
Sealed Container - A metal container sealed by means of a tightly fitting lid or other device so neither liquid nor vapors can escape at room temperature, Storage - The placing of items in a location and leaving them unattended until later use or disposal.
Temoorary - planned time duration of no more than 60 days.
Temporary Buildinas - Any temporary structure or supported protective covering where personnel a n move freely in and out of the structure.
Temporary Enclosure - Any supported protective covering where personnel cannot move in and out of the structure or where the protection is accomplished by draping the protective cover over the material and/or equipment. This definition also applies to temporary enclosures inside of permanent buildings.
Temaorarv Fixture - Something securely ptaced or attached as an appendage or as a structural part of a building (Le. door, portable steps, storage rack), constructed of a combustible material. This definition does not apply to furniture permanently installed in Station facilities.
Transient Combustible - Any combustible material introduced into a Safety-Related aredroom which will be used to complete an assigned task or left unattended for any period of time. It does not include items moved through areas in route to the final destination, unless the items are left un-monitored during transit.
REFERENCES NC. NA-AP.ZZ-0025 (Q), Operational Fire Protection Program N.J.A.C. 5:70-2 New Jersey State Fire Code Salem Updated Final Safety Anatysis Report (UFSAR), Section 9.5, - Fire Protection System Salem Fire Protection Report - Fire Hazards Analysis HCGS Updated Final Safety Analysis Report, (UFSAR} Section 9.5.1, - Fire Protection Program HCGS UFSAR, Appendix 9A, - Appendix R Comparison Appendix A to Branch Technical Position APCSB 9.5-1, Guidelines for Fire Protection for Nuclear Power Plants Docketed Prior to July I, 1976 Branch Technical Position CMEB 9.5-1, Guidelines for Fire Protection for Nuclear Power Plants NRC Supplemental Guidance, Nuclear Plant Fire Protection Functional Responsibilities, Administrative Controls and Quality Assurance, August 1977 Occupational Safety and Health Standards Title 29, Chapter XVII, Part 1910.36FR 10466 Common Page 17 of 22 Rev. 0
4LL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 20030210 8.11 8.12 8.13 8.14 8.15 8.16 8.17 8.18 8.18.1 8.18.2 8.18.3 8.18.4 8.18.5 8.18.6 8.18.7 8.18.8 8.1 8.9 8.1 8.1 0 8.18.11 8.10.12 8.1 8.13 8.1 8.14 8.18.15 8.18.16 8.1 9 8.19.1 8.1 9.2 8.19.3 8.19.4 8.1 9.5 8.19.6 8.18. I 7 NCFP-AP.22-0025 (Q) 1 OCFRSO Appendix R, Fire Prot. Program for Nuclear Power Facilities Operating Prior to Jan. I, 1979 NC. DE-PS.22-0001 (Q), Fire Protection Programmatic Standard IEEE Standard 634 - 1978 Hope Creek Station Generic Letter 86-10 Submittal, dated May 13, 1986 HC.DE-PS.22-0021 (F), Hope Creek Penetration Seal Program SC.DE-PS.ZZ-O035(Q), Salem Penetration Seal Program National Fire Protection Association Standard 701, Standard Methods for Fire Tests for Flame Propagation of Textiles and Films Cross References NFPA 30 - Flammable Liquids Code, National Fire Protection Association (NFPA)
NFPA 30B - Code for the Manufacturer and Storage of Aerosol Products NFPA 58 - Liquefied Petroleum Gas Code NFPA 703 - Standard for Fire Retardant Impregnated Wood and Fire Retardant Coatings for Building Materials Compressed Gas Association Pamphlet P-1, Safe Handling of Compressed Gases in Containers.
Property Loss Prevention Standards of Nuclear Generating Stations - Nuclear Services Organization (NSO).
NC. NA-AP.ZZ-O008(Q), Configuration Control Program NC,NA-AP.22-0011 (Q), Records Management Program (NAP-I 1)
NC. DE-AP.ZZ-O030(Q), Control of Temporary Modifications SC.FP.AP.ZZ-O003(Q), Actions for Inoperable Fire Protection - Salem HC.FP-AP.Z-OOM(Q), Actions For Inoperable Fire Protection-Hope Creek NC. FP-AP.ZZ-O005(Q), Fire Surveillance and Periodic Test Program NC. FP-AP.ZZ-O009(Q), Fire Department Training Program NC.FP-AP.ZZ-O010(Q), Fire Protection Impairment Program NC. FP-AP.ZZ-O012(Q), Safe Petformance of Hot Work NC. FP-AP.ZZ-0020(Q)q Compensatory Measure Fire Watch Program NC.PM-DG,ZZ-OOOl (Z) Storage and Handling of Compressed Gas Cylinders Commitment Documents: - Hope Creek CD-156A (NRC El Circular 80-09) (See FPAP-12)
CD-1736 (INPO SOER 82-09)
CD-'l74B (SOER 82-1OR02, 04, 07)
CD-226C (NRC IE Bulletin 75-04A)
CD300X (FSAR 8.1.4.14.3.1)
CD317Y (FSAR Q 430.85(B))
Common Page 18 of 22 Rev. 0
~~
ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 20030210 NC.FP-AP.ZZ-0025 (Q) 8.1 9.7 8.1 9.8 8.1 9.9 8.1 9. I O 8.19.11 8.19.12 8.1 9.13 8.19.14 8.19.15 8.19.16 8.1 9.1 7 8.1 9.1 8 8.1 9.1 9 8.1 9.20 8.19.21 8.1 9.22
- 8. 'I 9.23 8.20 8.20.1 8.20.2 8.21 8.21. I 8.21.2 8.21.3 CD-426X (FSAR 9.1.1.3,2)
CD-434E (IR 86-062) (LER 86-062)
CD-438A (INPO SER 055-81)
CD-463Y (FSAR Appendix 9k6.5)
CD-643A (INPO O&MR 64)
CD-652A (INPO O&MR 73) {See FPAP-12)
CD-682D (INPO SER 01-85) (See FPAP-12)
CD-753X (FSAR 9.5.1.1.10)
CD-7fi4X (FSAR 9.5.1.1.11)
CD-755X (FSAR 9.5.1.I
-33)
CD-764E (NHO IR 86-274)
CD-791X (FSAR 9.5.1.2.32)
CD-807X (FSAR Appendix 9A.IIl.K)
CD-821X (FSAR 9.5.1.5.3)
CD-830X (BTP CMEB 9.5-1.~.8.1)
CD-939F (NRC OPEN 354/88-18-F1)
Closina Documents - Salem C0403 (NRC-INFO 89-04]
C0549 (SR 272192-8) (See FPAP-I 0)
