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| issue date = 07/16/1985
| issue date = 07/16/1985
| title = Forwards Addl Info Re Reactor Vessel Level Indicating Sys (Rvlis),Per Sser License Condition.Info Documents Comparison of Emergency Operating Procedures (EOP) That Incorporate Westinghouse RVLIS Sys W/Generic EOP Guidelines
| title = Forwards Addl Info Re Reactor Vessel Level Indicating Sys (Rvlis),Per Sser License Condition.Info Documents Comparison of Emergency Operating Procedures (EOP) That Incorporate Westinghouse RVLIS Sys W/Generic EOP Guidelines
| author name = ZIMMERMAN S R
| author name = Zimmerman S
| author affiliation = CAROLINA POWER & LIGHT CO.
| author affiliation = CAROLINA POWER & LIGHT CO.
| addressee name = DENTON H R
| addressee name = Denton H
| addressee affiliation = NRC OFFICE OF NUCLEAR REACTOR REGULATION (NRR)
| addressee affiliation = NRC OFFICE OF NUCLEAR REACTOR REGULATION (NRR)
| docket = 05000400
| docket = 05000400
Line 16: Line 16:


=Text=
=Text=
{{#Wiki_filter:REGUL ATO~INFORMA T ION DISTRIBUTION TEH (R IDS)>~.AC~ESSION NBR: 8507260041 DOC~DATE: 85/07/16 NOTARIZED:
{{#Wiki_filter:REGUL ATO~INFORMAT ION       DISTRIBUTION               TEH (R IDS)
NO FACIL:50-400 Shearon Har r is Nuclear Power Plantg Unit 1~Carolina AUTH BYNAME AUTHOR AFFILIATION ZIMMERMANrS.R, Car ol ina Power L Light Co~RECIP~NAME.RECIPIENT AFFILIATION DENTON~H.RE Office of Nuclear Reactor Regulationg Director DOCKET-'05000000
        >~
.AC~ESSION NBR: 8507260041           DOC ~ DATE: 85/07/16 NOTARIZED: NO                                                          DOCKET-'
FACIL:50-400 BYNAME    Shearon Har   r is Nuclear Power Plantg Unit 1~ Carolina                                                       05000000 AUTH                   AUTHOR   AFFILIATION ZIMMERMANrS.R,       Car ol ina Power L Light Co ~
RECIP ~ NAME.         RECIPIENT AFFILIATION DENTON~H.RE           Office of Nuclear Reactor Regulationg Director


