ML18022A904
| ML18022A904 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 10/12/1992 |
| From: | Mccarthy D CAROLINA POWER & LIGHT CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML18010A852 | List: |
| References | |
| NLS-92-284, NUDOCS 9210280201 | |
| Download: ML18022A904 (157) | |
Text
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~ REGULAT(l INFORMATION DISTRIBUTION STEM (RIDE) i ACCESSION NBR'9210280201 DOC DATE'2/10/12 NOTARIZED'O DOCKET FACIL:50-400 Shearon Harris Nuclear Power Plant, Unit 1 Carolin 05000400 AUTH. NAME AUTHOR AFFILIATION MCCARTHY,D.C. Carolina Power 6 Light Co.
RECIP.NAME RECIPIENT AFFILIATION Document Control Branch (Docume Control Desk)
SUBJECT:
Forwards Modification Package for PCR-6547,Basis for Setpoint Calculation, including scaling, summary of evaluation l
of pipe stresses,CP6L position on TS requirement, info planned surveillance testing.
I D
DISTRIBUTION CODE: A001D TITLE: OR COPIES RECEIVED:LTR Submittal: General Distribution i ENCL SIZE: ISIS y NOTES:Application for permit renewal filed. 05000400)
RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD2-1 LA 1 1 PD2-1 PD 1 1 I
LE,N 2 2 0
INTERNAL: ACRS 6 6 NRR/DET/ESGB 1 1 NRR/DOEA/OTSBll 1 1 NRR/DST/SELB 7E 1 1 NRR/DST/SICBSH7 1 1 NRR/DST/SRXB 8E 1 1 NUDOCS-ABSTRACT 1 1 O~C- -.8MB 1 0 OGC/HDS1 1 0 EG FIL 01 1 1 RES/DSIR/EIB 1 1 EXTERNAL: NRC PDR 1 1 NSIC 1 1 NOTE TO ALL "RIDS" RECIPIENTS:
PLEASE HELP US TO REDUCE WASTE! CONTACT THE DOCUMENT CONTROL DESK.
ROOM P 1-37 (EXT. 504-2065) TO ELIMINATEYOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!
TOTAL NUMBER OF 'COPIES'REQUIRED: LTTR '22 ENCL 20
~ l' C$ QE, Carolina Power 8 Light Company OCT 12 'L992 SERIAL: NLS-92-284 United States Nuclear Regulatory Commission ATTENTION: Document Control Desk Washington, DC 20555 SHEARON HARRIS NUCLEAR POWER PLANT DOCKET NO. 50-400/LICENSE NO. NPF-63 HIGH HEAD SAFETY INJECTION ALTERNATE MINI-FLOW SYSTEM MODIFICATION Gentlemen:
As a result of the meeting on October 7, 1992 between CP&L and the NRC Staff on the modification to the Shearon Harris Nuclear Power Plant (SHNPP) High Head Safety Injection Alternate Mini-Flow System, the NRC Staff requested that CP&L provide the following specific items for NRC review:
- 1. Modification Package for PCR-6547;
- 2. Basis for Setpoint Calculation, Including Scaling;
- 3. Motor-Operated Valve Design Basis Review, Set-Up Criteria, and Differential Pressure Test Results;
- 4. Information on Flow Test Equipment Accuracy and Calibration Technique;
- 5. Pipe Support Calculations for 3 Redesigned Anchors and Summary of Evaluation for 29 Supports;
- 6. Summary of Evaluation of Pipe Stresses; 7 ~ Information on Planned Surveillance Testing, Calibration Process for Logic Modification; and
- 8. CP&L Position on Technical Specification Requirement Items 3, 5, 6, and 7 were transmitted to the NRC on October 9, 1992 by CP&L letter NLS-92-282., The purpose of this letter is to transmit the remaining items: 1, 2, 4, 8, and a revised item 3. These are included as Enclosures 1 through 5, respectively.
Should you have any questions about this information, please contact me at (919) 546-6901.
s very tr David C. McCarthy Manager Nuclear Licensing Section LSR/j bw Enclosures cc: Mr. S. D. Ebneter Mr. N. B. Le Mr. J. E. Tedrow 92 f 0280@0Qff
~ Fayetteville Street 4 P. O. Box 1551 I Raleigh. N. C. 27602 92fof2
PDR ADOCK 05000400 (1791HNP)
PDR,
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.9210280201 .
ENCLOSURE 1 SAFETY INJECTION ALTERNATE MZNZFLOM MODIFICATION PACKAGE
Design Package Mod. No. PCR-6547 Cover Sheet Field Rev. No. 0 DESIGN PACKAGE
Mod. No.. PCR-6547 .
Design Package Field Rev. No. 0 List of Effective Contents
~ET~IN V I D TOTAL CONTENTS CONTROL Effective Contents List 0 DESIGN PACKAGE Design Package-Mech./I&C 0 81 Design Package-Electrical 0 125
Design Package Mod. No. PCR-6547 Cover Sheet Field Rev. No. 0 DESIGN PACKAGE-Mech/IAC
Mod. No. PCR-6547 Design Package Field Rev. No. 1 List of Effective Contents CONTENTS CONTROL-Mech/I&C Effective Contents List DESIGN BASIS Design Basis References/Revisions DESIGN IMPACT EVALUATIONS Design Impact Summary Design Impact Statements DESIGN SUPPORT DOCUMENTS Calculations Alternate MiniQow Orifice Strainer 17 CSIP Alternate Miniflow Interlock Accuracy Calc. 24 Max. RCS Pressure for CSIP Min. Flow 19 Strainer Shielding Calculation 71 Minimum Wall Check on Strainer Housing 1 Setpoint Worksheets 2 DESIGN DOCUMENTATION REVISIONS Drawing Revision Sheet Design Document Revision Sheet MEQ Document Revisions SELF-ASSESSMENT RECORDS ALARAPre-Design Walk-Down Record Comment Resolution For ALARAPre-Design Walkdown Discipline Design Verification Record I&C Environmental Qualification Mechanical Civil Discipline Technical Review of Completed Design Package Fire Prot. 1 I&C 1 Environmental QualiQcation 2
~ ~ a a ~ ~ ~
Mod. No. PCR~7 Design Package Field Rev.
No. 1 List of EKective Contents (continued)
PAGES Mechanical 2 Civildtructural/Civi14tress 4 Inter-Discipline Review Requests(IRRs) 12 Appendix R 4
Design Package Mod. No. PCR-6547 Design Basis Field Rev. No. 0 Cover Sheet DESIGN BASIS
Design Package Mod. No. PCR-6547 Design Basis Field Rev. No. 0 Desi n Basis References/Revisions Pa e No. 1 1.0 DESIGN BASIS DOCUMENT FOR SAFETY INJECTION SYSTEM DBD f104 WAS REVIEWED FOR FOR IMPACT AND THE NECESSARY CHANGES ARE ATTACHED HEREIN.,
MOD. NO. PCR-6547 FIELD REV. NO. 0 PAGE NO.
DBD-104 SAFETY INJECTION SYSTEM To ensure that the Charging Pumps are protected in the event that RCS pressure exceeds the design shutoff head of the pump during injection, a separate ECCS minimum flow path has been added. The motor operated valves that normally isolate this path are opened by a safety injection signal coincident with high RCS pressure. Xf RCS pressure approaches the shutoff head of the pumps, the isolation valves open and provide sufficient flow to prevent pump damage. These isolation valves will close as the RCS depressurizes and in response to a safety injection signal to provide maximum injection
Design Package Mod. No, PCR-6547 Design Impact Ev'aluations Field Rev. No. 0 Cover Sheet DESIGN IMPACT EVALUATIONS
Design Package Design Impact Summary Rev. No.~
Mod. No. P R M7 DESIGN IMPACT
SUMMARY
The disciplines/specialty groups recorded below have design impacts which affect this Modification. An appropriate Design Impact Evaluation is attached for each of the affected disciplines/specialty groups.
Total Number of Evaluation Pa es Mechanical f)
HVAC [I [X]
f) pq Electrical pq [1 I&,C f]
Civil/Structural Seismic Appendix R Environmental Qualification pq pq pq pq
[X'/A
[l fl fl fl Wee se~Alh- w'Ls>
Human Factors [I [X]
Materials f]
+ee4e<<'A~e- <<3/oo<<<<,~ p(~~
NPMP REV- 4
Design Package Mod. No. PCR-6547 Design Impact Evaluations Field Rev. No. 1 Discs line Desi n Im act Statement Pae '
2.0 This modification removes the Alternate Mini-flow Relief Valves and installs restricting orifices upstream of the motor-operated isolation valves (1CS-746 & 1CS-752). The restricting orifices will ensure that the Charging Pumps are protected in the event that RCS pressure exceeds the design shutoff head of the pump during injection by allowing a 60 GPM flow back to the Refueling Water Storage Tank. Strainers will be installed upstream of the orifices to prevent 'clogging of the orifices. Calculation performed on strainer to ensure insignificant pressure drop.
The restrictive orifices are sized for a nominal flow of 60 gpm. A bench flow test will be performed at Wyle Test Laboratory to furnish a capacity performance curve of the orifice. This performance curve can be compared to the ultrasonic flow acceptance test results.
Pacific Pump recommended 60 gpm as the minimum flow for testing during normal operation. A limit of 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> per month was specified to ensure that pumps will operate continuously in the charging mode without maintenance between refueling outages. For the short duration of injection in the event of a LOCA, smaller flows can be tolerated.
A calculated performance curve for the orifice was provided by the manufacturer and the calculated flow at 6000'ead is just under 60 gpm. A tolerance was applied to the test acceptance value in case the ultrasonic flow meter erred on the low side. Past flow tests on these identical-orifices measured 60 gpm at 6000'ead pressure. The ultrasonic flow meters have a calibrated accuracy of +/-3/.
If flow through the orifices measures somewhat less than 60 test, the pump will not be damaged, since flow to the RCP gpm during the seals will be aligned. By the same token, during normal operation Reactor Coolant Pump seal injection will be maintained to ensure total CSIP flow exeeds 60 gpm during safety injection.
Protection of the weak pump has been demonstrated in the Mechanical Safety Analysis.
Design Package Mod. No. PcR-6547 Design Impact Evaluations Field Rev. No. 0 Discipline Design Impact Statement Page ISC Deci line Desi n Im act Statement This modification involves the introduction of an active pressure control system into the CSIP alternate mini-flow subsystem. This is accomplished by adding RCS pressure permissives in series with safety injection actuation logic to provide automatic control of motor operated isolation valve 1CS-746 and 1CS-752. Involved in this design change are modifications to protection cabinets 1 6 4, and train A 6 B of SSPS output bay 2.
The changes introduced by the subject modification represent a means for improving the existing alternate mi.ni-flow subsystem which has experienced integrity inadequacies. The design changes and subsequent effectes to equipment/system operability have been evaluated and found to be acceptable for applicational compatibilty, environmental congeniality, and equipment integrity. Results of a safety evaluation show that no detrimental affects to plant safety are introduced by the aforementioned design changes.
I
Design Package Mod. No. PCR-6547 Design Support Documents Field Rev. No. 0 Cover Sheet DESIGN SUPPORT DOCUMENTS
CA'ROLIHA ~HER & LIGHT COMPANY FORM 3 NUCLEAR PLANT [ ) BNP UNIT NUMBER PROJECT NUMBER MODIFICATION NUMBER MODIFICATION [X) HNP 1 RET-P-6547 PCR-6547 TRAVBLSR [ ) RNP ABSTRACT TITLE: [X] MODIFICATION
[ ] EMERGENCY MOD REASON FOR MOD: [ ] DOCUMENT CHANGE ONLY SYSTEM NUMBER(S):
QUALITY CLASSIFICATION: YES NO IMPACT:
[X] A. Q.LIST OR AFFECTS Q-LIST [ ] [X] UNREVIEWED SAFETY QUESTION B. REG. GUIDE 1.29 OR 1.97 [ ] [X) TECH SPEC CHANGE C. RADWASTE-Q [X) [ ] FSAR CHANGE D. FIRE PROTECTION-Q [ ] [X] SIGNIFICANT ENVIRONMENTAL IMPACT E. NON.Q [ ] [X] MAJOR RADWASTE MODIFICATION OTHER DESIGN REVIEWS/APPROVALS LEAD ENGINEER DAT
[X] DESIGN VERIFICATION 7
[ ] TECHNICAL DA E 10CFR50.59 REVIEW (See Safety Review Package)
QUALITY ASSURANCE DA E DESIGN RELEASE PRINCIPAL ENG/ENG SUPERVISOR ZS '7t-DATE PLANT REVIEWS/APPROVALS PLETED E
[X] [ ] ALARA 2
[x) [ ) E&RC FIRE PROTECTION 7 q-zS.~" (.( INSTALLATION 0 ?
[X]. ] ISI
[X] ] .MAINTENANCE - ELECTRICAL
[x] [. ] MAINTENANCE - MECHANICAL
[X] OPERATIONS/HUMAN FACTORS /a 9 f
[X] PNSC
[ ) [X) SECURITY
[X] SYSTEM ENGINEER
[X] [ ] TRAINING/SIMULATOR
[ ] [X] OTHER
[X] REVIEW COORDINATOR -z-ez.