Commitment Documents - Engineering 781 05 (Appendix A to BTP APSCB 9.5-1, Section 8.2) 781 11 (Letter from PSE&G to NRC, 07/26/78) 84220 (Memo from Vargas to PSE&G, 03/23/84)
Common Page 19 of 22 Rev. 0
NC.FP-APZ-0025 (Q)
TABLE 1 SALEM STATION COMBUSTIBLE CONTROL ZONES SALEM 1 & 2 Separation
- 2
' ccz-22 Hatch Hatch opening between 2FA-EP-78CQFA-PP-1 OOH Hatch opening between ZFA-AB1 WCEFA-A5848 Hatch opening between 2FA-AB4 22Bi2FA-AE1 OOC 2
ccz-23 nstch Hatch IOOAUX 122' A,,X 2
CCZ-24 1
I CCZ-25
' 88' TG Off Site Power Off Site Power from No 13 & 14 Power Transformer 2
CCZ-26 88' TG Off Site Power Off Site Power from No 23 & 24 Power Transformer 1
CCZ-27 100' RC 335 Panel Reactor Containment around the 335 Panel 1
CCCZ-28 81' RC Pressurizer Reactor Containment around the Pressurizer 1
CCCZ-29 203' RC Pressurizer Reactor Containment around the Pressurizer Platform 2
CCCZ-31 81' RC Pressurizer Reactor Containment around the Pressurizer 2
CCCZ-32. 103' RC Pressurizer Reactor Contalnment around the Pressurizer Platform Reference Drawings: 605810,605811,605812,605813,605815,605816, & 605817 Common Rev. 0
~
ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 20030210 NC.FP-AP.22-0025 (Q)
TABLE 2 COMMON TRANSIENT COMBUSTIBLES f ESTIMATED HEAT CONTENT Acetylene Common size acetylene cylinders found on site:
Cable insulation Cardboard Charcoal Cloth/cloh pcs Combustible liquid Dry ion resins Fiberglass ladder (6)
Flammable liquid Paper Plastic Rubber Titanium Wood 20,800 BTU/lb MC Size Cylinder (4 D x f2 H) = % Ib<
B Size Cylinder (6Dx ?9H) =
2lb R Size Cylinder (7Dx 25H) =
3 Ib, S Size Cylinder
(?Oft D x 30 H) = 15 Ib, M Size Cylinder (12 D x 36 H) = 18 lb:
L Size Cylinder (72 D x 39 H) = 24 Ib:
12,000 BTUllb 6,000 BTUllb 13,000 BTU/lb 16,000 STU/lb (I PC = ?4 LB or 8,000 BTUs) 160,000 BTU/gal 12,000 BTUAb 32,000 BTU/ea 90,000 BTWgaf 8,000 BTU/tb or 453,000 BTU/cu. ft.
(? case of copierpaper = 416,000 BTUs) 20,000 BTUAb (I sheet of herculite - SOx 50 = 50,000 BTUs) 1 0,000 BTUIlb (I - 36 hose, 50 = 50,000 BTUs) 8,500 BTU/LB 9,000 BTU/lb or 414,000 BTU/cu. ft.
(I plank - 2x 6x 10 = 277,500 BTUs)
(I foot = 5,333 B TUS)
Common Rev. 0 Page 21 of 22
NC.FP-AP.22-0025 (Q)
FORM I TRANSIENT COMBUSTIBLE PERMIT LOCATION FIRE A R A START DATE DURATION 1 ORDER#
I I
1 JOB SUPERVISOR I
EVALUATE TRANSIENT COMBUSTlBLES WHICH YOU ANTICIPATE BEING USED COMBUSTIBLES ESTIMATED HEAT CONTENT TOTAL BW's FLAMMABLE LIQUID 90,000 BTUlGAtLON 360,000 BTUlGALLON COMBUSTIBLE LIQUID or GREASE CHARCOAL 13,000 BTUPOUND 1
FIBERGLASS LADDER 1
5,333 BTU/FOOT OF LADDER I
CABLE INSULATION 12,000 BTUlPOUND CARDBOARD 6,000 BTU/POUND I
8,000 BTUPOUND I
(453,000 BTU/CUBIC FOOT)
PAPER I
16,000 BTU/POUND I
1PC = X POUND I
CLOTH and CLOTH PCs t
I I
PLASTIC 20,000 BTU/POUND 9,000 BTUlPound I
(414,000 BTUlCubic Foot)
WOOD
~
~
~ -7 12,000 BTUPound I
DRY ION RESINS TITANIUM 8,500 BTUPound ACEMENE 20,800 BTU/Pound OTHER Contact FIRE DEPT.
Engineering Assessment Needed Yes No Approved Yes [3 No Completed By Date TCP #
Total BTUs REMOVE AL L TRANSIENT COMBUSTIBLES UPON JOB COMPLf71ON Common Page 22 of 22 Rev. 0
ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE I 2003022 1 PSEG Internal Use Only Paae I of I PSEG NUCLEAR FIRE DEPARTMENT NC.FP-AP.ZZ-0012(4) - Revision 0 SAFE PERFORMANCE OF HOT WORK Sponsor Organization: Emergency Services - Fire Department REVISION
SUMMARY
I. This is a new procedure. This change is in support of revision number 6 to administrative procedure NC,NA-AP.ZZ-O025{Q), Operational Fire Protection Program. (80035483).
NAAP-25 is being revised to focus on high level process concepts, resulting in relocation of some process description detail and the associated forms, tables, and attachments to other lower tier documents.
- 2. This revision represents a significant editorial incorporation of the NAAP-25 Hot Work Program into a new stand alone procedure. The NAAP-25 sections utilized to create this procedure were primarily:
0 0
5.8, Ignitable Metals 5.5, Control of Ignition Sources - Hot Work
- 3. The associated NAAP-25 Hot Work forms, tables, and attachments have been relocated to this procedure.
- 4. This revision meets the requirements of the biennial review.
I IMPLEMENTING REQUIREMENTS Effective On:
I 8 / 03
ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 2003022 1 NC.FP-AP.22-0012 {Q)
SAFE PERFORMANCE OF HOT WORK TABLE OF CONTENTS NUCLEAR COMMON Page 1 of 14 Rev. 0 SECTION 1.o 2.0 3.0 4.0 5.0 6.0 7.0 TITLE Purpose Scope Responsibilities Hot Work Process Overview 4.1 Hot Work 4.2 General 4.3 High Risk Hot Work 4.4 Ignitable Metals 4.5 4.6 Exceptions Approval Process Records References Electromagnetic I Radio Frequency Interference PAGE 2
2 2
3 3
3 4
4 5
5 5
7 8
Table 1 Table 2 Form 1 Hot Work Permit Form 2 Ignitable Metals - Salem Ignitable Metals - Hope Creek Hot Work Authorization Log 9
11 12 14
4LL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 2003022 1 NC.FP-AP.22-0012 (a) 1.o 2.0 2.1 2.2 2.3 2.4 3.0 3.1 3.1.1 3.1.2 3.1.3 3.1.4 3-1
-5 3.2 3.2.1 3.2.2 3.2.3 3.2.4 3.2.5 3.2.6 3.2.7 PURPOSE Provide a method and instruction for preparing, authorizing, and tracking Hot Work Permits, to control and limit the introduction of potential fire hazards into plant areas. [CD300n SCOPE_
This procedure applies to the performance of hot work in the Salem 1 and 2 Stations, the Hope Creek Generating Station, and associated out buildings.