==SUBJECT:==
==SUBJECT:==
Forwards addi info re reactor vessel level indicating.
Forwards addi info re reactor vessel level indicating.
sys'RVLIS)<per SSER license condition+Info documents-comparison of emergency operating procedures (EOP)that incorporate>>
SSER license condition+Info documents- comparison                         sys'RVLIS)<per of emergency operating procedures (EOP) that incorporate>>
Westinghouse RVLIS sys w/generic EOP guidelines.,Lt DISTRIBUTION CODE: B001D COPIES RECEIVED:LTR gENCL L SIZE:='ITLE: Licensing Submittal:
Westinghouse RVLIS sys w/generic EOP guidelines.
PSAR/FSAR Amdts 8, Related Correspondence'OTES:
DISTRIBUTION CODE: B001D COPIES RECEIVED:LTR gENCL L SIZE:=
RECIPIENT ID CODE/NAME, NRR/OL/AOL NRR LB3.LA INTERNAL: ACRS 41 ELD/HDS1 IE/DEPER/EPB 36 NRR ROErM~L NRR/DE/CEB 11 NRR/DE/EQB 13 NRR/DE/HEB 18 NRR/OE/SAB 24 NRR/DHFS/HFEBPO NR R/OHF S/P SR B NRR/OS I/AEB 26 NRR/DSI/CPB 10 NRR/OS I/ICS8 16 NRR/OS I/PSB 19 NRRIOS I/RSB 23 RGiV2 EXTERNAL: 24X DMB/DSS (AMDTS)NRC PDR 02 PNL GRUELgR COPIES LTTR ENCL 1 0 1 0, 6 1 0 1 1 1 1 1 2 2 1 1 1 1 1 1 1 ,1 1 1 1 1 f 1 1 1 1 3 3 1 1 1 1 1 1 1 RECIPIENT ID CODE/NAME NRR LB3 BC BUCKLEY'S 01 ADM/LFHB IE;, FILE IE/DQA VT/QAB21 NRR/DE/AEAB NRR/DE/EHEB NRR/DE/GB 28 NRR/DE/HTEB 17 NRR/DE/SGEB 25 NRR/DHFS/LQB 32 NR R I D L'/S 8 P 8 NRR/DS I/ASB NRR/DS I'/CSB 09 iVRR/DS I/HETB 12 RAB 22 REG FILE 04 RIB BNL(AMDTS ONLY)LPDR.03-NSIC'5 COPIES LTTR ENCL 1 0 1 1 1 0 1 1=1 1 1 0.1 1 2 2 1 1 1 1 1 1 1 0 1 1 1 1 1 1 1 1 1 0.1 1 1 1 1 1 TOTAL NU>>>>lBER OF COPIES REQUIRED LTTR 52 ENCL f I  
                                                                                                                                  ,Lt Licensing Submittal: PSAR/FSAR Amdts 8, Related                   Correspondence'OTES:
'0 e I e 1 ('(l (r t I le'I%/,'I gael 6 1t3 f(t I, e y'h 1('"''lf ttl I g t c lt e Sv e'I'l V.mw.e eee-I''(I!V r 1 e u I k>l;f  
                                                                                                                        'ITLE:
~0 Carolina Power 8 Light Company JUL 1 6 1985 SERIAL: NLS-85-200 Mr.Harold R.Denton, Director Office of Nuclear Reactor Regulation United States Nuclear Regulatory Commission Washington, DC 20555 SHEARON HARRIS NUCLEAR POWER PLANT UNIT NO.I-DOCKET NO.50-000 REACTOR VESSEL LEVEL INDICATING SYSTEM
RECIPIENT             COPIES            RECIPIENT                                  COPIES ID CODE/NAME,           LTTR ENCL        ID CODE/NAME                                LTTR ENCL NRR/OL/AOL                 1    0    NRR   LB3 BC                                  1                    0 NRR  LB3. LA             1      0,    BUCKLEY'S            01                        1                      1 INTERNAL: ACRS                 41             6    ADM/LFHB                                        1                    0 ELD/HDS1                   1    0    IE;, FILE                                      1                    1=
IE/DEPER/EPB 36                   1    IE/DQAVT/QAB21                                  1                    1 NRR ROErM ~ L             1      1    NRR/DE/AEAB                                    1                    0.
NRR/DE/CEB        11      1      1    NRR/DE/EHEB                                    1                    1 NRR/DE/EQB        13      2      2    NRR/DE/GB            28                        2                   2 NRR/DE/HEB        18      1      1    NRR/DE/HTEB          17                        1                    1 NRR/OE/SAB        24              1    NRR/DE/SGEB          25                        1                    1 NRR/DHFS/HFEBPO            1      1    NRR/DHFS/LQB         32                       1                    1 NR R/OHF S/P SR B          1      1          I NR R D L'/S 8 P 8                               1                    0 NRR/OS I/AEB 26          ,1      1    NRR/DS   I/ASB                                 1                    1 NRR/DSI/CPB      10      1      1    NRR/DS I'/CSB 09 NRR/OS I/ICS8    16      1      f    iVRR/DS I /HETB 12                             1                     1 NRR/OS I/PSB      19      1     1                 RAB      22                        1                     1 NRRIOS I/RSB 23            1     1     REG FILE              04                        1                     1 RGiV2                      3      3                  RIB                              1                   0.
EXTERNAL: 24X                          1     1     BNL(AMDTS ONLY)                                1                     1 DMB/DSS (AMDTS)
NRC PDR PNL GRUELgR 02 1
1 1
1 1     NSIC'5 LPDR.                 03-                      1 1
1 1
TOTAL NU>>>>lBER   OF COPIES REQUIRED         LTTR   52     ENCL       f I
 
                            '0                                           e 1 ('(
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                                                                      'lf               t g
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~ 0 Carolina Power 8 Light Company SERIAL: NLS-85-200 JUL 1 6 1985 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation United States Nuclear Regulatory Commission Washington, DC     20555 SHEARON HARRIS NUCLEAR POWER PLANT UNIT NO. I - DOCKET NO. 50-000 REACTOR VESSEL LEVEL INDICATINGSYSTEM
 