/~o-Z- 4 INDEPENDENT SAFETY REVIEWS
[ CNSR (Prior to Implementation)
[X] {' CNSR (Review documented on Closeout Sheet - NPMP Form 15)
[ ] [X] NRC (Prior to Implementation)
Q ((((ORKING copy
NPMP - REV. 4
CA'ROLINA POWER 8 LIGHT COMPAHY FORM 3(CON'T)
NUCLEAR PLANT [ ] BNP UNIT NUMBER PROJECT NUMBER MODIFICATION NUMBER RET-P-6547 PCR.6547 TRAVELER(CON'T) [ ] RNP DESIGN ORGANIZATION INTERNAL APPROVALS Signatures below indicate that the appropriate areas of concern for the listed discipline/specialty group have been satisfactorily incorporated into the above document.
MECHANICAL [X] e/~/~<
ELECTRICAL [X] S a-I&C [x]
CIVIL/STRUCTURAL [X]
~a~ J >razes SEISMIC [x]
MATERIALS [X]
APPENDIX R [X]
ENVIRONMENTAL QUAL. [X] -25 Z NPMP - REV. 4
Installation Package Mod. No. P PCR-6547 List of Effective Pages Field Rev. No. 0 Page No. Al
~pa e No. ~ev. ~ae No. Rev. ~aetio. ~ev ~ae No. ~ev.
Al C29 0 D23 0 E16 A2 C30 0 D24 o~
0 E17
'l C31 0 D25 0 C32 0 D26 P~o 6-7 O Bl C33 0 D27 0 Fl 0 B2 C34 0 D28 0 F2 0 B3 C35 0 D29 0 F3 0 B4 C36 0 D30 0 O C37 0 D31 0 0 C38 0 D32 0 GLS~G2 G1A 0~~
C39 0 D33 P 0 Cl 0 C40 0 D34 0 G3 0 C2 0 C41 0 D35 0 G4 0 C3 0 C42 ~ 0 D36 0 GS 0 C4 0 C43 0 D37 0 G6 0 C4A 0 C44 0 D38 0 C4B 0 C45 0 D39 0 Hl C4C 0 C45A 0 D40 HX C4D 0 C46 0 D41 0
R>A, C47 0 D42 C4E'4F 0 C48 0 D43 HSA C5 0 C49 0 D44 HR C6 0 Cs4 C5 D45 H>
C7 0 D46 Hv C8 0 Dl 0 D47 H C9 0 D2 0 D48 Hah
,
Clp Cll C12 0
0 0
'4 D3 D5 0
0 0
D49 D50 D51
'HS H I C13 0 D6 ~
0 D52 I'g [05 C14 0 D7 0 C15 0 D8 0- El v'L (l A C16 0 D9 0 E2 C17 0 Dlp 0 E3 Hh>
C18 0 Dll 0 E4 H l5 C19 0 D12 0 E5 High C20 0 D13 0 E6 C21 0 D14 0 E7 H (<A C22 0 D15 0 E8 t5 C23 0 D16 0 E9 C24 0 D17 0 Elp HiaO C25 0 D18 0 Ell H lac C26 0 D19 P E12 C27 0 D20 0 ,
E13 C28 0 D21 0 E14 D22 0 E15
Installation Package Mod. No. PCR-6547 Table of Contents Field Rev. No. 0 Page No. A2 SECTION DESCRIPTION CONTENTS CONTROL List of Effective Pages Table of Contents B PROJECT
SUMMARY
Problem and Scope Recommended Solution Alternatives Considered INSTALLATION SUPPORT DOCUMENTS Quality Classification Evaluation Safety Review Attachment List Bill of Materials Spare Parts List INSTALLATION DRAWINGS Drawing List Connector List Instrument Data Sheets Component Level Q-List Form Equipment Data Base Forms Valve List Change Forms INSTALLATION INSTRUCTIONS Special Installation Instructions Penetration Breach Form TESTING REQUIREMENTS Accepatance Test PLANT DOCUMENTATION REVISIONS Plant Document Revision Sheet FSAR Changes H PLANT COMMENTS
Installation Package Mod. No. PCR-6547 Project Summary Field Rev. No. 0 Pa e No. Bl SECTION B PROJECT
SUMMARY
Installation Package Mod. No. PCR-6547 Project Summary Field Rev. No. 0 Pa e No. B2 1.0
~
Original - The High Head Safety Injection Alternate Mini-flow Relief valves may have caused water hammer in the Charging/Safety Injection piping and may have caused piping/valve damage. The piping configuration does not contain a high point vent directly under the relief valve.
1.2 Evolution to present - RET-P-5630 was released to NED for resolution.
2.0 2.1 History/Root Cause - In May 1991, the High Head Safety Injection Alternate Mini;flow Relief valves were determined to be inoperable.
This placed the plant in a common mode failure event which affected both trains of high head safety injection. The event was investigated and resolved. The root cause as identified in LER 91-008-01 was believed to be water hammer. The piping configuration of the subsystem does not contain a high point vent directly under the relief valve.
The elevation difference between the inlet pipe and the relief valve inlet is approximately 1 foot. After maintenance, this allows an air void to form between the relief valve and the motor operated isolation valve. At that time, the relief valve header was required to be
'manually filled (after relief valve maintenance) as a corrective action.
Prior to initial startup in 1985 the NRC issued IE Bulletin 85-03, "Motor-Operated Valve Common Mode Failures During Plant Transients Due to Improper Switch Setting". The bulletin was interpreted to require Harris Technical Support to develop and conduct a test to demonstrate compliance with the IE Bulletin. During this performance of this test a water hammer was developed during the opening of the inlet motor operated valves (MOVs) to the Alternate Mini-flow Relief Valves.
Subsequent manual operation of the individual MOVs exhibited no water hammer. Once the air was purged from the piping, no water hammer occurred.
The required ASME Section XI set pressures for the Alternate Mini-flow Valves is 2300 a69 psig. Historically, set point drift has been a problem with these valves. As an. example, during Refueling Outage g2, valve 1CS-744 was tested and found to lift at 2210 psig. (Originally set at 2310). this was 21 psig below the minimum required set point and amounted to a drop of 100 psig from the as left set point.
The proposed resolution will remedy this problem.
2.2 General Description - This modification will remove the Alternate Mini-flow Relief Valves and install restricting orifices upstream of the motor-operated isolation valves (1CS-752 6 1CS-746). The restricting orifices will ensure that the Charging Pumps are protected in the event that RCS pressure exceeds the design shutoff head of the during injection by allowing a 60 GPM flow back to the Refueling 'ump Water Storage Tank (RWST). Strainers will be installed upstream of the orifices to prevent clogging of the orifices. A blind flange will be mounted to the top of the strainers for venting/accessibility. Lead will be installed on strainers in case of crud buildup 'hielding inside strainers.
This modification introduces changes to the operating logic of CSIP alternate mini-flow isolation valves 1CS-746 and 1CS-752. The logic of
Installation Package Mod. No. PCR-6547 Project Summary Field Rev. No. 0 Pa e No. B3 these motor operated valves will be modified such that the valves will open upon high RCS pressure coincident with an "S" signal.
pressure approaches the shutoff head of the pumps, the isolation If RCS valves will open and provide sufficient flow to prevent pump damage.
Additionally, these isolation valves will close as the RCS depressurizes and in response to a safety injection signal to provide maximum injection flow. This will be accomplished by adding bistable circuitry to RCS wide range'pressure loops P-402 and P-403 'omparator cards (NAL) and solid state relay circuitry (NAS) will be added to protection cabinets 1 & 4. These bistables will energize/de-energize
'rotary relays (K711-A S K711-B) within the SSPS output bays. Contacts of these SSPS relays will be installed in series with contacts of safety injection relay K740 to provide automatic valve control.
The subject isolation valves presently receive an "S" signal from relay K636 located in SSPS output bay 1. This relay is manually reset early in the transient as directed by the emergency operating procedures. If this signal is reset prior to RCS pressure increasing to 2300 psig, the mini-flow valves may never open. Due to this concern, this modification will substitute a RWST-SI signal, which is not reset until the normal charging header is, aligned, for the present SI signal. This design change involves removing the K636 relay from valve circuitry and utilizing the"K740 relay (located in SSPS output bay 2) for the safety injection permissive.
2.3 Major Equipment - There is no major equipment specified for this modification. All items are in stock and have been placed on reserve.
See Bill-of-Material.
2.4 Control Features - Design changes introduced by this modification will include the addition of a high RCS pressure permissive coincident with the "S" signal into the isolation valve opening circuitry.
This design change also involves removing the K636 relay from valve circuitry and utilizing the K740 relay (located in SSPS output bay 2) for the safety injection permissive.
2.5 System Operations - No significant system operation changes.
2.6 Unit Performance - Unit performance is unaffected by this change.
2.7 Plant Impact - This change will be done during RFO g4.
2.8 ALEQVl - Radiation field in the area is 35 mR/hr. Based on a plant ALARA Group estimate, the estimated installation dose will be 4 Man-Rem.
3.0 3.1 The following design changes to the operating logic of these isolation valves have been explored to alleviate relief valve water hammer
.concerns and to maintain dead head protection:
(gl) CSIP discharge flow- This option involves the addition of reduntant flow loops which would provide an isolation valve opening scenario upon CSIP low discharge flow coincident with a "S" signal and pump running status. The prudence of this alternative is questionable due to the extent of logic modifications and doubtful cost effectiveness.
(g2) CSIP discharge pressure- This option introduces a CSIP high discharge pressure permissive to initiate isolation valve opening
Installation
~
Package Mod. No. PCR-6547 Project
~
Summary Field Rev. No. 0 Pa e No. B4 coincident with the safety injection signal. In addition to the introduction of safety-related discharge pressure instrumentation, dedicated train selectability would be required for CSIP 1C operability.
(g3) Alternate mini-flow line pressure- It was proposed to install ressure instrumentation immediately upstream of isolation valve 1CS-52 & 1CS-746. The two pressure devices (nne per protection path) will be interlocked to close the block valves upon low mini-flow line pressure. This low pressure permissive we'll provide dead head protection of the CSIPs and will allow for the elimination of the AMF safety relief valves. In lieu of the relief valves, orifice plates can be installed in non-class piping downstream of the block valves. A multi-port flow orifice in the code boundary of the piping was chosen instead.
Installation Package Mod. No. PCR-6547 Installation Support Documents Field Rev. No. 0 Pa e No. Cl SECTION C INSTALLATIONSUPPORT DOCUMENTS
Installation Package Mod. No. PCR-6547 Quality Classification Evaluation Field Rev. No. 0 Pa e No. C2 1.0 The orifices and strainers are being added to the Safety Class 2 portion of the Safety Injection piping. Since the relief valves are being removed, the Safety Class break will be moved back to valves 2CS-V758SB-1 &2CS-V760SA-1.
All valve logic changes will be made to safety related circuitry.
Based on the above, this modification is Quality Class "A".
REVZS ION 3 10CPR50. 59 PROGRAMÃ h MQAJM'TTACHMENT CP6LL SAFETY REVIEW PACKAGE Page 1 of 8 SAFETY REVIEW COVER SHEET DOCUMENT NO. REV. NO.
DESCRIPTION OR TITLE: ddd'RMA.
- l. Assigned Responsibilities:
Safety Analysis Preparer:
Lead 1st Safety Reviewer:
2nd Safety Reviewer:
2 ~ Safety Analysis Preparer: Complete PART I, SAFETY ANALYSIS Safety Analysis Preparer Lead 1st Safety Reviewer: Complete Part II, Item'Classification.
Lead 1st Safety Reviewer:
2 Part III is "yes," then Part IV is not required.
may be completed. If either question 1 or Lead 1st Safety Reviever: Determine vhich DISCIPLINES are required for review of this item (including own) and mark the appropriate block(s) belov.
c atu e Date Ste
[] Nuclear Plant Operations
[] Nuclear Engineering g Mechanical
[) Electrical
[) Instrumentation 6 Control
[ ) Structural
[] Metallurgy
[] Chemistry/Radiochemistry
[) Health Physics
[] Administrative Controls A QUALIFIED SAFETY REVIEWER will be assigned for each DISCIPLINE marked in step 5 and his/her name printed in the space provided. Each person listed shall perform a SAFETY REVIEW and provide input into thc Safety Review Package.
- 7. The Lead 1st Safety Reviewer will assure that a Part completed (see step 4 above) and a Part VI if III or Part IV is r'equired (see 9.d of Part II).
Each person listed in step 5 shall sign and date next to his/her name in step 5, indicating completion of a SAFETY REVIEW.
2nd Safety Reviever: PeaEorml a lBAFETY REVIEW in accordance vith Section 8.0.
2nd Safety Reviever Date -NAZ DISCIPLINE:
PNSC reviev required?
belov:
If "yes, attach Part V and mark reason Yes No,
[) Potential UNREVIEWED SAFETY QUESTION" Question 9 of Part IV nswere "Ye Other (specify): 0 (Form AP-011-6-A-1)
et M~ lt" ~<~$<7 Field Ree. tet~
+
REVISION 10CFR50 ~ S9 PROQR7LH Na
~ ~
3 Page ATTACHMENT A CP&L SAFETY REVIEW PACKAGE Page oi PART I: SAFETY ANALYSIS (See instructions in Section 8.4.1)
(Attach additional sheets as necessary.)
DOCUMENT NO. REV. NO.
DESCRIPTION OF CHANGE:
ANALYSIS
'EFERENCES
'Form AP-011-6-A-1)
PLANT MOO. NO.
FfELD REV. NO.
d Mechanical Safety Analysis PAGE NO.
PCR-6547 Page 1 SAFETY ANALYSIS FOR PCR-6547 BACKGROUND This modification replaces the safety injection (SI) alternate miniflow relief valves (1CS-744 and 1CS-755) with orifices open/close logic on the motor operated alternate miniflowand'nhanced isolation valves (1CS-746 and 1-CS-752). The isolation valves will receive an open signal on high RCS pressure to ensure minimum flow for protection of the charging/SZ (CSIP's) via the new orifices. They will also receive pumps a close signal on low RCS pressure to ensure adequate SZ flow is delivered to the core. The purpose of the alternate miniflow sub system is to provide an alternate flow path for protection of the CSIP's for those postulated accidents in which RCS pressure can increase above CSIP shutoff head following SZ actuation.