Implementation of this procedure provides a log of the areas where hot work has been authorized including High Risk jobs and hot work in approved fab shops.
This procedure ensures that the Emergency Services - Fire Department (FD) duty Nuclear Fire Protection Supervisor (NFPS) is aware of all hot work jobs, possible fire hazards and their effect on station safety.
This procedure also allows for the tracking of flammable compressed gas cylinders utilized for hot work.
RESPONSIBILITIES Nuclear Fire Protection Supervisor {NFPS) or designee Compile information on the hot work permit ( H W ) log prior to authorizing a permit.
Authorize Hot Work Permit.
Maintain log of authorized hot work permits.
Recognize the extent of hot work jobs in progress during his shift.
Inspect and authorize high risk hot work jobs.
Job Supervisor Ensure their personnel are trained and qualified to perform hot work.
Prepare their jobsite in accordance with the H W.
Inspect their jobsite prior to requesting a HWP number.
Maintain their jobsite in accordance with the HWP.
Ensure precautions are taken regarding work practices that have the potential to generate electromagnetic interference and radio frequency interference Notify the OSlCRS of the affected station prior to welding in the vicinity of a Radiation Monitor System (RMS) component 0
Spiking of the RMS may result, causing unexpected alarms of the RMS.
[TS980622174].
Assign qualified fire watch to the hot work job site.
0 Additional firewatch may be assigned due to sparks or slag streams affecting adjacent fire areas or elevations.
NUCLEAR COMMON Page 2 of 14 Rev. 0
ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 2003022 1 3.2.8 3.2.9 3.3 3.3.1 3.3.2 3.3.3 3.3.4 3.3.5 4.0 4.1 4.1.I 4.1.2 4.1.3 4.2 4.2.1 4.2.2 4.2.3 4.2.4 4.2.5 4.2.6 NC.FP-AP.22-0012 (Q)
Term inate work activities properly to ensure that flammable compressed gas cylinders are not left unattended.
Ensure fire watch remains for 30 minutes after hot work is completed.
Hot Work Fire Watchers:
Present and observant at all times of the work being performed and the surrounding area.
Ensure that the work area is maintained as described on the HWP Inspection at all times while hot work is being performed.
Notify the control room immediately of any fire occurring after ensuring the safety of their co-workers.
Extinguish fires within the capabilities of their training after the fire has been reported.
Remain at their post for 30 minutes following the last hot work.
HOT WORK PROCESS OVERVIEW Hot Work Work involving ignition sources, such as welding, cutting, burning, grinding or open flame sotdering, is considered Hot Work.
Hot work is a fire watch required activity.
A Hot Work Permit (HWP) shall control work involving ignition sources, such as welding, cutting, burning, grinding or open flame soldering. [CD1226C, General Daily Hot Work permits are valid for the one calendar day in which they are issued.
Annual hot work permits, with specific numbers, are issued individually for each approved shop facility on site for compliance with the New Jersey State Uniform Fire Code.
Fire watch are not required in shops where welding and work involving ignition sources are routinely performed or in yard areas where no flammable or combustible materials are located within 35 feet of the hot work, Separate Hot Work tracking logs should be established for the Hope Creek and Salem stations.
The NFPS should review the logs at the beginning of each operating shift to become familiar with hot work that has been authorized for the shift and what detection zones may be affected.
Hot work requires the fire watch to be maintained for 30 minutes after hot work is completed or suspended, to ensure the work area is safe from fire.
CD900X, CD-317Y8 CD-754X, CD-8214 NUCLEAR COMMON Page 3 of 34 Rev. 0
ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 2003022 1 4.2.7 4.2.8 4.3 4.3.1 4.3.2 4.3.3 4.4 4.4.1 4.4.2 NC. F P-AP.U-0012 (Q)
Fire watch is responsible for monitoring the Work area, including levels below floor grating and floor openings.
A portable fire extinguisher, inspected within the previous 30 days, is required for the hot work job.
Hiwh Risk Hot Work Hot Work activities will be defined as HIGH RISK if certain criteria are met.
HIGH RISK hot work activities will require additional administrative controls and supervisory oversight, beyond the normal hot work controls.
The following hot work activities are considered HIGH RISK:
0 Work on tanks, vessels or piping that contained, or had contained combustible or flammable chemicals, solvents, gases, etc.
+ Work in areas identified as having hydrogen as part of equipment or processes.
0 Use of arc gouging or plasma arc welding which produce copious amounts of heat, with molten slag and sparks.
0 Work in cable shafts, filter housings, or ventilation ductwork 0 Work in Hope Creek Drywell [CD-682-D].
0 Work on equipment containing combustible (ignitable) metals, which are listed in Table 1 and 2 of this procedure lqnitable Metals Ignitable metals utilized at Salem and Hope Creek Stations are identified in Table I and Table 2.
The following precautions apply when working on or in proximity of equipment containing ignitable metals:
Metals are considered potentially combustible whenever exposed.
Notify the Fire Department (FD) of work activities involving ignitable metals prior to the exposure of the metal.
Class D fire extinguisher (identified by a yellow star with a "D", located on the extinguisher body), is readily available.
Ignition mechanisms consists of oxy-acetylene or carbon-arc welding, band saw friction cutting, abrasive cutting wheels and grinding operations.
Heat sink effect dissipates ignition source heating when in bulk form.
The ignitable metal hazard increases significantly when shavings, chips, dust, etc., are involved.
Metal fines produced from combustible metals, increase the hazard potential because they act like kindling.
Proper cleanup and removal of accumulations of metal fines, produced in such operations as band saw cutting, are essential for fire prevention and safety.
NUCLEAR COMMON Page 4 of 14 Rev. 0
ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 20030221 NC.FP-AP.ZZ~0012 (Q) 4.5 4.5.1 4.5.2 4.5.3 4.6 4.6.1 4.6.2 5.0 5.1 5.2 Electromaqnetic I Radio Frequency Interference Plant areas susceptible to electromagnetic interference (EMf) and radio frequency interference are listed in NC.NA-AP.ZZ-O005(Q) (NAAP-5), Station Operating Practices, Prior to commencing electric arc welding in plant areas susceptible to EM1 or RFI, the job supervisor should ensure work practice precautions and Main Control Room notifications have been addressed in accordance with NAAP-5.
Hot Work Permit (Form I) provides the job supervisor with the direction to inspect the work area against the precautions for EMI and RFI.
Exceptions Work involving wire wheeling, needle gun use, or hot iron soldering is N m a fire watch required activity and a HWP is not required. However precautions identified in this procedure for hot work activities should be taken to ensure a fire safe work area.