==Dear Mr. Denton:==
 
Carolina Power R Light Company (CPRL) hereby submits additional information concerning the Shearon Harris Nuclear Power Plant (SHNPP) Reactor Vessel Level Indicating System (RVLIS). This information is submitted in response to the Safety Evaluation Report (SER) license condition concerning Inadequate Core Cooling Instrumentation. As discussed on page 0-0 of SER Supplement l, the NRC requires that prior to criticality CPRL demonstrate that the SHNPP Emergency Operating Procedures (EOP) that incorporate the generic Westinghouse RVLIS system conform to generic EOP guidelines relating to the use of the RVLIS, or that deviations be identified and explained. The attached information documents CPRL's completion of this evaluation.
If you have any questions concerning this subject or require additional information, please contact me.
Yours very truly, S    . Zi merman nager Nuclear Licensing Section 3HE/ccc (1670GAS)
Attachment Cct    Mr. B. C. Buckley (NRC)                                  Mr. Wells Eddleman Mr. G. F. Maxwell (NRC-SHNPP)                            Mr. 3ohn D. Runkle Dr. 3. Nelson Grace (NRC-RII)                          Mr. G. A. Schwenk Mr. Travis Payne (KUDZU)                                Dr. Richard D. Wilson Mr. Daniel F. Read (CHANGE/ELP)                          Mr. G. O. Bright (ASLB)
Wake County Public Library                              Dr. 3. H. Carpenter (ASLB)
Mr. 3. L. Kelley (ASLB)
        '11 b.
    . 85072b004i 8507105000 PDR  'ADOCN E
Fayettevilte Street  ~ P. O. Box 1551 o Raleigh, N. C. 27602
 
V I'
 
ATTACHMENT The Westinghouse Owners'roup Emergency Response Guidelines (ERG) Revision I, high pressure version incorporates a generic Reactor Vessel Level Indicating System (RVLIS) which is described in the executive volume of the ERGs under the generic issues section entitled, "Reactor Vessel Liquid Inventory System." A comparison of the generic RVLIS against the Reactor Vessel Level Instrument System Manual that describes the SHNPP RVLIS system shows that the SHNPP RVLIS is consistent with the generic RVLIS in terms of instrument readout capabilities (i.e., upper range/full range/dynamic head range) and application of RVLIS reading to different plant situations.
Enclosures  l and 2 contain the necessary cross-reference information to show that the plant  specific use of RVLIS in SHNPP EOPs conform to the generic ERGs relating to the use of RVLIS. Enclosure 2 lists the comparison between generic and plant specific usage of RVLIS and any deviations are identified and explained.
(1674GAS/ccc )