After this modification is implemented, the safety injection system will be capable of performing its design safety functions. The following specific safety functions could be impacted:
- 1) adequate SI flow must be delivered to the core
- 2) at least 60 gpm must be delivered by each operating CSZP under the most limiting pump condition and operational lineup ANALYSIS
- 1) Adequate SZ Flow Delivered to the Core Calculation HNP-I/INST-1044, Rev. 1 selected a MOV closure setpoint of 1750 psig in the RCS as measured by wide range pressure channels P-402 and P-403. The calculation technique made allowances for:
uncertainty associated with the sensing channel (i.e. P-402/P-403),
uncertainty associated with the pressurizer pressure channels which initiate SI, and an additional margin for conservatism. The MOVs will be expected to close only during those accidents where SI is actuated and RCS pressure is expected to increase above the MOV opening setpoint. Zn these accidents however (inadvertant SI and feedline break) SZ flow provides little benefit. Consequently, MOV closure is not critical to ensure satisfactory performance of the SI system. Automatic closure does, however, increase plant safety by providing an additional backup. (See nuclear Fuels Section Safety Analysis)
=2) At Least 60 gpm Delivered by Each Operating CSZP Under the Most Limiting Conditions To ensure their integrity and long-term availability for accident
Mod No. cd- Mechanical Safety Analysis Fiekt Rev. No.
Page No. PCR-6547 Page 2 mitigation, each operating CSIP must pass at least 60 gpm. First, the new orifices are sized to ensure at least 60 gpm will be passed by each pump in the condition of maximum degradation that will satisfy the .ECCS analyses assumptions.
The CSIP performance curve which defines the minimum allowed performance for ECCS analyses was transmitted to CP&L by Westinghouse in letter no. 92CP*-G-0096. At 60 gpm the TDH is approximately 5623 ft. Pressure downstream of the new orifices at 60 gpm is 84 psig or 194 ft (60 psi dynamic plus 24 psi static).
Calculated orifice performance curve shows an expected flow of 56 gpm at 5429 ft TDH (5623 ft minus 194 ft). Seal injection flow to the reactor coolant pumps will be maintained during safety injection. This flow will be sufficient to ensure total CSIP flow exceeds 60'gpm.
Second, the MOV opening setpoint is low enough to ensure at least 60 gpm is passed by the weakest CSIP in the most limiting ~
configuration. Parallel operation of the weakest and strongest CSIP's is the most limiting configuration due to the increased total flow and head loss. From pre-op test data (1-2080-P-04), at 60 gpm the weakest CSIP total developed head (TDH) is 6140 ft.
Also from 1-2080-P=04, the instrument uncertainty associated with this data is .14 of span (0-3000 psi) of 3 psi. Assuming a reading error of 23 ft yields:
(3 psi) (2.31 ft/psi) + 23 ft = 30 ft Therefore, the minimum weak pump TDH at 60 gpm is:
6140 ft 30 ft = 6110 With the weak pump delivering 60 gpm at 6110 ft, the maximum strong pump flow at the same TDH is 179 gpm. The total flow required to ensure at least 60 gpm is passed by the weakest pump's therefore:
60 gpm + 179 gpm = 239 gpm From pre-op 1-2080-P-04, the SI system resistance through FE-943 ls ~
= 7 10T x 10 ft/gpm (678.9 gpm)~
The head loss through the SI system with the weak pump delivering 60 gpm is:
~h = (7.107 x 10~ ft/gpm~) (239 gpm)~ = 406 ft The head loss from the SI injection points in the cold legs to the pressure transmitter sense line connections in the hot legs is:
Mod No.
Fl RR eNL
.R .~
CHC.
Cold Leg Piping Mechanical Safety Analysis 4 ' ft PCR-6547 Page 4 Reactor Vessel Hot Leg Piping 127 4.3 ft ft 136 ft The head gain between the RWST and the SI injection points in the cold legs is:
RWST elev.
RCS C/L elev.
273
~254 ftt (min level) 19 ft The maximum RCS pressure measured by P-402 and P-403 that will ensure at least 60 gpm is passed by the weakest CSIP during two pump operation is:
P = 6110 ft 406 ft2.31 136 ft + 19 ft ft/psi P = 2419 psig Calculation HNP/INST-1044, Rev 1 selected a MOV opening setpoint of 2300 psig. The instrument channel uncertainty associated with the setpoint is 100 psig , and additional conservatism of 37 psig was applied. The maximum RCS pressure which ensures protection of the weakest pump is above the nominal setpoint plus instrument channel uncertainty. Conservatism in the calculation of maximum pressure ensures pump protection without reliance on the 37 psig conservatism in the setpoint calculation. Very little margin exists, however, for future pump degradation.
Lastly, the combination of MOV opening setpoint and MOV stroke time provide adequate pump protection for the highest expected rate of RCS pressure increase.
Pipe Hanger Anchors CS-H-4400, CS-H-4403, and CS-H-4406 have been redesigned as a result of this modification. Calculations and drawings were revised for each anchor based on loads provided by
'the pipe stress analysis sub-unit. Twenty-nine (29) other pipe hangers remain to be reviewed for load increases prior to declaring this modification.
plant MocL N Page Na Nuclear Fuels Section Safety Analysis PCR-6547 Page 1 This safety analysis focuses an the impact of the apen and close setpoints associated with alternate miniflaw valves 1CS-746 and 1CS-752 on analysis performed in the SHNPP FSAR Chapter 15. The valves will open at a wide range RCS pressure between 2405 and 2195 psig (reference 1) coincident with a Safety Injection actuatian signal. The valves will close at a wide range RCS pressure between 2205 to 1995 psig (reference 1) coincident with a Safety Injection Actuation signal. The open and close pressure ranges averlap; however, this will not cause undue cycling af the valves since the apen and close setpoints are established hy the same instrument channel such that the setpoints move away from the naminal value in the same direction.
The wide range pressure setpoint ta open the valve is established at a law enough RCS pressure to ensure that at least the minimum required flaw is maintained through both operating CSIPs in any Chapter 15 event. This setpoint is above RCS pressures which would occur coincident with a Safety Injection actuatian signal. The Safety Injection (SI) actuatian signals listed in Technical Specification Table 3.3-4 are High Containment Pressure, Low Pressurizer Pressure, and Low Main Steam Pressure, and Manual Safety Injectian.
The Chapter 15 events in which a Safety Injectian actuation occurs are as fallows:
- 1. Inadvertent Opening af a Steam Generator Relief or Safety Valve (FSAR Sectian 15.1.4)
- 2. Steam System Piping Failure (FSAR Section 15.1.5)
- 4. Inadvertent Operation af the Emergency Core Cooling System During Pawer Operatian (FSAR Section 15.5.1.1)
- 5. Inadvertent Opening of a Pressurizer Safety or Power Operated Relief Valve (FSAR Section 15.6.1).
- 6. Steam Generator Tube Rupture .(FSAR Section 15.6.3).
- 7. Loss of Coolant Accidents (FSAR Section 15 6.5)
Events 1 and 2 could actuate SI via Low Steam Line Pressure. In both events, RCS pressure will he substantially below the pressure required to open the alternate miniflow valves at the time of SI actuation. RCS pressure will increase in the event only after
c~-
N N ~+ Safety Analysis pcR-6547 Page 2 critical care parameters have been stabilized. Operators would be expected to secure 8Z prior to reaching RCS pressures vhich exceed that vhich would open the alternate mini.flow -valves. However, opening of the valves at this paint in either event vould have no adverse impact.
Event 3 could actuate SZ via Lov Steam Line Pressure. The safety injectian system has little added benefit in this event. RC8 pressure vill he significantly belav the pressure required to open the alternate miniflow valves at the time of SZ actuation. Zn this event, RCS pressure increases to the pressuri.zer relief setpoint some.200 seconds after SZ actuation. As a consequence, RCS pressure vill increase above the pressure necessary to open the alternate miniflav valves. Therefore, shortly after 8Z actuation occurs, Safety Znjectian only serves to worsen the event hy adding inventory and thereby increasing RCS pressure. As such, opening the alternate mini.flow valves during thi.s portion of the event protects the CSZPs (the CSZPs are needed later for lang term recovery) and reduces the severity of the event (by decreasing the inventary added to the RCS).
Event 4 may cause an increase in RCS pressure above the RCS pressure required to apen the valves coincident. wi.th an SZ signal.
Havever, SZ flav ta the RCS in this event i.s actually detrimental.
As such, opening the alternate minflov valve and thereby reduci.ng
.the amount of SZ flov delivered to the RCB vould lessen the severity of the event.
5, 6, and 7 will
'vents actuate SZ at a pressure vhich i.s significantly less than the RCS pressure necessary to open the alternate miniflow valves. The amount of SZ flov delivered in these events via the CSZPs significantly affects the autcome af these events; however, flow vill not be lost through the alternate miniflav lines since the isolatian valves will remain closed.
The isolatian valve clase setpoint is based on ensuring safety injection flow is established in accordance with analysis of events which actuate SZ on lov'pressuri.zer pxessure. While i.t is true that the manual safety injection may be actuated in events where lav pressuri.zer pressure occurs somewhat later than the actuating signal (such as low steam li.ne pressure during a main steam line break),
pressure aperators which will vill nat actuate SZ manually above an RC8 apen the valves. Operators will manually actuate SZ based on plant symptoms such as lov Volume Control Tank level, lov Reactar Coolant Temperature, low steam pressure, and lov pressurizer level or pressure. 8uch symptoms wi.ll simply not occur at the RCS wi.de range pressures associ.ated with opening the valves
plant W CZ Field Rev. ga Nuclear Fuels Section Safety Analysis, Page Na p
page 3 coincident with an SI actuation (between 2405 and 2195 psig). As such, the RCS wide range low pressure coincident with SI actuation close setpoint serves as an additional backup which exceeds the requirements of the existing licensing bases.
References
- 1. Calculation HNP-I/INST-1044 Rev. 1.
- 2. PSAR Amendment 43a (see sections references above)
i'~nt Mu" h cZ-C <7 F@d Rev. No.
REVISION 3 IOCrRSO. S9 XR00WuC @amer, ATTACHHENT A Page NO.
CP&L SAFETY REVIEW PACKAGE age t
PART II: ITEN CIASSIPICATION DOCUMENT NO. 'REV. NO.
Yes No
- 1. Does this item represent:
- a. A change to the facility as described in the SAFETY g ANALYSIS REPORT?
- b. A change to the procedures as described in the SAFETY ANALYSIS REPORT?
- c. A test or experiment not described in the SAFETY (1 ANALYSIS REPORT?
- 2. Does this item involve a change to the individual plant Operating License or to its Technical Specifications?
- 3. Does this item require a revision to the FSAR?
- 4. Does this item involve a change to the Off-Site Dose Calculation Manual?
- 5. Does this item constitute a change to the Process Control ]]
Program?
- 6. Does this item involve a ma)or change to a Radvaste Treatment []
System?
7 ~ Does this item involve a change to the Technical (]
Specification Equipment List (BSEP and SHNPP only)?
8 ~ Does this item impact the NPDES Permit (all 3 sites) or ,(]
constitute an "unrevt.eved environmental question" (SHNPP Environmental Plan, Section 3.1) or a "significant environmental impact" (BSEP)?
9 ~ Does this item involve a change to a previously accepted:
- a. Quality Assurance Program []
- b. Security Plan (including Training, Qualification, and (]
Contingency Plans)?
- c. Emergency Plan?
- d. Independent Spent Fuel Storage Installation license?
(If "yes," refer to Section 8.4.2, "Question 9," for special considerations. Complete Pert VI in accordance Mith Section 8.4.6)
SEE SECTION 8.4.2 FOR INSTRUCTIONS FOR EACH "YES" ANSWER.
REFERENCES. List FSAR and Technical Specification references used to ansver questions 1-9 above. Identify specific reference sections used for any "Yes" answer.
5A . I 7. /. ] . 2.
I5. 4.
(Form AP-Oil-6-A-1)
N+Cf2~5 ~
REVZBZON 3 10CFR50 e 59 PROQRAH ATTACHMENT A N7LHU7LL
- F e"le< lg>V~
Field Ree age Na Na~
CP&L SAFETY REVIEW PACKAGE PART III: UNREVIEWED SAFETY QUESTION DETERMINATION SCREEN DOCUMENT NO. 'EV . NO.
~es No
- l. Is this change ~u addressed by another completed [)
UNREVIEVED SAFETY QUESTION determination? (See Sections 7.2.1, 7.2.2.5, and 7.9.1.1)
REFERENCE DOCUMENT: REV. NO.
~es No
- 2. Far procedures, is the change a non-intent change which ~o (check all that apply): (See Section 7.2.2.3)
[] Corrects typographical errors which do not alter the meaning or intent of the procedure; or,
[] Adds or revises steps for clarificatian (provided they arc consistent with the original purpose or applicability of the procedure); ar,
[) Changes thc title of an organizational position; or,
[] Changes names, addresses, or telephone numbers of persons; or,
[] Changes the designation of an item of equipment where the equipment is the same as the original equipment or is an authorized replacement; or,
[] Changes a specified tool or instrument to an equivalent substitute; ar,
[] Changes the format .of a procedure without altering the meaning, intent, or content; or
[] Deletes a part or all of a procedure, the deleted portions of which are wholly covered by approved plant procedures?