Equipment for pre-heating such as Cooper Heat (trade mark) resistance heaters, or similar devices, is not ordinarily considered Hot Work.
0 At the discretion of tbe NFPS, formal Hot Work administrative controls may be required.
0 The job supervisor is not exempt from enforcing the intent of maintaining any work area in good housekeeping condition.
At the discretion of the job supervisor, formal Hot Work administrative controls may be requested. If any uncertainty exists, regarding the nature of the work activity, the job supervisor should review the activity with the NFPS.
APPROVAL PROCESS Prior to requesting a HWP, the job supervisor should ensure that all of the inspection criteria on the HWP have been satisfied and the job site inspected.
FD personnel develop and record the following baseline information fields in the Hot Work Authorization (HWA) Log:
Job location.
0 Work order number.
0 Type of work (including use of flammable compressed gas cylinders),
0 Responsible supervisor and telephone extension.
0 Affected detection zones.
I Time and date of authorization.
NUCLEAR COMMON Page 5 of 14 Rev. 0
ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 2003022 1 NC.FP-AP.22-0012 (Q) 5.3 5.3.1 5.3.2 5.3.3 5.3.4 5.4 5.4.1 Evaluate For HIGH RISK FD personnel evaluate the hot work activity request against the following HIGH RISK criteria, prior to providing the hot work authorization number:
Work on tanks, vessels or piping that contained, or had contained combustible or flammable chemicals, solvents, gases, etc.
0 Work in areas identified as having hydrogen as part of equipment or processes.
0 Use of arc gouging or plasma arc welding which produce copious amounts of heat, with molten slag and sparks.
Work in cable shafts, filter housings, or ventilation ductwork Work in Hope Creak Drywell [CD-682-D].
Work on equipment containing combustible metals, which are listed in Table I and 2 of this procedure.
IF the job is HIGH RISK, THEN no authorization number will be provided until after the job site is inspected by the Fire Protection Supervisor or his designee.
0 The inspection will determine whether the jobsite complies with the minimum standards set forth on the HWP or whether more stringent requirements need to be implemented.
0 A FD field inspection is REQUIRED to be performed prior to issue of a HIGH RISK HWP.
The NFPS should notify the Operations Superintendent (OS) of High Risk work being conducted.
FD reviews the Impairment tog for any fire protection components in the work area that are inoperable and identifies additional compensatory measures as necessary.
Hut Work Authorization Hot Work may be authorized by FD after completion of the following:
0 FD completes baseline information fields in the HWA Log (Form 2) 0 FD completes a HIGH RISK activity criteria evaluation.
e FD completes a work area inspection and signs/dates the field HWP, for activities designated HIGH RISK.
NUCLEAR COMMON Page 6 of 14 Rev. 0
~ _ _
ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 2003022 1 5.4.2 5.4.3 5.4.4 5.5 5.5.1 5.5.2 5.5.3 5.5.4 5.5.5 6.0 6.1 6.2 6.3 NC.FP-AP.ZZ-OOI2 (Q)
Authorization to perform hot work is indicated by an authorization number.
The authorization number is developed using the following format and associated legend:
Format:
Legend:
o X-YY-MM-DD-ZZ o X - Station where work is performed. (S=Salem, H=Hope Creek) o YY - Last 2 digits of current year o MM-Month o DD - Day that work is being performed o ZZ - Sequential number beginning with 01 each day.
FD enters the authorization number for each activity into the HWA Log.
FD initials the HWA tog and records the time the HWP was issued.
Work Area Hot Work Permit Upon issuance of 8 hot work authorization number, the job supervisor should complete the following fields of the HWP:
HWPNumber Job Supervisor Name Date The job supervisor should ensure the HWP is posted in the work area.
The job supervisor should ensure that the work area is maintained in accordance with the HWP Work Area Inspection criteria.
IF hot work is discontinued for longer than 8 hrs, THEN re-inspection of the area in accordance with Part 2 of the HWP (Form I) is required prior to the commencement of work.
The FD will verify HWP criterialconditions on a random basis after issuance of hot work permits.
RECORDS Hot work authorization logs should be maintained for a minimum of 30 days.
Hot Work Permits and Transient Combustible Permits shall be retained with the work package in accordance with NC.NA-AP.ZZ-001 I (Q), Records Management Program (NAP-1 1 )- [CD-434E]
The FD, in accordance with NAP-I 'l shall retain logs pertaining to hot work permits. [CD-434E]
NUCLEAR COMMON Page 7 of 14 Rev. 0
4LL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 2003022 1 NC.FP-AP.ZZ-0012 (Q) 7.0 7.1 7.2 7.3 7.4 7.5 7.6 7.7 7.8 7.9 7.1 0 7.1 1 7.12 7.1 3 7.14 7.15 7.16 7.16.1 7.162 7.16.3 7.16.4 7.16.5 7.16.6 7.16.7 7.16.8 7.
16.9 REFERENCES
NC-NA-AP.22-0025 (Q) Operational Fire Protection Program New Jersey Administrative Code - NJAC 570-2 NJ State Fire Code NC.PM-DG.ZZ-0001 (Z) Storage and Handling of Compressed Gas Cylinders Compressed Gas Association Pamphlet P-I Property Loss Prevention Standards of Nuclear Generating Stations, -
Nuclear Electric Insurance Limited (NEIL). NFPA 30 - Flammable Liquids Code, National Fire Protection Association NC. IS-TM.22-0001 (Z) PSEG Nuclear Industrial Health and Safety Manual Salem Updated Final Safety Analysis Report (UFSAR), Section 9.5, - Fire Protection System Salem Fire Protection Report - Fire Hazards Analysis HCGS Updated Final Safety Analysis Report, (UFSAR) Section 9.5.1, - Fire Protection Program HCGS UFSAR, Appendix 9A, - Appendix R Comparison Appendix A to Branch Technical Position APCSB 9.5-1, Guidelines for Fire Protection for Nuctear Power Plants Docketed Prior to July 1, 1976 Branch Technical Position CMEB 9.5-1, Guidelines for Fire Protection for Nuclear Power Pfants NRC Supplemental Guidance, Nuclear Plant Fire Protection Functional Responsibilities, Administrative Controls and Quality Assurance, August I977 Occupational Safety and Health Standards Title 29, Chapter XVII, Part 1910.36FR 10466 1 OCFRSO Appendix R, Fire Prot. Program for Nuclear Power Facilities Operating Prior to Jan. 1, 1979 Commitment Documents: - Hope Creek CD-I 56A (NRC El Circular 80-09)
CD-226C (NRC IE Bulletin 75-04A)
CD300X (FSAR 8.1.4.14.3.1)
CD317Y (FSAR Q 430.85(6))
CD-434E (IR 86-062)
(LER 86-062)
CD-652A (INPO O&MR 73)
CD-754X (FSAR 9.5.1.I
.I I)
CD-821X (FSAR 9.5.1.5.3)
NUCLEAR COMMON Page 8 of 14 Rev. 0
ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 2003022 1 Location
- Units 1 & 2 Turbine Bldg.