==Dear Mr.Denton:==
Carolina Power R Light Company (CPRL)hereby submits additional information concerning the Shearon Harris Nuclear Power Plant (SHNPP)Reactor Vessel Level Indicating System (RVLIS).This information is submitted in response to the Safety Evaluation Report (SER)license condition concerning Inadequate Core Cooling Instrumentation.
As discussed on page 0-0 of SER Supplement l, the NRC requires that prior to criticality CPRL demonstrate that the SHNPP Emergency Operating Procedures (EOP)that incorporate the generic Westinghouse RVLIS system conform to generic EOP guidelines relating to the use of the RVLIS, or that deviations be identified and explained.
The attached information documents CPRL's completion of this evaluation.
If you have any questions concerning this subject or require additional information, please contact me.Yours very truly, 3HE/ccc (1670GAS)Attachment S.Zi merman nager Nuclear Licensing Section Cct Mr.B.C.Buckley (NRC)Mr.G.F.Maxwell (NRC-SHNPP)
Dr.3.Nelson Grace (NRC-RII)Mr.Travis Payne (KUDZU)Mr.Daniel F.Read (CHANGE/ELP)
Wake County Public Library Mr.Wells Eddleman Mr.3ohn D.Runkle Mr.G.A.Schwenk Dr.Richard D.Wilson Mr.G.O.Bright (ASLB)Dr.3.H.Carpenter (ASLB)Mr.3.L.Kelley (ASLB).85072b004i 85071 b.PDR'ADOCN 05000 E'11 Fayettevilte Street~P.O.Box 1551 o Raleigh, N.C.27602 V I' ATTACHMENT The Westinghouse Owners'roup Emergency Response Guidelines (ERG)Revision I, high pressure version incorporates a generic Reactor Vessel Level Indicating System (RVLIS)which is described in the executive volume of the ERGs under the generic issues section entitled,"Reactor Vessel Liquid Inventory System." A comparison of the generic RVLIS against the Reactor Vessel Level Instrument System Manual that describes the SHNPP RVLIS system shows that the SHNPP RVLIS is consistent with the generic RVLIS in terms of instrument readout capabilities (i.e., upper range/full range/dynamic head range)and application of RVLIS reading to different plant situations.
Enclosures l and 2 contain the necessary cross-reference information to show that the plant specific use of RVLIS in SHNPP EOPs conform to the generic ERGs relating to the use of RVLIS.Enclosure 2 lists the comparison between generic and plant specific usage of RVLIS and any deviations are identified and explained.
(1674GAS/ccc
)
Q.
Q.
ENCLOSURE I: EOP TITLE CROSS REFERENCE CPRL WOG Guidelines Title EPP-I EPP-2 EPP-3 EPP-0 EPP-5 EPP-6 EPP-7 EPP-3 EPP-9 EPP-10 EPP-I I EPP-12 EPP-13 EPP-IO EPP-15 EPP-16 EPP-17 EPP-13 EPP-19 EPP-20 EPP-21 EPP-22 ECA-O.I ECA-O.I ECA-0.2 ES-O.I ES-0.2 ES-0.3 ES-0.0 ES-I.I ES-1.2 ES-1.3 ES-1.0 ECA-I.I ECA-1.2 E-2 ECA-2.1 N/A ES-3.1 ES-3.2 ES-3.3 ECA-3.1 ECA-3.2 ECA-3.3 Loss of AC Power to IA-SA and IB-SB Busses Loss of All AC Power Recovery Without SI Required Loss of All AC Power Recovery With SI Required Reactor Trip Response Natural Circulation Cooldown Natural Circulation Cooldown With Steam Void in Vessel (with RVLIS)Natural Circulation Cooldown With Steam Void in Vessel (Without RVLIS)SI Termination Post-LOCA Cooldown and Depressurization Transfer to Cold Leg Recirculation Transfer to Hot Leg Recirculation Loss of Emergency Coolant Recirculation LOCA Outside Containment Faulted Stea'm Generator Isolation Uncontrolled Depressurization of All Steam Generators SGTR Isolation Post-SGTR Cooldown Using Backfill Post-SGTR Cooldown Using Blowdown Post-SGTR Cooldown Using Steam Dump SGTR With Loss of Reactor Coolant: Subcooled Recovery SGTR With Loss of Reactor Coolant: Saturated Recovery SGTR Without Pressurizer Pressure Control FRP-S.I FRP-S.2 FRP-C.I FRP-C.2 FRP-C.3 FRP-H.I FRP-H.2 FRP-H.3 FRP-H.O FRP-H.5 FRP-P.I FRP-P.2 FRP-3.1 FRP-J.2 FRP-3.3 FRP-I.I FRP-I.2 FRP-I.3 FR-S.I FR-S.2 FR-C.I FR-C.2 FR-C.3 FR-H.I FR-H.2 FR-H.3 FR-H.O FR-H.5 FR-P.I FR-P.2 FR-Z.I FR-Z.2 FR-Z.3 FR-I.I FR-I.2 FR-I.3 PATH-1 E-0 2 E-I PATH-2 E-3 Response to Nuclear Power Generation/ATWS Response to Loss of Core Shutdown Response to Inadequate Core Cooling Response to Degraded Core Cooling Response to Saturated Core Cooling Response to Loss of Secondary Heat Sink Response to Steam Generator Overpressure-Response to Steam Generator High Level Response to Loss of Normal Steam Release Capability Response to Steam Generator Low Level Response to Imminent Pressurized Thermal Shock Conditions Response to Anticipated Pressurized Thermal Shock Conditions Response to High Containment Pressure Response to Containment Flooding Response to High Containment Radiation Level Response to High Pressurizer Level Response to Low Pressurizer Level Response to Voids in Reactor Vessel Reactor Trip or Safety Injection/Loss of Reactor Secondary Coolant Steam Generator Tube Rupture NOTE: This cross-reference list may change depending on future WOG ERG revision and plant specific needs.(1674GAS/ccc
ENCLOSURE       I: EOP TITLE CROSS REFERENCE WOG CPRL       Guidelines                                   Title EPP-I       ECA-O. I         Loss of AC Power to IA-SA and IB-SB Busses EPP-2      ECA-O. I        Loss of All AC Power Recovery Without SI Required EPP-3      ECA-0.2          Loss of All AC Power Recovery With SI Required EPP-0      ES-O. I          Reactor Trip Response EPP-5      ES-0.2          Natural Circulation Cooldown EPP-6      ES-0.3          Natural Circulation Cooldown With Steam Void in Vessel (with RVLIS)
)  
EPP-7      ES-0.0          Natural Circulation Cooldown With Steam Void in Vessel (Without RVLIS)
EPP-3      ES-I. I          SI Termination EPP-9      ES-1.2          Post-LOCA Cooldown and Depressurization EPP-10      ES-1.3          Transfer to Cold Leg Recirculation EPP-I I    ES-1.0          Transfer to Hot Leg Recirculation EPP-12      ECA-I. I        Loss of Emergency Coolant Recirculation EPP-13      ECA-1.2          LOCA Outside Containment EPP-IO      E-2              Faulted Stea'm Generator Isolation EPP-15      ECA-2.1          Uncontrolled Depressurization of All Steam Generators EPP-16      N/A              SGTR Isolation EPP-17      ES-3.1          Post-SGTR Cooldown Using Backfill EPP-13      ES-3.2          Post-SGTR Cooldown Using Blowdown EPP-19      ES-3.3          Post-SGTR Cooldown Using Steam Dump EPP-20      ECA-3.1          SGTR With Loss of Reactor Coolant: Subcooled Recovery EPP-21      ECA-3.2          SGTR With Loss of Reactor Coolant: Saturated Recovery EPP-22      ECA-3.3          SGTR Without Pressurizer Pressure Control FRP-S. I   FR-S. I          Response  to Nuclear Power Generation/ATWS FRP-S.2     FR-S.2          Response  to Loss of Core Shutdown FRP-C. I   FR-C. I          Response  to Inadequate Core Cooling FRP-C.2     FR-C.2          Response  to Degraded Core Cooling FRP-C.3     FR-C.3          Response  to Saturated Core Cooling FRP-H. I   FR-H. I          Response  to Loss of Secondary Heat Sink FRP-H.2     FR-H.2          Response  to Steam Generator Overpressure FRP-H.3     FR-H.3          -Response  to Steam Generator High Level FRP-H.O     FR-H.O          Response  to Loss of Normal Steam Release Capability FRP-H.5     FR-H.5          Response  to Steam Generator Low Level FRP-P.I     FR-P. I          Response  to Imminent Pressurized Thermal Shock Conditions FRP-P.2     FR-P.2          Response  to Anticipated Pressurized Thermal Shock Conditions FRP-3.1     FR-Z. I          Response  to High Containment Pressure FRP-J.2     FR-Z.2          Response  to Containment Flooding FRP-3.3     FR-Z.3          Response  to High Containment Radiation Level FRP-I. I   FR-I. I          Response  to High Pressurizer Level FRP-I.2     FR-I.2          Response  to Low Pressurizer Level FRP-I.3     FR-I.3          Response  to Voids in Reactor Vessel PATH-1      E-0  2 E-I      Reactor Trip or Safety Injection/Loss of Reactor Secondary Coolant PATH-2      E-3              Steam Generator Tube Rupture NOTE:    This cross-reference list may change depending on future WOG ERG revision and plant specific needs.
(1674GAS/ccc )
 