If the answer to either Question 1 or Question 2 in PART III is "Yes," then PART ".:
need not be complctcd.
(Form AP-011-6-A-1)
P Field Rev. Na REVISION 3 10CFR50 ~ 59 PROQRAH HAKJ7LL ATTACHMENT h Page Na CP&L SAFETY REVIEW PACKAGE age PART IV: UNREVISED SAFETY QUESTION DETERMINATION DOCUMENT NO. 'EV..NO.
Using the SAFETY ANALYSIS developed for the change, test or experiment, as veil as other required references (LICENSING BASIS DOCUMENTATION, Design Dravings, Design Basis Documents, codes, etc.), thc preparer of the Unreviewed Safety Question Determination must directly answer each of the follovtng seven questions and make a determination of vhether an UNREVIBKD SAFETY QUESTION exists.
h WRITTEN BASIS IS REQUIRED POR EACH ANSQER Yes No
'ay the proposed activity increase the probability of l) occurrence of an accident evaluated previously in thc SAFETY ANALYSIS REPORT?
r Ct J C
- 2. May the proposed activity increase the consequences of an accident evaluated previously in thc SAFETY ANALYSIS REPORT?
L7
- 3. May the proposed activity increase the probability of () p; occurrence of a malfunction of equipment important to safety evaluated previously in the SAFETY ANALYSIS REPORT?
rA P7 l ~v WrCE &cK&VG.
- 4. May the proposed activity increase the consequence of a malfunction of equipment important to safety evaluated previously in the SAFETY ANALYSIS REPORT?
rS AD r-'rc47rcd
- 5. May the proposed activity create the possibi.lity of an accident of a different type than any evaluated previously in thc SAFETY ANALYSIS REPORT' J
Aec.r p I J r=do (Form AP-011-6-A-1)
r--n. i;~. t~~cg=C+g Fietd ReV. NO REVISION 3 10CPR50 ~ 59 PROQRAH HAÃUhL pg m ATTACHMENT h CP&L SAFETY REVIEW PACKAGE Page 6 of 8 PART IV: (Continued)
DOCUMENT NO. REV. NO ~
~es ~o
- 6. May the proposed activity create the possibility of a fl malfunction of equipment important to safety of a different type than any evaluated previously in the SAFETY ANALYSIS REPORT?
ee.r.m A
C Cgr ~ MW Sl~drerAC, CAWr g~y Co~VeaC.
Does the proposed activity reduce the margin of safety as mO
() X defined in the basis of any Technical Specification?
I/4Lg/6 c ~ C~ a avH~
de'2 Zo<
0 J Based on the answers to questions 1 - 7, does this item ()
result in an UNREVIEWED SAFETY QUESTION? If any of the questions 1-7 is "Yes," chen the item is the answer to considered co constitute an UNREVIEWED SAFETY QUESTION ~
- 9. Is PNSC reviev required for any of the folloving reasons? (l If, in answering question 1 or 3 "No," it vas determined that the probabili increase was small relative to the uncertainties; or, in ansvering question "
or 4 "No," ic vas determined that the doses increased, but the dose was less than the NRC ACCEPTANCE LIMIT; or, in ansvering question 7 "No," a st'l parameter vould be closer to the NRC ACCEPTANCE LIMIT, but the end result was still vithin the NRC ACCEPTANCE LIMIT; then PNSC reviev is required.
REFERENCES:
raw) ~
This Unrevieved Safety Question Determination is for the folloving DISCIPLINE(s):
(Additional Part IV forms may be included as appropriate.)
Nuclear Plant Operations S true tural Nuclear Engineering Metallurgy Mechanical Chemistry/Radiochemistry Electrical Health Physics Instrumentation & Control Administrative Controls (Form AP-011-6-A-1)
~ Em n". M~ NO '+</g., <S4T REVISION 3 10CPR50 ~ 59 PROQRAH HAHU~
ATTACHMENT A C ~
CP&L SAFETY REVIEW PACKACE PART V: PNSC REVIEW DOCUMENT NO.
p-A
&
Determination/Evaluation:
<J7& '4 Qr 4 5 8-a P<
f p s
/PJJ n ~ 4'-ev y7 gJJ'~ J
'V
]-z/ CJ Wg Action Taken:
e Basis: ~m /n J~E
/NCf/
~ ~
p
~
Vl cM7 vK uc, oP PNSC Chairman:
(Form AP-011-6-A-1)
.. Blod ko P~<-d$ M REVZSZON 3 ioCTR50.59 PROQRMC HARUM'.
ATTACHMENT A II ~D~
CP&L SAFETY REVIEW PACKAGE Page 8 of 8 PART VI: ISFSI CHANCES (10CFR?2 '8)
DOCUMENT NO. REV. NO.
~es No Does this item represent:
- a. A change to the Independent Spent Fuel Storage () t)
Installation (ISFSI) as described in the ISFSI Safety Analysis Report?
A change to the procedures as described in the !) ()
ISFSI Safety Analysis Report2 A test or experiment not described in the ISFSI ll (!
Safety Analysis Report2
- 2. Does this item involve a change to the license conditions incorporated in the ISFSI Operating License?
- 3. Does this item result in a significant increase in [) ()
occupational exposure?
- 4. Does this item result in a significant unrevieved environmental impact?
SEE SECTION 8.4.6 FOR INSTRUCTIONS FOR EACH "YES" ANSWER.
REFERENCES. List ISFSI SAR and Technical Specification references used to ansver questions 1 and 2 above. Identify specific reference sections used for any "Yes" answer.
(Form AP-Oil-6-A-1)
REVZSZON 3 IOCFRS 0 ~ 5 9 PROQRAH KNURL ATTACHMENT h g @57 CP&L SAFETY REVIEW PACKAGE Page I o~ 8 7@nt Viod t
~~ ggy. W P8gg 50 4 SAFETY REVIEW COVER SHEET DOCUMENT NO. Pd.IZ- 547 REV. NO.
DESCRIPTION OR TITLE: ui - WLo&
Assigned Responsibilities:
Safety Analysis Preparer:
Lead 1st Safety Reviewer:
2nd Safety Reviewer:
Safety Analysis Preparer: Comnlete PART I SAFETY ANALYSIS Safety Analysis Preparer
- 3. Lead 1st Safety Reviewer: Complete Part II, Item Classification.
III may If ed'ture Lead 1st Safecy Reviewer: Part be completed. either question 1 or 2 is "yes," then Par't IV is not required.
Lead 1st Safety Reviewer: Determine vhich DISCIPLINES are required for review of this item (including own) and mark the appropriate block(s) belov.
C Rc u ate Ste ?4
[) Nuclear Plant Operations
[) Nuclear Engineering
[] Mechanical Q Electrical I
[) Instrumentation & Control
[) Structural
[] Metallurgy
[) Chemistry/Radiochemistry
[) Health Physics
[) Administrative Controls A QUALIFIED SAFETY REVIEWER vill be assi,gned for each DISCIPLINE marked in step 5 and his/her name printed in the space provided. Each person listed shall perform a SAFETY REVIEW and provide i.nput into the Safety Revie~
Package.
The Lead 1st Safety Reviewer vill completed (see step 4 above) and a Part VI assure that a Part if III or Part IV is required (see 9.d of Part II)..
Each person listed in step 5 shall sign and date next to his/her name i.n step 5, indi.cati.ng completion of a SAFETY REVIEW.
2nd Safety Reviever: Perform a SAFETY REVIEW in accordance vith Section 8.0.
~ n n n f
2nd Sa e ty Reviever Date DISCIPLINE: I
- 9. PNSC reviev required?
belov:
If "yes," attach Part V and mark reason Yes No Potential UNREVIEWED SAFETY QUESTION [) ~
Question 9 of Part IV ansvered "Yes" Other (specify):
(Form AP-011-6-A-1)
'E ~ ~ l
REVZSZON 3 10CFRSO ~ S9 PROQRhH HhHUAL s.
ATTACHMENT A CP&L SAFETY REVIEW PACKAGE Page ~ ~
of p N PART I: SAPETT ANALYSIS (See instructions in Section 8.4.1)
(Attach additional sheets as necessary.)
DOCUMENT NO. REV. NO. CO DESCRIPTION OF CHANGE:
ANALYSIS: 8 SIS
REFERENCES:
5IEM glR )TTShl, 5+5)5 .2 (Form AP-011-6-A-1)
piBntvioc Mg PEA'w~qp REVISION 3 10CPRSO 59 PROGRAM HAÃJAL
~ ~
~
ATTACHMENT A CP&L SAFETY REVIEW PACKAGE Page of PART II: ITEM CLASSIPICATION DOCUME'K NO. REV. NO.
Yes No
- 1. Does this item represent:
- a. A change to the facility as described in the SAFETY (]
ANALYSIS REPORT?
- b. A change to the procedures as described in the SAFETY [)
ANALYSIS REPORT?
- c. A test or experiment not described in the SAFETY ()
ANALYSIS REPORT?
- 2. Does this item involve a change to the individual plant Operating License or to its Technical Specifications?
- 3. Does this item require a revision to the FSAR?
- 4. Does this item involve a change to the Off-Site Dose Calculation Manual?
- 5. Does this item constitute a change to the Process Control ()
Program?
- 6. Does this item involve a ma)or change to a Radvaste Treatment (]
System?
- 7. Does this item involve a change to the Technical '(]
Specification Equipment List (BSEP and SHNPP only)?
- 8. Does this item impact the NPDES Permit (all 3 sites) or t) constitute an "unrevieved environmental question" (SHNPP Environmental Plan, Section 3.1) or a "significant environmental impact" (BSEP)?
- 9. Does this item involve a change to a previously accepted:
- a. Quality Assurance Program l)
- b. Security Plan (including Training, Qualification, and )]
Contingency Plans)?
- c. Emergency Plan?
- d. Independent Spent Fuel Storage Installation license?
(If "yes," refer to Section 8.4.2, "Question 9," for special considerations. Complete Part VI in accordance vith Section 8.4.6)
SEE SECTION 8.4.2 FOR INSTRUCTIONS FOR EACH "YES" ANSWER.
REFERENCES. List FSAR and Technical Specification references used to ansver questions 1-9 above. Identify specific reference sections used for any "Yes"
+l 4 PA (Form AP-011-6-A-1)
pent ICc jq: Peg-Cg~7 '
REVISION 3 10CFR50 ~ 59 PROGRAM Hh?rgb Page NO, C ATTACHMENT A CP&L SAFETY REVIEW PACKAGE PART ZZZ: UNREVIEWED SAFETY QUESTION DETERMINATION SCREEN DOCUMENT NO. REV. NO.
~es h'o
- 1. Is this change ~u addressed by another completed [) 5 UNREVZEWED SAFETY QUESTION determination7 (Sec Sections 7.2.1, 7.2.2.5, and 7.9.1 1) ~
REFERENCE DOCUMENT: REV. NO.
~es ~o 2.
vhich ~
For procedures, is the change a non-intent change (check all that apply): (See Section 7.2.2.3)
Corrects typographical errors vhich do not alter
[I I
[)
the meaning or intent of the procedure; or,
[) Adds or revises steps for clarification (provided they are consistent vith the original purpose or applicability of the procedure); or,
[) Changes the title of an organizational position; or,
[) Changes names, addresses, or telephone numbers of persons; or, f) Changes the designation of an item of equipment vhere the equipment is the same as the original equipmcnt or is an authorized replaccmcnt; or, f] Changes a specified tool or instrument to an equivalent substitute; or, f] Changes the format .of a procedure vithout altering the meaning, intent, or content; or
[] Deletes a part or all of a procedure, thc deleted portions of vhich are vholly covered by approved plant procedurcs7 If the ansver to either Question 1 or Question 2 in PART III is "Yes," then PART '."
need not be completed.
(Form AP-011-6-A-1)
lee e" ) ', 'Cg 45yP ieu~
~
Field Ree.
RZVISZON 3 10CPR50 59 PROQlQLH H)Q?UAL Pdg8 +I
~
~ NCL ATTACHMENT A CP6L SAFETY REVIEW PACKAGE Page of 8 PART IV: UNREVIEWED SAFETY QUESTION DETERMINATION DOCUMENT NO. +C.R-6 7 REV. NO.
Using the SAFETY ANALYSIS developed for the change, test or experiment, as well as other required references (LICENSING BASIS DOCUMENTATION, Design Drawings, Design Basis Documents, codes, etc.), the preparer of the Unreviewed Safety Question Determination must directly answer each of the following seven questions and make a determination of whether an UNREVIEWED SAFETY QUESTION exists.
A WRITTEN BASIS IS REQUIRED POR EACH ANSWER Yes No
- 1. May the proposed activity increase the probability of E) .5 occurrence of an accident evaluated previously in the SAFETY ANALYSIS REPORT?
I PA
- 2. May h
the proposed activity increase the consequences of an accident evaluated previously in the SAFETY ANALYSIS REPORT?
t) I
- 3. May'he proposed activity increase the probability of () 0 occurrence of a malfunction of equipment important to safety evaluated previously in the SAFETY ANALYSIS REPORT?
S~ v4Ki I S.
4, May the proposed activity increase the consequence malfunction of equipment important to safety evaluated of a fl I previously in the SAFETY ANALYSIS REPORT?