El. 100'
- Units 1 & 2 Auxiliary Bldg.
El. 84'
- Unit I Auxiliary Bldg.
El. 84'
- Unit 1 Auxiliary Bldg.
El. 100'
- Unit2 Auxiliary Bldg.
El. 100'
- Units 1 & 2 Inner Piping Pen Area, El. 100'
- Units 1 & 2 Service Water Intake Structure NC.FP-AP.ZZ-0012 (Q)
TABLE 1 lgnitable Metals - Salem Page 1 of 2 System(s)
Component Turbine Auxiliary Heat Exchangers Cooling (TAC)
Nos. 11,12,21,22 Component Cooling Heat Exchangers Nos. 11,21,22 Component Cooling Heat Exchanger No. 12 Service Water Diesel Generator Lube Oil Coolers
& Jacket Water Heat Exchangers
{Nos. lA, lB, IC)
Service Water Diesel Generator Lube Oil Coolers
& Jacket Water Heat Exchangers
{Nos. 2A, 2B,2C)
Service Water C hi1 ler Condensers Service Water Service Water Pump Upper Motor Bearing Coolers 25 & 26 NOS. I 1-1 684 21 -23, NUCLEAR COMMON Page 9 of 14 Ignitable Metal Titanium Tubes Titanium Tubes Titanium Plates Titanium Tubes & Heads Titanium Tubes & Heads Titanium Plates Nozzles & Flanges Water Box Housings Turn Around Heads Titanium Tube Coils Rev. 0
4LL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 2003022 1 Location
- Units 1 & 2 Turbine Bldg.
El. 100'
- Units I
& 2 Auxiliary Bldg.
El. 84'
- Units 1 & 2 Auxiliary Bldg.
El. 84'
- Unit I Auxiliary Bldg.
El. 84'
- Unit 2 Auxiliary 3ldg.
El. 84' NC.FP-AP.ZZ-OO?Z (Q)
TABLE I ignitable Metals - Salem Page 2 of 2 System@)
Component Ignitable Metal Turbine Auxiliary Condensate Pump Titanium Tubes Cooling (TAC)
Upper Motor Oil Finned Coil Coolers Service Water Charging Pump Titanium Tubes Bearing Oil Coolers
& Tube Sheets Nos. 11, 12,21,22 Service Water Safety Injection Titanium Tubes
& Tube Sheets Pump Bearing Oil Coolers Nos. 11, 12, 21,22 Service Water ChglSafety Inj.
Titanium Tubes Pump
& Tube Sheets Speed Increaser Gear Lube Oil Cooler Nos. 11 & 12 Service Water ChgBafety Inj.
Titanium Tubes Pump
& Tube Sheets Speed Increaser Gear Lube Oil Cooler Nos. 21 & 22
- Circulating Water Circ Water Intake Structure Circ. Water Pump Titanium Tubes Motor Coolers NUCLEAR COMMON Page I O of 14 Rev. 0
ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 2003022 1 NC-FP-AP.ZZ-0012 (Q)
TABLE 2 Ignitable Metals - Hope Creek Esubment 1Al E-201 1 ME-201 I B I E-201 1 B2E-201 1 A-E-2 1 7 1 B-E-217 I
A-E-I 08 I
B-E-I 08 1 C-E-108 Description SACS Heat Exchangers Tubing RACS Heat Exchangers Tubing Main Condensers Tubing Rx Fuel Channels Rx Fuel Spacers Rx Fuel Elements Tubing End Plugs Limitorque Actuators Motor Core Rotor Ignitable Metal Titanium Titanium Titanium Zirconium4 Zirconium -2 Magnesium NUCLEAR COMMON Page I?
of 14 Locations Reactor Bldg., El. 102' Room 4309 Room 4307 Reactor Bldg., El. 77' Room 421 1 Turb Bldg., El. 54' - 102' Rooms 131 0 to 1312 Reactor Bldg El. 201' Drywell Rev. 0
ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 2003022 1 NC.FP-AP.U-0012 (Q)
FORM I HOT WORK PERMIT Page I of 2 PART I:
GENERAL INFORMATION TYPE OF WORK:................................ 0 W 0 C 0 8 G c3 OFS OXY-ACENLENE RIG TO BE USED:....I7 YES NO LOCATION:....................................... SGS U/1 I7 SGS U/2 0 HCGS 0 YARD LOCATION DESCRIPTION:
ANNUAL PERMIT:.............................. 0 YES ORDER NUMBER:
JOB SUPERVISOR (PRINT):
NO PART 2: WORK ARA INSPECTION 0
0 0
0 0
0 Burning, cutting, and welding equipment is in good repair.
Ignitable metals have been considered (Le. titanium).
Floor has been swept clean of loose combustibles within 35 feet.
Watt, floor openings, cable trays and readily ignitable combustibles within a 35 foot radius which may be subjected to ignition sources have been covered.
Combustibles have been moved away from opposite sides of walls or ceilings.
Enclosed spacelequipment has been cleaned and purged of combustibles.
[CD-682D]
Fire extinguisher is available (obtain extinguisher from Firehouse/Storeroom).
Fire Watch will be provided during and 30 minutes after work completion to ensure work area is safe from fire danger.
Precautions required for protection of filtrationladsorption ventilation systems have been taken (per NAP-9). [CD-652A]
For electric arc welding, electromagnetic interference (EMI) precautions have been taken. (per NAP-5) [CD-I 56A]
Signature on page 2 indicates inspection was satisfactory and hot work activity may proceed following the acquisition of the DAILY hot work permit number.
NUCLEAR COMMON Page 12 of 14 Rev. 0
ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 20030221 NC.FP-AP.27-0012 (a)
FORM 3 HOT WORK PERMIT Page 2 of 2
~
PART 3: FINAL CHECKJOB COMPLETION Work areas and all adjacent areas have been inspected following job completion and no evidence of fire is present. All equipment has been returned to its proper location.
VERIFIED BY:
~
Job Supervisor Datd Time NUCLEAR COMMON Page 13 of 14 Rev. 0
KEY: W =Welding C = Cutting Hot Work Number NC.FP-AP.Z&0012(Q)
Work Location /
Type of Order Job Supv.
Hi Det.
FD Equip Involved work Number Ext Risk Zone Initials/
Time w, c,G, Yes B, OF$, OA X-No w,c,G, YeS B, OFS, OA X-No W,C,G, YeS B, OFS, OA X-No w IC,G, Yes B, OFS, OA X-No w,c,G, Yes B, OFS, OA X-NO W, C,G, Yes B, OFS, OA X-No W, C,G, YeS B, OFS, OA X-NO w IC,G, Yes B, OFS, OA X-No FORM 2 HOT WORK AUTHORlZATlON LOG B = Burning STATION:
OFS = Open Flame Soldering G= Grinding OA = Oxygen Acetylene unit NUCLEAR COMMON Page 14 of 14 Rev. 0
ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 2003022 1 PSEG Internal Use Only Page 1 of 2 PSEG NUCLEAR FIRE DEPARTMENT NC.NA-AP.ZZ-0025 (Q) - Rev. 6 OPERATIONAL FIRE PROTECTION PROGRAM Sponsor Organization: Emergency Services REVISION
SUMMARY
Biennial Review performed Yes X
No -.
I. This is a full revision to NC.NA-AP.ZZ-0025 (Q) - Rev. 5. (80035483).
- 2. Due to the extensive nature of the revision, change bars have not been used. This revision represents a significant editorial change with regard to content, formatting, process descriptions, and section numbering.
- 3.
This procedure has been revised to focus on high level process concepts, resulting in relocation of process description detail and associated forms, tables, and attachments to other lower tier administrative procedures.
- 4.
These changes are editorial. There is no reduction in program content or control, as 8 result of this revision.
- 5.
This revision relocates all Commitment Dowment numbers to the departmental administrative procedures implementing those commitments.
- 6.
Revised titles to reflect the current organizational assignments.
- 7. Replaced Nuclear Electric Insurance Limited (NEIL) with Nuclear Services Organization (NSO) throughout the procedure.
IMPLEMENTATION REQUIREMENTS:
Effective on: I /a?J03
- 4.
Fire Protection Impairment, Hot Work, and Transient Combustible Permits prOCeSS8d by the Fire Department under Revision 5 as of the issue date, do not have to be changed in the field to meet Revision 6 requirements.
Nuclear Outlook article and PSEG Nuclear L.L.C. lntranet website announcement describing the revision.
Provide a global electronic Power Point presentation to all PSEG Nuclear personnel.
- 2.
- 3.
APPROVED:
%rector.bc Buqness Support
ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 20030221 PSEG lntemal Use Only NC.NAWAP.ZZ-0025(Q), Revlslon 6 OPERATIONAL FIRE PROTECTION PROGRAM Revision Summary (Contlnued)
- 8.
Revised company name from PSE&G Nuclear Business Unit (NBU) to PSEG Nuclear.
- 9.
Removed detailed information on the Combustible Control Program. All detailed information is now found in NC.FP-AP.ZZ-O025(Q).
IO. Removed detailed information and forms for fire protection impairments. This information is now located in NC.FP-APZ-0025 (Q) and NC FP-APZ-0010 (Q).
1 1, Removed detailed information and forms for hot work operation and the control of ignition sources. This information is now located in NC.FP-AP.77-0025 (GI) and Page 2 of 2 NC.FP-AP.ZZ-0012(Q).
SECTION ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 2003022 1 NC. NA-APZZ-0025( Q)
OPERATIONAL FIRE PROTECTION PROGRAM TABLE OF CONTENTS I
.o 2.0 3.0 4.0 5.0 6.0 7.0 8.0 Common I_
TITLE PAGE 2
PURPOSE...........,.........._.._.
SCOPE.....,.................,,.,..,,..,..,.........,.,...,..,...,,.... _.. _....................._.............2 RESPONSIBILITIES.....,......,..,.,,.,,....~.............................~.~....,...................Z PROCESS DESCRIPTION...,...,..,............._.......................................-........4 PROCEDURE..................._.................................,........................... -........... 4
- 5. I 5.2 5.3 5.4 5.5 5.6 5.7 5.8 5.9 5.10 Response to Fire Emergencies..,.._.,_.......
.......... 4 Fire Prevention Requirements........................................................-. 5 Control of ignition Sources - Hot Work........_..._................................
5 High Risk Hot Work...................................,.._.._....._...........
............... 6 Combustible Material Control........................................................_.. 6 Control of Flammable and Combustible Liquids and Gases............. 6 6
Control of Portable Space Heaters.................._................................
7 Fire Protection Impairment Program..,......,.............,............._.
Compensatory Measure Fire Watch Program............_..................... 7 7
Ignitable Metals.._....,..,..............................,......._....._.
RECORDS,.....,............,..........,...,.,...................,...........................-............. /
DEFINITIONS..............................,.........................,...........,..,..............,......
8 9
REFERENCES....................-..-....,........,..,....-.......................................-..-...
8.1 General Documents..........._.._.._..
....................._....._.......................... 9 8.2 Cross References......................................................,,...,...........-.....9 Page 1 of 9 Rev. 6
4LL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 2003022 1 NC. NA-AP.ZZ-0025( Q)
I
.o 2.0 2.1 3.0 3.1 3.2 3.3 3.4 3.5 3.6 PURPOSE This procedure addresses the Operational Fire Protection Program (OFPP) for PSEG Nuclear.
SCOPE This procedure applies to all organizations and personnel engaged in activities which may affect the Operational Fire Protection Program (OFPP) for PSEG Nuclear.
RESPONSiBlLlTlES All Personnel Comply with the requirements of this procedure.
Job Supervisors - Salem and Hope Creek Stations Comply with fire prevention requirements of the Fire Protection Impairment, Hot Work, and Transient Combustible Control processes 0 Adhere to controls for ignition sources and flammable liquids and gases.
All Directors and Manacmrs Ensure their organizations implement the applicable fire prevention requirements within facilities they maintain and in which they conduct work.
Chemistrv Manager Ensure the performance of the Hope Creek Diesel Driven Fire Pump Fuel Oil Storage Tank sampling surveillance.
Operations Department Manacler - Salem Ensure the performance of the Diesel Driven Fire Pump Fuel Oil Storage Tank sampling surveillance 0
Ensure that personnel within their organization, designated as the Operation's liaison to the Fire Department (FD), are available as required, during fire fighting activities.
Operations Deoartment Manaser - Hope Creek Ensure that personnel within their organization designated as the Operation's liaison to the FD, are available as required, during fire fighting activities.
Common Page 2 of 9 Rev. 6
ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 2003022 1 NC.NA-AP.ZZ-O025(Q) 3.7 Radiation Protection Manaaet Provide required radiological monitoring when their personnel respond with FD personnel to fires located within the Radiological Controlled Areas (RCAs).
Work Control Center Personnel - Salem and HoPe Creek Operations Notify the FD when a fire protection systemharrier is being affected and when it is being returned to service.
3.8 3.9 Manaaer - Ememencv Services a
a a
0 e
Ensure General Employee Training (GET) discusses required employee actions in the event of fire.
Ensure a qualified, trained fire department is present on-site in accordance with NC.FP-AP.ZZ-O009(Q) Fire Protection Training Program.
Develop and maintain FD surveillance and test procedures.
Perform walk downs to monitor control of transient combustibles.
Authorize Hot Work, Transient Combustible, and Fire Protection impairment Permits (FP1Ps) and ensuring the aggregate impact is considered.
Maintain daily communication with the Operations Superintendent andor Control Room Supervisors on the status of Salem and Hope Creek Station FPIPs.
Ignition source control and monitoring for fire prevention.