S uQRrrrSd &@514
- 5. May the proposed accident of activity create a different type than any the possibility of an evaluated previously
[), I in the SAFETY ANALYSIS REPORT?
vs@, ( 5l&
(Form AP-Oil-6-A-1)
4~~
hh~ N~cg 4 A%7 REVZSZON 3 10CFR50 ~ 59 PROGRAM MAKJ7LL a
Fiekf fbw. Na~
ATTACHMENT A CPSL SAFETY REVIEW PACKAGE Page 6 of 8 PART IV: (Continued)
DOCUMENT NO. R- &5 7 REV. NO. O
~Ye ~o
- 6. May the proposed activity create the possibility of a (l 5 malfunction of equipment important to safety of a different type than any evaluated previously in the SAFETY ANALYSIS REPORT?
- 7. Does the proposed activity reduce the margin of safety as defined in the basis of any Technical Specification?
I AcSt 2
- 8. Based on the answers to questions 1 - 7, does this item result in an UNREVIEWED SAFETY QUESTION?
any of the questions 1-7 is "Yes," then the item is If the answer to considered to constitute an UNREVIEWED SAFETY QUESTION.
- 9. Is PNSC review required for any of the'ollowing reasons? [1 If, in ans~ering question 1 or 3 "No," it was determined that the was small relative to the uncertainties; or, in answering question "
probabili='ncrease or 4 "No," it was determined that the doses increased, but the dose was scil'.
less than the NRC ACCEPTANCE LIMIT; or, in answering question 7 "No," a parameter ~ould be closer to the NRC ACCEPTANCE LIMIT, but the end result was still ~ithin the NRC ACCEPTANCE LIMIT; then PNSC review is required.
REFERENCES:
WR)
This Unreviewed Safety Question Determination is for the following DISCIPLINE(s):
(Additional Part IV forms may be included as appropriate.)
fj Nuclear Plant Operations Structural Nuclear Engineering Hetallurgy Mechanical Chemistry/Radiochemistry Ij Electrical
!
[ Instrumentation 6 Control Health Physics Administrative Controls (Form AP-011-6-A-1)
~ ~ p ~ ~
REVISION 3 10CFR50 ~ 59 PROQRhH HhRU7LL ATTACHMENT A
-. ni rm aw.
P lir m~
"- t ~PcZ c~~y P CP&L SAFETY REVIEW'ACKACE PART V: PNSC REVIEW DOCUMENT NO. REV . NO.
Determination/Evaluation:
Action Taken:
Basis:
PNSC Chairman: Dace:
(Form AP-011-6-A-1)
..ra i, ~Cg -g5g7 REVISION 3 ~ 59 PROQRAH H7NU7LL 8'0CFR50 ATTACHMENT A CP&L SAFETY REVIEW PACKAGE Page 8 of 8 PART VI: ISFSI CHANCES (10CFR72.48)
DOCUMENT NO. REV. NO.
~es Po
- 1. Docs this item represent:
[) I Safety Analysis Report?
- b. A change to the procedures as described ISFSI Safety Analysis Report?
in thc .') I
- c. A test or experiment not described in the ISFSI Safety Analysis Report?,
- 2. Does this item involve a change to the license conditions incorporated in the ISFSI Operating License?
- 3. Does this item result in a significant increase in )) lI occupational exposure?
- 4. Docs this item result in a significant unreviewed [) S environmental impact?
SEE SECTION 8.4.6 FOR INSTRUCTIONS FOR EACH YES" ANSWER.
REFERENCES. List ISFSI SAR and Technical Specification references used to answer questions 1 and 2 above. Identify specific reference sections used for any "Ycs" answer.
(Form AP-011>>6-A-1)
Plant Modification Mod. No. PCR-6547 Safety Review Continuation Page Field Rev. No. 0 Page No. 2.1 PART 1. SAFETY ANALYSIS CONTINUED 1.0 ~
Descri tion of Chan e: continued The electrical portion of this PCR deals with providing cable and raceway information only. There are four new safety related cables and two existing safety cables which have been rerouted that requires additional raceway to be added to complete the circuits. These cables are part the Charging Safety In)ection Pump (CSIP) Alternate Mini.-flow System operating logic for valves 2CS-V757SA-1 and 2CS-V759SB-1 located in the Reactor Auxiliary building.
2.0 ~nal sis Cables 10317F-SA and 10319F-SB will to be rerouted from equipment SSP OUTPUT gl to SSP OUTPUT P2 train "A" and "B" respectively. New cables 10317N-SA, 10317P-SA, 10319N-SB, and 10319P-SB have been added to complete the circuit changes for the operating logic of the valves indicated above. Penetrations will be required to be breached for cables indicated. Conduits 10317N-SA-2.0" and 10319N-SB-2.0" were added to complete cable routes and to avoid overloaded trays. Equipment within the affected area is designed to accepted multiple cables of different voltage classes as well as separation requirements. The cable changes and additions mentioned above will maintain separation as required by FSAR Table 8.3.1-10 and 2166-B-060. MCC P1A21-SA and MCC P 1821-SB located in the RAB is fed from safety related 480V Emergency Bus, 1A3-SA and 183-SB respectively. The increase in electrical load to MCC glA21-SA and MCC P 1B21-SB is a increase to the 480V safety related power distribution systems. This load addition has been evaluated for impact to the DAC>>l and E6001 calculations and was determined to be acceptable.
Plant Modification Mod. No. PCR-6547 Safety Review Continuation Page Field Rev. No. 0 Page No. 2.2 3.0
References:
continued
- 1. FSAR: Index, Chapter 8, 15, Section, 8.0 2 ~ Tech Spec Index
res- t.. fag-C5+7,".
Plant Modification Safety Review Continuation Page C.2 1 Field Rev. No. 0 Page No. 5.1 PART IV: UN1~IRWED SAFETy UESTION DETERMINATION CONTINUED
- 1. The electrical portion of the this PCR adds safety related cables and raceway. These changes are in compliance with FSAR Table 8.3.1-10, 2166-B-060 for separation, FSAR Section 8.3.2.30 for overloaded raceway, Design guides/criterion and related plant procedures. The changes mention in the PCR will not increase the probability of occurrence of an accident.
- 2. The electrical portion of the this PCR adds safety related cables and raceway. These changes are i.n compliance with FSAR Table 8.3.1-10, 2166-B-060 for separationg FSAR Section 8.3.2.30 for overloaded raceway, Design guides and related plant proceduresi Therefore, the changes mention in the PCR will not increase the consequences of an accident.
3 ~ The additional safety related cables and raceway added to the plant tray systems to complete the circuits are in compliance with FSAR Table 8.3.1-10, 2166-B>>060 for separation, FSAR Section 8.3.2.30 for overloaded raceway, Design guides/criterion and related plant procedures. The changes mention in the PCR will not increase the probability of occurrence of a malfunction of equipment important to safety.
The additional safety related cables and raceway added to the plant tray systems to complete the circuits are in compliance with FSAR Table 8.3.1-10, 2166-B-060 for separation, FSAR Section 8.3.2.30 for overloaded raceway, Design guides/criterion and related plant procedures, The changes mention in the PCR will not increase the consequence of a malfunction of equipment important to safety. 'I
- 5. The electrical portion of the this PCR adds safety related cables and raceway. These changes are in compliance with FSAR Table 8.3.1-10, 2166-B-060 for separation, FSAR Section 8.3.2.30 for overloaded raceway, Design guides/criterion and related plant procedures. The changes mention in the PCR will not create the possibility of an accident of a different type than
,any evaluated in the safety analysis report.
0
0
'=.~~au~ Pm>-~~
Fad Res. N~ C Plant Modification Mod. No. PCR-6547 Safety Review Continuation Page Page Na Field Rev. No. 0 Page No. 6.1 PART IV: lBG&VIEWED SAFETY UESTION DETERMINATION CONTINUED
- 6. The electrical portion of the thi.s PCR adds safety related cables and raceway. These changes are in compliance with FSAR Table 8.3.1-10, 2166-B-060 for separation, FSAR Section 8.3.2.30 for overloaded raceway, Design gui.des/criterion and related plant procedures. The changes mention in the PCR will not create the possibility of a malfunction of equipment important to safety as evaluated in the safety analysis report.
7 ~ The additional safety related cables and raceway added to the plant tray systems for the CSIP Alternate Mini.-flow System which is connected to class 1E power source (MCC flA21-SA and MCC f 1B21-SB) wi.ll not reduce the margin of safety as defined in the basis of any Tech. Spec. The additional loads have been analyzed in accordance wi.th applicable design criterion/procedure/guidelines.
~Summar Since the proposed modification does not requi.re a change to the Tech.
Specs. nor involve an unreviewed safety question, in accordance wi.th 10CFR50.59, the proposed changes may be implemented without prior approval of the NRC.
REVZSZOH 3 XOCPR50 59 PROGRAM MMiUAL ATTACHMENT A
, Pc/-g CP6L SAFETY REVIEW PACKAGE Page 1 of 8 "dd i Page'NO. SAFETY REVIEW COVER SHEET DOCUMENT NO. NO. q'EV.
0 DESCRIPTION OR TITLE: C E & p/8'spy
- l. Assigned Responsibilities:
,Safety Analysis Preparer:
Lead 1st Safety Reviewer:
2nd Safety Reviewer:
2; Safety Analysis Preparer: PART I. SAFETY ANALYSIS Preparer'omnletm Safety Analysis
- 3. Lead 1st Safety Reviewer: mplete Part II, Item Classification.
1st Safety Reviewer: Part III may be completed. If ed'ture 4, Lead either question 1 or 2 is "yes," then Part IV is not required.
- 5. Lead 1st Safety Reviewer: Determine which DISCIPLINES are required for review of this item (including own) and mark the appropriate block(s) below.
D SC Re u ate'te 7
[] Nuclear Plant Operations
[] Nuclear Engineering
[] Mechanical
[] Electrical
[] Instrumentation 6 Control c2 Structural
[] Metallurgy
[] Chemistry/Radiochemistry
[] Health Physics
[] Administrative Controls A QUALIFIED SAFETY REVIEWER will be assigned for each DISCIPLINE marked in step 5 and his/her name printed in the space provided. Each person listed shall perform a SAFETY REVIEW and provide input into the Safety Review Package.
7, The Lead 1st Safety Reviewer will assure that a Part completed (see step 4 above) and 'a Part VI if III or Part IV is required (see 9.d of Part II).
Each person listed in step 5 shall sign and date next to his/her name in step 5, indicating completion of a SAFETY REVIEW;
- 8. 2nd Safety Reviewer: Perf~ AAFETY3tCVIEW 1< accordance with Section 8.0.
2nd Safety P~riawar Date DISCIPLINE: /
- 9. PNSC below:
review required? If "yes," attach Part V and mark reason Yes No
] Potential UNREVIEWED SAFETY QUESTION [] 4l Question 9 of Part IV answered "Yes" Other (specify):
(Form AP-011-6-A-1)
- -'icnt M~ Na ~<X
- ~X~
Field Reii. Nc.
REVISION 3 10CPR50 ~ 59 PROGRAM HllMHLL ATTACllMENT A N CP&L SAFETY REVIEW PACKAGE PART I: SAFETY ANAIYSIS (See instructions in Section 8.4.1)
(Attach additional sheets as necessary.)
DOCUNENT NO. I RE.- HC-k2(
I aZV, NO. 0 DESCRIPTION OF CNANGE: QohlbU IT4 IOSI7M-5A-'1 ) IOSIR 4-5h-2 kRE 7b 9j MQ Aun SQPPDaTED lu RAG ARIA CAGLK IOEl7 P P /OSI9 P Ai2E TD EEAODED TO EXISTW6 e.aaLC TPAV5 Vl 'DAB.
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Ocean/~C. Ia r~e AM OP-SAP-En'ZnAaw EaljIPIumr
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~
A55LIQ.ES THE 572VZT'L/d if I IhlYE6R.I TY Qt-" THCSi= MAAPDA/w~TS j9A/D ~BVIPAAEIJT AQ2O ~TiZLId TLIRT UJ HATHI 4 l H SR. IQ F'LLIEkldE REFERENCF I E'C.H.
AP-/J F
8:
'R W6AR. TABLE aW 5PCd, SO/
L Mur~i5 '
IdhEX
'.a.l
'. 5 9'M oO/'M I. /3 I S.D oo2'AP-oo&'I (Form AP-011-6-A-1)
,N JcE-C5 y~~
I'4J ..2 r Old Rcv. No. O REVZSZON 3 10CPRSO 59 PROGRAM MMUAL
~ ~
~
ATTACHMENT A page Na CP&L SAFETY REVIEW PACKAGE age of PART ZI: ITEM CLASSIFICATION DOCUMENT NO. REV. NO.
Yes ~o
- 1. Does this item represent:
- a. A change to the facility as described in the SAFETY ANALYSIS REPORT?
- b. A change to the procedures as described in the SAFETY [)
ANALYSIS REPORT?
- c. A test or experiment not described in the SAFETY [)
ANALYSIS REPORT2
- 2. Does this item involve a change to the individual plant [)
Operating License or to its Technical Specifications?
- 3. Does this item require a revision to the FSAR?
- 4. Does this item involve a change to the Off-Site Dose Calculation Manual2
- 5. Does this item constitute a change to the Process Control Program'?
- 6. Does this item involve a ma)or change to a Radwaste Treatment System?
- 7. Does this item involve a change to the Technical Specification Equipment List $ BSEP and SHNPP only)?
- 8. Does this item impact the NPDES Permit (all 3 sites) or constitute an "unreviewed environmental question" (SHNPP Environmental Plan, Section 3.1) or a "significant environmental impact" (BSEP)?
- 9. Does this item involve a change to a previously accepted:
- a. Quality Assurance Program
- b. Security Plan (including Training, ~liflcation, and .
Contingency Plans)?
- c. Emergency Plan?
- d. Independent Spent Fuel Storage Installation license?
(If "yes," refer to Section 8.4.2, "Question 9," for special considerations. Complete Part VI in accordance with Section 8.4.6)
SEE SECTION 8.4.2 FOR INSTRUCTIONS FOR EACH "YES" ANSWER.
REFERENCES. List FSAR, and Technical Specification references used to answer questions 1-9 above. Ident.'=".y specific reference sections used for any "Yes" answer.
(Form AP-Oll-6-A-l)
,.;.41od NcLcg /A~7 REVZSZON 3 ATTAQQKNT A Page Na CP&L SAFETY REVIEW PACKAGE PART III: UNREVIEWED SAFETY QUESTION DETEMGNATION SCREEN DOCUMENT NO. ZVR- L -(o REV., NO.
Yes ~o
- l. Is t'his change ~fu addressed by another UNREVIEWED SAFETY QUESTION determination?
completed (See I
Sections 7.2.1, 7.2.2.5, and 7.9.1.1)
REFERENCE DOCUMENT: REV. NO.
Yes Po
- 2. For procedures, is the change a non-intent change which ~o (check all that apply): (See Section 7.2.2.3) [] M.
[] Corrects typographical errors vhich do not alter the meaning or intent of the procedure; or, f] Adds or revises steps for clarification (provided they are consistent vith the original purpose or applicability of the procedure); or,
[] Changes the title of an organizational position; or,
[] Changes names, addresses, or telephone numbers of persons; or,
[] Changes the designation of an item of equipment vhere the equipment is the same as the original equipment or is an authorized replacement; or,
[] Changes a specified tool or instrument to an equivalent substitute; or,
[] Changes the format. of a procedure vithout altering the meaning, intent, or content; or
[] Deletes a part or all of a procedure, the deleted portions of vhich are vholly covered by approved plant procedures2 If the answer to either Question 1 or Question 2 in PART III is "Yes," then PART IV need not be completed.
(Form AP-011-6.-'A-1)
.;. Mac N~+~ - Hg, REVISION 3 ATTACHMENT A CP&L SAFETY REVIEW PACKAGE Page ~ of 8 PART ZV: UNREVIEWED SAFETY QUESTION DETERMINATION DOCUMENT NO. J REV. NO.
Using the SAFETY ANALYSIS developed for the change, test or experiment, as well as other required references (LICENSING BASIS DOCUMENTATION, Design Drawings, Design Basis Documents, codes, etc.), the preparer of the Unreviewed Safety Question Determination must directly answer each of the following seven questions and make a determination of whether an UNREVIEWED SAFETY QUESTION exists.
h WRITTEN BASIS IS REQUIRED FOR EACH ANSWER Yes No
- 1. May the proposed activity increase the probability of f) N occurrence of an accident evaluated previously in the SAFETY ANALYSIS REPORT?
THE ANIILYSI5 b'F THE 5I)PPDILTC AND 1 HE &Albl)ll 7PIIETLi&fI Rl QP'HE SY57i3A I5 /NAINTAIA/EQ ARID IZBVICVl5 A55ijkE'r9QE'i P IAN IRWIN ED
- 2. May the proposed activity increase the consequences of an accident evaluated previously in the SAFETY ANALYSIS REPORT?
[) I A5 5 ATCb Jhl 41. ~
ABOVE 'THE 50PPD7LT AhIALV5/S Aap 7ZEUIC~5 ASSI)~C 7///Ir 77IE SA~ S~ie7UQAL INTEhiZITY l5 MAIAITA/hICb - TH/5 AAOD I R dATIQAI E MD IM~ASE LOMSELhJEXI~E
- 3. May the proposed activity increase the probability of ()
occurrence of a malfunction of equipment important to safety evaluated'reviously in the SAFETY ANALYSIS REPORT?
/la Sis/'uS5Ea lu 4f ~+ Z 18DVd'HE STROCrllml lu7FSeITY'ElihAIII5 llNeIIA//G 4A/D THE 5LIPPD85 RAVE bib EWPEP7 DAI EBLIIPMEU7 7 HM.F IDEE 7HSK I5 h/D ulCRCA5E. Ju MOBABiLirY02 A MAL/=LJNZTIOH
. 4. May the proposed activity increase the consequence malfunction of equipment important to safety evaluated of a l) I previously in the SAFETY ANALYSIS REPORT?
TOE QJSCVMIDiu JAI 4 3 ABOVE'DLLD4IA THA7 A/D ZHAkl~E /Al STZlJCMr8 /IJ 4/
Ahlb 5UPPDR75 HAI/IAI6 Al Go&KEd DEW d. E5 DF'uc /=/='ECT Dht EQIJIPiiIEA/7 TIO
- 5. May the proposed activity create the possibili,ty of an accident of a different type than any evaluated previously I) I
. in the SAFETY ANALYSIS REPORT?
7//E FuhliiTIO~ b> TIIE SOPPDPTA IIAVE BEEN P/ieODO5I Y EVALuATE7) Aub hD 4 o7 d.RESTE AmmCur O/=' Z>>FFr~7 7Y E //W AIVV i>REVIDODLY EvALu~r-(Form AP-Oil-6-A-I)
,. Mw i M~~~'.
REVISION 3 10CFR50 ~ 59 PRtMMM MMUAL .
ATTAQiMENT A p CP&L SAFETY REVIEW PACKAGE Page 6 of 8 DOCUMENT NO. TVQ- HE- 6)'2 PART t'EV.
IV: (Continued)
NO.
~es Po
- 6. May the proposed activity create the possibility of a malfunction of equipment important to safety of a different f) I type than any. evaluated previously in the SAFETY ANALYSIS REPORT?
15 $7ÃlPD kl 44 77/L FVn'LT)O)J Ok 7PE SuN'd)2G HAS 8)W N&/OV5LY LVACI)A)LT)
Adb /7 ht A LP'll d.TI 0
- 1005 TL~ D
- 7. Does the proposed activity reduce the margin of safety as defined in the basis of any Technical Specification? l) *1 7 AW CDPCDRT AIIIALY5K An/D REU)EUU SATISFY ALL L)Q~Z)n/I& Z)p~nyGVl Zt EQ I RE MELlTS J.
']
- 8. Based on the answers to questions 1 - 7, does this. item []
result in an UNREVIEWED SAFETY QUESTION?
any of the questions 1-7 is "Yes," then the item is If the answer to M
considered to constitute an UNREVIEWED SAFETY QUESTION.
- 9. Is PNSC review required for any of the following reasons? jg If, in answering question 1 increase was small relative or 3 "No," it was determined that the probability to the uncertainties; or, in answering question 2 or 4 "No," it was determined that the doses increased, but the dose was still
'less than the NRC ACCEPTANCE LIMIT; or, in answering question 7 "No," a l
'I - ~
parameter would be closer to the NRC ACCEPTANCE LIMIT, but the end result was still within the NRC ACCEPTANCE LIMIT; then PNSC revie~ is required.
REFERENCES:
This Unreviewed Safety Question De>ermination is for the following DISCIPLINE(s):
(Additional Part IV forms may be c.eluded as appropriate.)
Nuclear Plant Operations Nuclear Engineering I Structural
[) Metallurgy Mechanical [) Chemistry/Radiochemistry Electrical [) Health Physics Instrumentation & Control () Administrative Controls (Form AP-011-6-A-1)
Rev. NQ REVZSZOM 3 I.OCPR50 59- PROQRhH HAMUAL Field ATTACHMENT A Page Na CP&L SAFETY REVIEW PACKAGE PART V: PNSC REVZEV DOCUMENT No. Z RQ. M C fo 2. Ln aEV. NO. 0 Determination/Evaluation:
Action Taken:
Basis'NSC Chairman: Date:
(Form AP-Oll-6-A-1)
sMnt Mod. N PC WaRe.No ~
REVISION 3 lOCPRSO ~ 59 PROI7kM KQiUAL ATTAQiMENT A page Na CP&L SAFETY REVIEV PACKAGE PART VI: ISFSI CHANCES (10CFR72.48)
DOCUMENT NO. ERR. AE 6 ZG REV. NO. O Yes No Does this i represent
- a. A cha e the In pendent pent Fuel Storage t) [)
Insta atio (ISFS as des ibad in the ISFSI Safe Analy s Re rt?
- b. A ch e toth r edur as described in the t) t)
ISFSI afety Ana s s Re ort? .
A tes or experime n t descri d n the ISFSI [)
Safety Analysis Rep r '?
- 2. Does this item involve a hange to the licens () ()
conditions incorporated n the IS SI Operating cense?
- 3. Does this item result in a si ficant increase in l) l) occupational exposur
- 4. Does this item result in a significant unrevieved f) l) environmental impact?
SEE SECTION 8.4.6 FOR INSTRUCTIONS FOR EACH "YES" ANSWER.
REFERENCES. List ISFSI SAR and Technical Specification references used to answer questions 1 and 2 above. Identify specific reference sections used for any "Yes" answer.
.(Form AP-011-6-A-1)
REVZSZON 3 3.On'RSO.59 MOD NO.
PaaaRMC Munx7Lri PAGE NO.
ATTACHMENT h OO O~
CP&L SAFETY REVIEW PACKAGE Page 1 of 8 SAFETY REVIEW COVER SHEET DOCUMENT NO. REV. NO.
DESCRIPTION OR TITLE:
Assigned Responsibilities:
Safety Analysis Preparer:
Lead 1st Safety Reviever:
2nd Safety Raviever:
- 2. Safecy Analysis Preparer: Complete PART I, SAFETY ANALYSIS r
Safety Analysis Preparer
- 3. Lead 1st Safety Reviever: Complete Part II, Item Classification.
- 4. Lead 1st Safety Raviever: Part III may be completed. If either question 1 or 2 is "yes,'hen Part IV is noc required.
- 5. Lead 1st Safety Ravievet". Determine vhich DISCIPLINES are required for reviev of this item (including ovn) and mark tha appropriate block(s) belov.
a e a e te 7'l
[] Nuclear Plant Operations
[) Nuclear Engineering
'[] Mechanical f] Electrical I Structural Instrumentation & Control
[]
[) Metallurgy
[) Chemistry/Radiochemis try
[] Health Physics
[) Administrative Concrols.
- 6. h QUALIFIED SAFETY REVIEWER vill be assigned for each DISCIPLINE marked in scop 5 and his/her name printed in the space provided. Each person listed shall perform a SAFETY REVIEW and provide input into the Safecy Reviev Package.
- 7. The Lead 1st Safety Reviever vill assure that a Part completed (see step 4 above) and a Part VI if III or Part IV is required (see 9.d of NPart II).
Each person listed in scop 5 shall sign and date next to his/her name in seep 5, indicating completion of a SAFETY REVIEW.
- 8. 2nd Safety Reviever: Perform a.ShFETY REVIEW in accordance vith Section 8.0.
2nd Safety Reviavar Date Z
'ISCIPLINE:
- 9. PNSC reviev required?
belov:
If "yes," attach Part V and mark reason ~as No Potential UNREVIEWED SAFETY QUESTION
[] I Question 9 of Part IV ansvered "Yes" Other (specify):
(Form AP-011-6-A-1)
XOCPR50 ~ 59 PROORhH lQDiUAL 0
RMNDN
~
Na.~
REVZSZOM 3 ATTACHMENT h CP&L SAFETY REVIEW PACKAGE ppQf Page ~ ~of PART I; SAFETY ANALYSIS (See instructions in Section 8.4.1)
(Attach additional sheets as necessary.)
DOCUMENT NO. REV. NO.
DESCRIPTION OF CHANGE:
ANALYSIS:
REFERENCES:
(Form AP-011-6-A-1)
NIL~
MOD IKL RRISIONlla~
REVZSZON 3
~ ~
ASK ATTACHMENT h CP&L SAFETY REVIEW PACKAGE Page of PART ZZ: ZTEH CIASSIFZCATZON DOCUMENT NO. REV. NO.
- 1. Docs this item represent:
- a. h change to the facility as described in the SAFETY g ANALYSIS REPORT?
- b. h change to the procedures aa described in the SAFETY []
ANALYSIS REPORT?
- c. h test or experiment not described i.n the SAFETY []
ANALYSIS REPORT?
- 2. Docs this item involve a change to the individual plant Operating License or to its Technical Specifications2
- 3. Does this item require a revision to the FSAR2
- 4. Does thii item involve a change to the Off-Site Dose Calculation Manual?
- 5. Does this item constitute a change to the Process Control [)
Program?
- 6. Does this item involve a ma)or change to a Radwaste Treatment []
System?
- 7. Does this item involve a change to the Technical []
Specification Equipment List (BSEP and SHNPP only)?
- 8. Docs this item impact the NPDES Permit (all 3 sites) or constitute an "unreviewed environmental question" (SHNPP Environmental Plan, Section 3.1) or a "significant environmental impact (BSEP) 2
- 9. Does this item involve a change to a previously accepted:
- a. Quality Assurance Program
- b. Security Plan (including Training, ~lification, and Contingency Plans)?
- c. Emergency Plan?
- d. Independent Spent Fuel Storage Installation license?
(If "yes," refer to Section 8.4.2, "Question 9," for special considerations. Complete Part VI in accordance with Section 8.4.6)
SEE SECTION 8.4.2 FOR INSTRUCTIONS FOR EACH "YES" ANSWER.
REFERENCES. List FSAR and Technical Specification references used to answer questions 1-9 above. Identify specific reference sections used for any "Yes" answer.
(Form AP-011-6-A-1)
REVISION 3 10CFR50 ~ S9 PROORhH M~~ AGE ~
MOO NO.