Fire incident reports and investigations.
Interface with fire insurance inspectors, state and local enforcement agencies, and with offsite fire departments.
Define non-FD responsibility for required testing and inspection of fire protection equipment.
Performance of preventive and corrective maintenance on fire protection systems and components assigned to the FD.
Administration and control of the Fire Watch Program.
3.10 Vice President.- Endneering Establish programmatic requirements for implementing design related aspects of the Fire Protection and Fire Barrier Penetration Seal Programs.
a Development and review of Design Change Packages for controlling configuration of both the Fire Protection and Fire Barrier Penetration Seal Programs.
e Ensure a member grade Society of Fire Protection Engineer (SFPE) person exists within the organization, [CR96M1621 I]
Page 3 of 9 Rev. 6 Common
ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 20030221 NC. NA-AP.U-002qQ) 3.11 4.0
- 4. I 4.2 4.2.1 4.2.2 4.2.3 4.2.4 4.2.5 4.2.6 4.2.7 4.2.8 5.0 5.1 5.1.1 Vice President - Operations 0
Ensure trained maintenance personnel perform preventive or corrective maintenance of Fire Protection systems in a prompt and effective manner.
Ensure performance of preventive maintenance surveillances for fire protection components at both the Salem and Hope Creek Stations, as agreed upon with the Manager - Emergency Services.
0 Ensure PSEG Nuclear facilities are maintained in compliance with the applicable state fire codes and insurance regulations.
PROCESS DESCRIPTiON This procedure identifies elements of the Operational Fire Protection Program categorized as Fire hventlon and Readiness to Detect and Suppress Fires.
This procedure is implemented in full detail by the following Fire Department administrative procedures:
NC.FP.AP.Z4025(Q); Precautions Against Fire SC.
FP.AP.ZZ-0003(Q), Actions for Inoperable Fire Protection - Salem HC.FP-AP.ZZ-O004(Q), Actions For Inoperable Fire Protection-Hope Creek NC.FP-AP.ZZ-0005(Q), Fire Protection Surveillance I Periodic Test Program NC.FP-AP.z7-0009Q), Fire Protection Training Program NC. FP-APZ-001 O(Q), Fire Protection Impairment Program NC.F P-AP.ZZ-OOIZ(Q), Safe Hot Work Practices NC. FP-AP.ZZ-0020(Q), Compensatory Measure Fire Watch Program PROCEDURE Response to Fire Emeraencies Refer to NC.FP-AP.ZZ-0025(Q) (FPAP-25) for detailed instructions on response to fire emergencies. Key concepts are reviewed here.
0 In the event a fire alarm sounds within an office facility, evacuate the building in accordance with the established routes. Accountability should be performed.
NEVER use the elevator during a fire emergency building evacuation.
e For fire alarms within the Salem and Hope Creek Stations, listen to the instructions via the page system provided.
Common Page 4 of 9 Rev. 6
5.1.
I (Continued)
Upon discovering a fire, your first duty is to report the fire. The Fire Department will respond with personnel and equipment to combat the emergency.
a The communication systems vary between the Salem and Hope Creek stations. Know the differences between the plants. Refer to FPAP-25.
a After reporting a fire, employees should fight the fire if its within their training and capabilities. If not, they should alert others in the area, evacuate to a safe distance and standby to direct FD personnel to the scene.
5.2 Fire Prevention Reauirements 5.2.1 5.2.2 Refer to NC.FP-AP.ZZ-O025(Q) (FPAP-25) for detailed instructions on fire prevention work practices and requirements.
Some key concepts are briefly reviewed here.
e 1
e 1
Fire Doors shall be maintained closed,. except for normal passage, unless provided with an approved permit It is imperative that ALL PERSONNEL make sure fire doors close behind them.
Fire Protection deficiencies observed should be immediately reported to the FD (Ext. 2800/2803) for evaluation.
Restrictions apply to combustible materials introduced into any safety-related aredroom.
Restrictions apply to locating combustible materials in the vicinity of permanent bu i Idings.
Restrictions apply to use of fire protection water for non-fire protection purposes.
Storage of material should not affect access to, or obstrud, fire protection systems or fire fighting equipment.
Fire extinguishers mounted in the stations are for emergency use only.
Clearance, free of combustible material, shall be maintained around energized electrical panels.
Extension cords and flexible cords shall not be a substitute for permanent wiring.
5.3 5.3.1 5.3.2 Control of lanition Sources - Hot Work Work involving ignition sources, such as welding, cutting, burning, grinding or open flame soldering, is considered hot work.
A Hot Work Permit will be required to control these ignition sources.
~
ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 2003022 1 5.3.3 5.4 5.4. I 5.4.2 5.4.3 5.5 5.5.1 5.5.2 5.6 5.6. I 5.7 5.7. I 5.7.2 Plant work involving ignition sources, such as welding, cutting, burning, grinding or open flame soldering is considered a fire watch required activity.
An exception to this requirement is allowed for work performed in shop areas.
Hot work permits are required for all hot work activities.
Hlqh Risk Hot Work High risk hot work is a PSEG Fire Department term applied to work with well defined and special hazard criteria.
This type of hot work is distinguished from non-high risk hat work.
High Risk hot work still involves ignition sources, such as welding, cutting, burning, grinding or open flame soldering.
High risk work involves:
Vessels and systems that contain, or had contained, combustible or flammable chemicals, solvents, or gases.
0 Work involving ignitable metals.
0 Work involving processes that produce copious amounts of heat with molten slag and sparks. For example, current industrial technology may employ equipment such as plasma arc cutting machines. This technology produces copious amounts of energetic slag and sparks.
Special and additional controls apply to high risk hot work. These controls are covered under NC.FP-AP.ZZ-O012(Q), Safe Hot Work Practices.
Combustible Material Control General Use and Controls - AH areas:
0 Combustibles Material will be controlled IAW NC.FP-APZ-0025 (a).
for Safety Related Areas Only:
Storage of combustibles is NOT permitted in Salem and Hope Creek Stations safety related areaskooms except when approved by Engineering.
Control of Flammable and Combustible Liauids and Gases Detailed guidance on the handling and control of flammable and combustible liquids and gases is provided in NC.FP-AP.ZZ-O025(Q).
Control of Portable SPace Heaters The use of portable space heaters is prohibited in safety related areas except as indicated IAW NC.DE-AP.ZZ-0030 Q)., Control of Temporary Modifications.
Space heaters in non-safety related areas are controlled IAW NCFP-AP.ZZ-0025 (GI),
Common Page 6 of 9 Rev. 6
ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 20030221 5.0 5.8.1 5.8.2 5.8.3 5.8.4 5.8.5 5.9 5.9.1 5.9.2 5.10 5.10.1 5.1 0.2 5.10.3 6.0 6.1 6.2 6.3 Fire Protection lmpalrrnent Proaram FD is the fire protection impairment authority.