KID etaall IKI.~
ATTACHMENT h CP6L SAFETY REVIEM PACKAGE Page 4 of 8 PART III: UNREVISED SAFETY QUESTION DETERMINATION SCREEN DOCUMENT NO. REV. NO ~
- 1. Is this change ~
UNREVIEWED SAFETY QUESTION addressed by anather completed determination? (See Sections 7.2.1, 7.2.2.5, and 7.9.1.1)
~es
[]
Po I
REFERENCE DOCUMENT: REV. NO.
Yes No 2.
vhich ~
For procedures, is the change a non-intent change (check all that apply): (See Section 7.2.2.3)
Corrects typographical errors which do not alter
[) I
[]
the meaning or intent of the procedure; or,
[] Adds or revises steps for clarification (provided they are consistent vith the original purpose or applicability of the prcicedure); oz,
[) Changes the ticle of an organizational position; or,
[] Changes names, addresses, or telephane numbers of persons; or,
[) Changes the designation of an item of equipment vhere the equipment is the same as the ariginal equipment or is an authorized replacemenc; or,
[] Changes a specified tool or instrument to an equivalent substitute; or,
[] Changes the format. of a procedure vithaut altering the meaning, intent, or content; or
[] Deletes a part or all of a procedure, the deleted portions of
~hich are @holly covered by approved plant procedures2 If the answer to either Question 1 or Question 2 4n PART III is "Yes,'hen PART '."
need nat be completed, (Form AP-011-6-A-1)
REVZBZON 3 ATTACHMENT h CPRL SAFETY REVIEV PACKhGE MOD MO, LOCI R50 ~ 59 PROQKLM HANU~ AGE No "KNnall N Page ~ of
.~
8 PART IV: UNREVISED SAFETY QUESTION DETERMINATION DOCUMENT NO. REV. NO.
Using the SAFETY ANALYSIS developed for the change, test or experiment, as well as other required references (LICENSING BASIS DOCUMENTATION, Design Drawings, Design Basis Documents, codes, etc.), the preparer of the Unreviewad Safety Questt.on Determination must directly answer each of the fallowing seven questions and make a determination of whether an UNREVIEWED SAFETY QUESTION exists, h WRITTEN BASIS ZS REQUIRED FOR EACH ANSWER
~es Na
- 1. May the proposed activity increase the probability of () N accurrence of an accident evaluated previously in the SAFEIY ANALYSIS REPORT7
- 2. May the proposed activity increase the consequences of an accident evaluated previously in -the SAFETY ANALYSIS REPORT?
- 3. May the proposed activity increase the probability of () 0 occurrence of a malfunction of equipment important to safety evaluated previausly in the SAFETY ANALYSIS REPORT7
- 4. May the proposed activity increase the consequence malfunction of equipment important to safety evaluated of a [) r previously in the SAFETY ANALYSIS REPORT?
- 5. May the proposed activity create the possibility of an accident of a different type than any evaluated previously in the SAFETY ANALYSIS REPORT7 (Form AP-011-6-A-1)
MOO HO.
REVZBZON 3 10CFR50e59 PROORAH FIELD REAStOH MRS~ AGE gg,
~
ATTACHMENT A CP&L SAFETY REVIEV PACKAGE Page 6 of 8 PART IV: (Continued)
DOCUMENT NO. REV. NO.
~o
- 6. May the proposed activicy create the possibility of a ()
malfunction of equipment important to safety of a different type than any evaluated previously in the SAFETY ANALYSIS REPORT2
- 7. Docs thc proposed activity reduce the margin of safety as defined in the basis of, any Technical Specification2
- 8. Based on the ansvars to questions 1 - 7, does this item result in an UNREVIBKD SAFETY QUESTION2 If any of the'questions 1-7 is "Yas," then the item is the ansver to considered to constitute an UNREVIEWED SAFETY QUESTION.
- 9. Is PNSC rcvicv required for any of the folloving reasons2 (J If, in ansvering question 1 or 3 "No," it vas determined that thc probabilit:
increase vas small relative to the uncertainties; or, in ansvering question '"
or 4 "No," it vas determined that the doses increased, but tha dose vas still less than the NRC ACCEPTANCE LIMIT; or, in answering question 7 "No," a parameter would be closer to the NRC ACCEPTANCE LIMIT, but the end result was still within the NRC ACCEPTANCE LIMIT; then PNSC revicv is required.
REFERENCES:
This Unreviewed Safety Question Determination is for the following DISCIPLINE(s):
(Additional Part IV forms may be included as appropriate.)
() Nuclear Plant Operations Structural Nuclear Engineering Me tallurgy Mechanical Chemistry/Radiochemistry
() Electrical Health Physics g Instrumentation & Control Administrative Controls.
(Form AP-011-6-A-1) ~ g 4 '
REVZSZON 3 DOCUMENT NO.
hTThCHMENT h CP&L ShFETY REVIEV PhCKhCE PhRT V: PNSC REVIEV MOD No.
lie Page
~
REUNION REV. NO.
7 of ~
Determination/Evaluation:
hction Taken:
Basis'NSC Chairman: Date:
(Form AP-Oil-6-A-1)
MOO NO.
F)ELD RKV%NN W.
REVISION 3 IOCFR50 ~ 59 PROQRAH HlLRJ)LL PPQE ATTACHMENT A
~
CP&L SAFETY REVIEW PACKhGE Page 8 of 8 PART VI: ISFSI CHANCES (10CFR?2.48)
DOCUMENT NO. REV. NO.
~es ~o Does this item represenr,:
() I Safety Analysis Report?
- b. A change to the procedures as described in the ISFSI Safety Analysis Report?
CI A test or experiment not described in the ISFSI Safety Analysis Report?
- 2. Does this item involve a change to the license conditions incorporated in the ISFSI Operating License?
- 3. Does this item result in a significant increase in occupational exposure?
- 4. Does this item result in a significant unreviewed [) W environmental impact?
SEE SECTION 8.4.6 FOR INSTRUCTIONS FOR EACH YES" ANQKR.
REFERENCES. List ISFSI SAR and Technical Specification references used to answer questions 1 and 2 above. Identify specific reference sections used for any "Yes" answer.
(Form AP-011-6-A-1)
Installation Package Mod. No. PCR-6547 Safety Review Continuation Page Field Rev. No. 0 Page No.t" qg ART I AFETY A ALY I n inu 1.0 DE RIPTI N F HAN E'continued)
This modification introduces changes to the operating logic of CSIP alternate mini-flow isolation valves 1CS-746 (2CS-V757SA-1) and 1CS-752 (2CS-V759SB-1). The logic of these motor operated valves will be modified such that the valves will open upon high RCS pressure coincident with an "S" signal. IfRCS pressure approaches the shutoff head of the pumps, the isolation valves will open and provide sufficient flow to prevent pump damage. Additionally, these isolation valves will close as the RCS depressurizes and in response to a safety injection signal to provide maximum injection flow; This will be accomplished by adding bistable circuitry to RCS wide range pressure loops P-402 and P-403. Comparator cards (NAL) and solid state relay circuitry (NAS) will be added to protection cabinets 1 & 4. These bistables will energize/de-energize rotary relays (K711-A & K711-B) within the SSPS output bays.
Contacts of these SSPS relays will be installed in series with contacts of safety injection relay K740 to provide automatic valve control.
The subject isolation valves presently receive an "S" signal from relay K636 located in SSPS ouput bay 1. This relay is manually reset early in the transient as directed by the emergency operating procedures. Ifthis signal is reset prior to RCS pressure increasing to 2300 psig, the mini-flow valves may never open. Due to this concern, this modification will substitute a RWST - SI signal, which is not reset until the normal charging header is aligned, for the present SI signal. This design change involves removing the K636 relay from valve circuitry and utilizing the K740 relay (located in SSPS output bay 2) for the safety injection permissive.
2.0 A~NALY I ' I'l The intent of this modification is to replace the existing passive pressure control system with an active pressure control system consisting of solid state instrumentation for the purpose of eliminating problematic relief valves possessing high failure potential. The parallel dead head protection valves will be controlled by independent protection instrument loops. RCS wide range pressure loop P-402 via
Installation Package Mod. No. PCR-6547 Safety Review Continuation Page Field Rev. No. 0 Page No. t" 90 Y I in protection cabinet 1 and in series with SSPS train "A" output logic will provide automatic pressure control for valve 2CS-V757SA-1. Redundantly, RCS wide range pressure loop P-403 will automatically control valve 2CS-V759SB-1 through protection cabinet 4 circuitry in series with SSPS "B" train output logic. Each independent train is physically separated to preserve redundancy and to ensure that no single credible event will create common failure. All of the new materials introduced to the protection cabinets and SSPS output bays are 1E qualified and are identical to those originally supplied by Westinghouse to provide protection features.
Changes to the isolation valve operating logic will result in an automatic opening.
permissive which will occur upon high RCS pressure (M2300 psig) coincident with an "S" signal. Automatic valve closure will be initiated by a low RCS pressure (C1750 psig) event and an "S" signal. The low pressure permissive is a function of bistable deadband to preclude the possibility of introducing unbiased errors into the opening and closing logical features. The open pressure permissive setpoint was calculated using a value for RCS pressure (Hot Leg) that insured a low enough pump discharge pressure to allow a minimum required flow of 60 gpm. The setpoint is low enough to protect the CSIP's by assuring isolation valve opening prior to RCS pressure reaching the shutoff head of the pump. The low RCS pressure permissive (bistable reset) is low enough so that the alternate mini-flow MOV open permissive will not be in effect when the PORV's are open, but also high enough to insure adequate injection from the CSIP's. The setting was calculated by taking the P-11 permissive setpoint (insuring the PORV's will be closed) and subtracting instrument uncertainties associated with P-11 and wide range RCS pressure bistables.
Modifications to the Solid State Protection System involve the removal of the K636 relay from valve logic. This change involves the sparing of contacts from the subject relay and has no detrimental effect upon the ability of this device to perform its intended protective function. The newly added K711 relays will be energized/de-energized by the aforementioned pressure bistables. K711 relay contacts will be wired in series with contacts of safety injection slave relay K740. This alignment does not introduce any adverse scenario that would prohibit the slave relays from performing other intended protective functions.
Installation Package Mod. No. PCR4547 Safety Review Continuation Page Field Rev. No. 0 Page No. C-+ t ART I'FETY NALV I n in With the introduction of automatic valve logic, manual valve control aspects are limited to the extent that manual over-ride via MCB control switches will not be possible. Hence, once the MOV's receive a shut signal, operator action cannot re-open the valve. The ability of operations'ersonnel to control alternate mini-flow is unchanged by this modification since the present relief valve pressure control system dictates recirc. to the RWST.
In conclusion, the design intent of the CSIP alternate mini-flow system is unchanged by this modification. The implement for accomplishing the intended system function(s) has been altered to remedy operability concerns as delineated by NRC 0445. Although an active pressure control system. introduces different failure" mechanisms than those associated with a passive system, the ability of the system to tolerate a single active failure and perform its intended safety function is not compromised. This is accomplished by a combination of suitable redundancy, protection instrumentation, and proper bistable actuation to preclude pump damage and to ensure injection flow.
3.0 l~lZFEl!RN: ( th d)
FSAR SECTION 6.3.2.1; 7.1.2; 7.2.2; 15.1.5; 15.2.8; 15.6.5 FSAR TABLE 6.3.2-3; 7.3.1-5 DBD-104 TECH. SPEC. 2.1.2; 3/4.1.2; 3/4.3.2; 3/4.4.9; 3/4.5
Installation Package Mod. No. PCR-6547 Safety Review Continuation Page Field Rev. No. 0 Page No.
ART IV'VIEWED AFETV N DETERMINATI N n inue BA I F R 1 (continued)
- 1. As identified in FSAR section 15.5.1.1, spurious Emergency Core Cooling System (ECCS) operation at power could be caused by operator error or a false electrical signal from the safety injection system actuation channels. The subject design change modifies the Solid State Protection System such that contacts of an existing safety injection related slave relay will be wired in series with a newly introduced RCS pressure'elay. This circuit configuration does not introduce any potential unanticipated adverse reactions that could affect the safety injection capabilities of the slave relay. The potential of inadvertent operator induced ECCS actuation is not credibly linked to any design changes introduced by this modification.
BA I F R TI N 2 (continued)
- 2. Events which result in a safety injection actuation are as follows:
I. Inadvertent Opening of a Steam Generator Relief or Safety Valve;
- 2. Steam System Piping Failure;
- 3. Feedwater System Pipe Break;
- 4. Inadvertent Operation Of The ECCS System;
- 5. Inadvertent Opening of a PORV;
- 6. Steam Generator Tube Rupture;
- 7. Loss of Coolant Accidents.
Only events pertaining to items 3 and 4 are expected to cause the alternate mini-flow valves to open. Chapter 15 transient curves show RCS pressure approaching bistable setpoints only during these two events. The purpose of alternate mini-flow is to provide protection for the CSIP's for these postulated accidents in which RCS pressure can increase above the shutoff head follow'ing SI actuation. Based upon bistable setpoint methodology, pump protection is provided during events 3 & 4 and mitigation is not compromised. During the other five events, maximum injection flow is provided for accident mitigation.
Installation Package Mod. No. PCR<547 Safety Review Continuation Page Field Rev. No. 0 Page No.
PART IV' DETERMINATI N ntin ed BA I F R TI N (continued)
- 3. The ECCS system is designed and analyzed to tolerate a single active failure. Table 6.3.1.1 provides a failures modes and effectes analysis which demonstrates the capability of the ECCS to perform following a single active failure. This analysis shows that single failures, such as the loss of a CSIP, do not compromise the ability to prevent or mitigate accidents. This modification introduces suitable redundancy, protective instrumentation, and materials possessing less failure potential of those contained in the existing pressure control system and, therefore, does not augment the failure effects as analyzed.