Impairments are formal controls that recognize, evaluate, and administratively approve and track either planned actions or emergent conditions that impact the operability of the fire protection system or a fire protection component.
Job supervisors prepare and submit impairment requests for work activities that will place fire protection equipment or system in an inoperable condition.
Nuclear fire protection supervisors (NFPS) evaluate impairment requests and emergent system and equipment issues to identify the full scope, or aggregate impact, of the issue.
NFPS identifies the appropriate compensatory measures and generates an impairment, accordingly.
Prior to issuing impairments, the NFPS notifies and reviews the impairment issue with the Operations Superintendent or Control Room Supervisor (OSICRS), of the affected station, and obtains OS/CRS-approval to issue the impairment,.
Cornpensatow Measure Fire Watch Program Adequate compensatory measures are required to be implemented to supplement inoperable fire protection systems and equipment Procedures SC.FP-AP.ZZ-O003(Q) and HC. FP-AP.ZZ-0004(Q) identify those systems and components requiring compensatory actions to be taken when a fire protection system or component is determined to be inoperable.
lanitable Metals The metals titanium, zirconium, and magnesium represent a special combustible material hazard class that warrants extra care and control.
These metals are classified as ignitable metals. These are metals that can burn readily, vigorously, and with a significant release of heat Controls relative to working on or in the vicinity of ignitable metals are provided in procedure MC.FP-AP.Z74012(Q), Safe Performance of Hot Work.
RECORDS Hot Work Permits and Transient Combustible Permits shall be retained with the work package in accordance with NC.NA-AP.ZZ-0011 (Q), Records Management Program (NAP-11).
The FD, in accordance with NAP-11 shall retain transient combustible permits and logs pertaining to hot work permits and fire protection impairment permits for a minimum of six months.
FPlRs that were used to create FPlPs should be retained with work packages. The FD maintains a hard copy log of all FPlPs generated from FPIRs.
Common Page 7 of 9 Rev. 6
ALL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 20030221 NC.NA-APJZ-O025( Q) 7.0 7.1 7.2 7.3 7.4 7.5 7.6 7.7 7.8 7.9 7.10 7.1 I 7.12 7.13 7.14 DEFINITIONS Combustible Liquid - Liquid heving a flash point at or above I00 deg-F (37.8' C).
Combustible Material - Material, which in the form it is used and under the conditions anticipated, WILL ignite, burn, support combustion, or release flammable vapors when subjected to fire and heat.
Fire Barrier - Those construction elements (walls, floors, and their supports),
including beams, columns, penetration seals or closures, fire doors and dampers, that are rated by approving laboratories in hours of resistance to fire and used tu prevent the spread of fire.
Fire Watch - An individual who has satisfactorily completed fire watch training and is designated by the FD for compensatory measure, hot work, of both fire watch type duties.
Flammable Liauid - A liquid with a flash point below 100 deg-F and a vapor pressure not exceeding 40 pounds per square inch absolute at 100 deg-F.
Also known as a Class 1 liquid.
Hot Work -Work that involves ignition sources.
lanition Source - Heat or flame from any source capable of igniting combustible or flammable material.
impairment - Any condition affecting the intended operation of a fire protection system or barrier or posing as a fire hazard within the stations.
Noncombustible Material - Material which, in the form it is used and under the conditions anticipated, will NOT ignite, burn, support combustion, or release flammable vapors when subjected to fire and heat.
Permanent - A condition with an indefinite period of time.
Safety Related Area - An area that contains systems and components required to shut down the reactor, mitigate the consequences of postulated accidents or maintain the reactor in a safe cold shutdown condition.
Storaqe - The placing of items in a location and leaving them unattended until later use or disposal.
TemDorary - planned time duration of no more than 60 days.
Transient Combustible - Any combustible material introduced into a Safety-Related arealroom which will be used to complete an assigned task or left unattended for any period of time. It does not include items moved through areas in route to the final destination, unless the items are lef? un-monitored during transit.
Page 8 of 9 Rev. 6 Common
4LL ACTIVE ON-THE-SPOT CHANGES MUST BE ATTACHED FOR FIELD USE 2003022 1 8.0 8.1 8.1.1 8.1.2 8.1.3 8.1.4 8.1.5 8.1.6 8.t.7 8.1.8 8.1.9 8.1.10 8.1.1 I 8.1.I2 8.1.13 8.1.I4 8.1.I 5
8.1.16 8.2 8.2.1 8.2.2 8.2.3 8.2.4 8.2.5 8.2.6 8.2.7 8.2.8 8.2.9 8.2.1 0 NC.NA-AP.ZZ-O025( Q)
REFERENCES General Documents Salem Updated Final Safety Analysis Report (UFSAR), Section 9.5, - Fire Protection System Salem Fire Protection Report - Fire Hazards Analysis HCGS Updated Final Safety Analysis Report, (UFSAR) Section 9.5.1, - Fire Protection Program HCGS UFSAR, Appendix 9A, - Appendix R Comparison Appendix A to Branch Technical Position APCSB 9.5-1, Guidelines for Fire Protection for Nuclear Power Plants Docketed Prior to July I, 1976 Branch Technical Position CMEB 9.5-1, Guidelines for Fire Protection for Nuclear Power Plants NRC Supplemental Guidance, Nuclear Plant Fire Protection Functional Responsibilities, Administrative Controls and Quality Assurance, August 1977 Occupational Safety and Health Standards Title 29, Chapter XVII, Part 191 0.36FR 10466 10CFR50 Appendix R, Fire Prot. Program for Nuclear Power Facilities Operating Prior to Jan. 1, 1979 NC. DE-PSZ-0001 (Q), Fire Protection Programmatic Standard IEEE Standard 634 - I978 Hope Creek Station Generic Letter 86-1 0 Submittal, dated May 13, 1986 HC.DE-PS.ZZ-0021 (F), Hope Creek Penetration Seal Program SC.DE-PS.Z-O035(Q), Salem Penetration Seal Program HC.FP-EO.ZZ-0001 (Z), Control Room Fire Response - HCGS SC.FP-EO.=-0001 (Z), Control Room Fire Response - Salem Cross
References:
NC.NA-AP.ZZ-001 1 (Q), Records Management Program NC.C)E-AP.ZZ-O030(Q), Control of Temporary Modifications SC.FP.AP.ZZ-0003(Q), Actions for Inoperable Fire Protection - Salem HC. FP-AP.ZZ-O004(Q), Actions For Inoperable Fire Protection-Hope Creek NC.FP-AP.ZZ-0005(Q), Fire Surveillance and Periodic Test Program NC.FP-AP.ZZ-O009(Q}, Fire Department Training Program NC.FP-AP-ZZ-001 O(Q), Fire Protection Impairment Program NC.FP-AP.ZZ-0012(Q), Safe Performance of Hot Work NC.FP-AP.U-O020(Q), Compensatory Measure Fire Watch Program NC. FP.AP.ZZ-O025(Q); Precautions Against Fire Common Page 9 of 9 Rev. 6