BA I F R TI N 4 (continued)
- 4. Analyzed effects analyses show that the consequences of single active failure will not jeopardize ECCS capabilities to perform required protective functions. This modification does not reduce system redundancy and does not downgrade support system performance necessary for reliable operation. The consequence of motor operated isolation valve actuation circuitry failure presents no greater consequence than that associated with existing relief valve failure.
BA I F R TI N (continued)
- 5. The equipment/system operating parameters introduced by this design change do not alter the design intent of the CSIP alternate mini-flow system. Although the method of accomplishing adequate pressure control is changed, this improved means of pressure control does not introduce any transients not bounded by FSAR assumptions. In essence, the possibility of loss of high head safety injection is decreased by this design change.
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Safety Review Continuation Page Field Rev. No. 0 Page No.
PART IV EVIEWED FETV N DETER INATI n in A I F R TI N (continued)
- 6. Although the active pressure control system introduced by this modification presents different failure mechanisms than those associated with a passive system, this modification does not introduce previously unanticipated failure mechanisms at a system level which, as presently analyzed in FSAR chapter 6.3.1, assumes the worst system wide single failure - the loss of a CSIP.
BA I F R TI N 7 (continued)
- 7. The changes introduced by this modification do not affect any safety limit and/or limiting safety system setting as governed by the technical specifications. This modification does not decrease or otherwise contradict the conservatism established in the basis for any ESF, ECCS, remote shutdown, or accident monitoring related technical specification.
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SK-6547-2-001 2166-B-401 317 YES SK-6547-2-002 2166-B-401 319 YES SK-6547-Z-003 1364-46574 NO SK-6547-Z-004 1364-46577 NO SK-6547-2-005 2166-S-PRC0402 NO SK-6547-Z-006 2166<<S-PRC0403 NO SK-6547-Z-007 1364-1328 S29 NO SK-6547-2-008 1364-1328 S29 NO SK-6547-2-009 1364>>10929 S02 NO SK-6547-Z-010 1364-10929 S05 NO SK-6547-2-011 1364-51840 NO SK-6547-2-012 1364-51840 NO SK-6547-Z-013 1364-51840 NO SK-6547-Z-014 1364-92103 NO SK-6547-Z-015 1364-92103 NO A SK-6547-Z-016, 1364-92103 ~
NO SK-6547-2-017 1364-92103 NO SK-6547-2-018 1364-51837 NO SK-6547-Z-019 1364-51837 NO SK-6547-2-020 1364-51837 NO SK-6547-2-021 1364-51837 NO SK-6547-2-022 1364-92103 NO SK-6547-Z-023 1364-92103 NO SK-6547-2-024 1364-92103 NO SK-6547-Z-025 1364-92103 NO SK-6547-Z-026 1364-2776 S26 NO SK-6547-2-027 1364-2776 828 NO SK-6547-Z-028 1364-45841 S59 NO SK-6547-2-029 1364-45841 S58 NO
Installation Package Mod. No. PCR-6547 Drawing List Field Rev. No. 0 Page No. D3 Drawin No. Ref. Dw . No. ~OVS EV.
SK-6547-Z-030 1364-37747 NO SK-6547-Z-031 1364-37747 NO SK-6547-2-032 1364-37747 NO SK-6547>>Z-033 1364-37747 NO SK-6547-Z-034 1364-2776 S30 NO SK-6547-Z-035 1364-45841 S49 NO SK-6547-Z-036 1364-37746 NO SK-6547-Z-037 1364-37746 NO SK-6547-Z-038 2166-8-2020 S28 YES
INSTALLATION PACKAGE MOD. NO. PCR-6547.
DRAWING LIST FIELD REV. NO. 0 PAGE NO. D4 DRAWING NO. REVISION NO.
SK-6547-M-2000 B OVSD SK-6547-M-2001 B OVSD SK-6547-M-2002 B OVSD SK-6547-M-2003 A OVSD SK-6547-M-2004 A SK-6547-C-1001 PG.1 A SK-6547-C-1001 PG.2 A SK-6547-C-1002 PG.1 A SK-6547-C-1002 PG.2 A SK-6547-C-1003 PG.1 A SK-6547-C-1003 PG.2 A SK-6547-C-1004 B SK-6547-C-1005 B
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SK-6547-C-1000 N/A NO CONDUIT SUPPORTS FOR CONDUITS SHEET 1 OF 2 10317N AND 10319N REACTOR AUXILIARY BUILDING EL.305'-0" SK-6547-C-1000 N/A NO CONDUIT SUPPORTS FOR CONDUITS SHEET 2 OF 2 10317N AND 10319N REACTOR AUXILIARY BUILDING EL@ 305 0 SK-6547-E-3300 2166-G-322 NO REACTOR AUXILIARY BUILDING TRAYS'ROUNDING EL.305 ~
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OVALITY LEVEL:
CAROLINA POWER & LIGHT COMPANY NUCLEAR ENGINEERING DEPARTMENT PLANT. HARRIS NUCLEAR PROJECT - UNIT I SCALE: NONE LE: PCS PRINTED CIRCUIT CARD LIST CABINET 81 REVISION FOR UNIT 1 PCR-6547 ~ ~
REVI DATE DESCRIPTION DSNEHK DV DPE DPPE tg NO 1 ~64 51 848 IREV. 5 SHT: OF NO SK 6547 Z 0 I?
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NSS P IN ASSIGNMENTS CABINETS 1 -8, 1 7, 1 8, AND1 9 REYlSION FOR UNIT 1 PCR-6547 EV ( PATE PESCRIPTION DSN HK DVIDPE DPPE DENT I 364-921 83 REY 5 ( SHT.
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PCR QUALITY LEVEL:
6547 CAROLINA POWER & LIGHT COMPANY NUCLEAR ENGINEERING DEPARTMENT PLANT: HARRIS NUCLEAR PROJECT UNIT I SQQfONE TITLE:
NSS PIN ASSIGNMENTS CABINETS 1-8, 17, 18, AND19 REVISION FOR UNIT A PCR-6547 I ~
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6547 CAROLINA POWER 4 LIGHT COMPANY NUCLEAR ENGINEERING DEPARTMENT ~I%
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PCR QUALITY LEVEL:
6547 CAROLINA POWER & LIGHT COMPANY NUCLEAR ENGINEERING DEPARTMENT PLANT..HARRIS NUCLEAR PROJECT - UNIT I SCALE: NONE TITLE:
NSS PIN ASSIGNMENTS CABINETS 1-8, 17, 18, AND19 REVISiON FOR UNIT 1 PCR-6547 REVI DATE DESCRIPTION WDDSN)HK DV DP DPPE DP1D1DDPD 1364-92193 REV. 5 SHT: OF No ~SK 6547 Z 0I7
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QUALIIY lEVEL:
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CAROLINA POWER 8c LIGHT COMPANY NUCLEAR ENGINEERING DEPARTMENT ~i~
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PCR 6547 EE SH.II.) IK FOR CIRC.UlT i'HEKT6 PROFESSIONAL ENGINEER'UALITY LEVEL:
CAROLINA POWER & LIGHT COMPANY NUCLEAR ENGINEERING DEPARTMENT CPS L PIJLNT: HARRIS NUCLEAR PROJECT - UNIT I SCALE'NONE PCS-PRINTED CIRCUIT CARD LIST CABINET 84 REVIS1ON FOR UNIT 1 PCR-6547 RDI DATE DESCRIPTION DSN HK DV DPE DPPE 1364-51837 DwC. NO. REV 4 SHT: OF MOD SKETCH NP SK-6547- Z-020
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QUALITY LEVEL:
CAROLINA POWER & LIGHT COMPANY NUCLEAR ENGINEERING DEPARTMENT PLANT: HARRIS NUCLEAR PROJECT - UNIT I NONE mLE'CS-PRINTED CIRCUIT CARD LIST CABINET 84 REVISION FOR UNIT 1 A +gp PCR-6547 EV DATE DESCRIPTION WNDSN HK DV DPE DPPE DwC".'NO. 1364-5 8a7 REV. 4 SHT: OF SKETCH NO SK 6547 Z 02 I 2
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A QUALITY LEVEL:
PCR 6547 CAROLINA . POWER h LIGHT COMPANY NUCLEAR ENGINEERING DEPARTMENT PLANT: HARRIS NUCLEAR PROJECT - UNIT HQHE NSS PIN ASSIGNMENTS CABINETS 1-8, 17, 18, AND19 REVISION FOR UNIT 1 9 PCR-6547 I DATE DESCRIPTION IJIINDSN Nl< DV DPE DPPE 1364 92183 5 OIYG NP REV SHT: OF MOD S~ETCH wo. SK-6547-Z-022
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A QUALITY LEVEL:
PCR 6547 CAROLINA POWER &: LIGHT COMPANY NUCLEAR ENGINEERING DEPARTMENT PLANT: HARRIS NUCLEAR PROJECT - UNIT I NONE TITLE: NSS PlN ASSIGNMENTS CABINETS 1-8, 17, 18, AND19 REVISlON FOR UNIT 1 PCR-6547 REVi DATE DESCRIPTION PNlssNPslsv(Pvsl>>ss ts~p'.p i364-92193 spy 5 sHs, ps SKETCH NO SK-6547-Z.- 029
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PCR 6547 3l Chk PROFESSIONAL ENGINEER:
OLIALITY LEVEL:
CAROLINA POWER R I IGHT COMPANY NUCLEAR ENGINEERING DEPARTMENT PLANT: HARRIS NUCLEAR PROJECT - UNIT I SCALE; HONK NSS PIN ASSIGNMENTS CABINETS 1-8, 17, 18, AND19 REVISION FOR UNIT 1 PCR-6547 r EV) DATE DESCRIPTION I1QDEP)HK DV DPE DPPE 1354-92183 5 DP DPD PD PEP EE, SKETCH NO SK-6547-Z-OP "l
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A QUALITY LEVEL:
PC 6547 CAROLINA POWER & LIGHT COMPANY NUCLEAR ENGINEERING DEPARTMENT PLANT: HARRIS NUCLEAR PROJECT - UNIT I SCALE: NONE TITLE:
SSPS-INTERCONNECTION DIAGRAM REVISION FOR UNIT 1 A PCR-6547 DATE OESCRIPTION YNOSN HK OV OPE OPPE tg NP 1364 2776 REV 18 SHT.S28OF MOO'KETCH NO SK-6547-Z-027
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CAROLINA POWER k LIGHT COMPANY NUCLEAR ENGINEERING DEPARTMENT CP&L P~T: HhRRIS NUCLEAR PROJECT UNIT I S~E'ONE TITLE:
SSP S- SCHEMATIC DIAGRAM UN1T 1 REVISION FOR PCR-6547 DATE DESCRIPTION WNDSN HK DV DPE DPPE 1 364- 45841 REV. 7 SHT:S59OF SKETCH NO.
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OI7ALI1Y LEVEL CAROLINA POWER k LIGHT COMPANY NUCLEAR ENGINEERING DEPARTMENT '~i~
PLANT: HARRIS NUCLEAR PROJECT - UNIT 1 SCALE: NONE SSPS OUTPUT CABINET 2 REVISION FOR WIRE LIST PCR-6547 UNIT 1 REV DATE DESCRIPTION DSN HK DV DPE DPPE DWG. NO.
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QUALITY LEVEL:
CAROUNA POWER & LIGHT COMPANY NUCLEAR ENGINEERING DEPARTMENT CO&L PLAN7: HARRIS NUCLEAR PROJECT UNIT I SCALE: NONE
, DATE REVISION FOR PCR-6547 DESCRIPTION DSN HK Dv DPE DPPE Dwc'NO SSPS OUTPUT CABINET 1364-37747 WIRE LIST UNIT 1 RE 2 2
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CAROLINA POWER & LIGHT COMPANY NUCLEAR ENGINEERING DEPARTMENT PLANT: HARRIS NUCLEAR PROJECT - UNIT 1 SCALE: NONE SSPS SCHEMATIC REVISION FOR DIAGRAM A 9~ PCR-6547 UNIT 1 PUWT REV DATE DESCRIPTION W DSN HK OV DPE OPPE r DgG NP 1364-45841 REV 7 SHT.S4 OF MOD NO SK 6547 Z 09& 12
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QUAUTY LEVEL:
CAROLINA POWER & LIGHT COMPANY NUCLEAR ENGINEERING DEPARTMENT CO&L PLANT: HARRIS NUCLEAR PROJECT - UHLT 1 SCALE: NOHS TITl.E:
SSPS OUTPUT CABINET 1 REVLSLON fOR WIRE LIST PCR-6547: UNIT 1 REV DATE DESCRIPTION W DSN HK DV DPE DPPE DWC. NO. 1364-37746 REv 2 SHT: OF SK-6547- Z.-0&6
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152B SAFETY RELATED CAROLIHA POWER Bc LIGHT COMPANY NUCLEAR EHGINEERIHG DEPARTMENT DWG'S FOR REVISION: PLANT: HARRIS NUCLEAR PROJECT UNIT 1 SCALE: NONE 5-G-884,5- S-1384 TITLE:
5-G-885.5-S-1385 CHEMICAL AND VOLUME FSAR FIG. 9.3.4-82 FSAR FIG. 9.3.4-83 CONTROL SYSTEM FLOW DIA. REVISION PLANT PwG NP SEE LEFT REV. SHT: 1 PF 1 CAE FILE1 2000 SK-6547-M-2000 17
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