ENCLOSURE 2: RVLIS CROSS-REFERENCE SHNPP                            WOG EOPs                            ERGs                                Differences
: l. EPP-6/la                        ES-0.3/la                          None
: 2. EPP-6/15                        ES-0.3/5                            None
: 3. EPP-6/2 1c                      ES-0.3/11c                          None EPP-12/13a                      ECA-l. 1/13a                        None
: 5. EPP-12/18a                      E CA-l. 1/18a                      None
: 6. EPP-21/15                        ECA-3.2/16                          See Note  1
: 7. EPP-21/19b                      ECA-3.2/20a                        See Note  1
: 8. EPP-22/8c                        ECA-3.3/7c                          None
: 9. EPP-22/12                        ECA-3.3/11                          None
: 10. Foldout I/a 2)                  ECA-3.2 Foldout/1                  See Note 1
: 11. Foldout 3/a 2)                  ECA-3.3 Foldout/1                  See Note 2
: 12. FRP-C. 1/6a                      FR-C. I/6a                          None
: 13. FRP-C.1/16c                      FR-C. 1/16c                        None
: 10. FRP-C.1/23                      FR-C.1/23                          None
: 15. FR P-C.2/5a                      FR-C.2/5a                          None
: 16. FRP-C.2/7a                      FR-C.2-7a                          None
: 17. FRP-C.2/18                      FR-C.2/18                          None
: 18. FR P-P. I/5                     FR-P.1/5                            None
: 19. FRP-P. 1/12                      FR-P.1/12                          None
: 20. FRP-I.3/8b                      FR-I.3/8b                          None
: 21. FRP-I.3/10                      FR-I.3/10                          None
: 22. FRP-I.3/18                      FR-I.3/18                          None
: 23. FRP-I.3/20                      FR-I.3/20                          None
: 20. PATH-2/J-9                      E-313S                              None
: 25. CSF-2                           F-0.2                               None
: 26. CSF-6                            F-0.6                              None NOTE 1:    The generic guidelines only allow the operator to run one RCP as part of ERG recovery actions. The plant specific EOPs allow the operator to run more than one RCP and therefore the plant specific steps account for more than one RCP running.
NOTE 2:   The ERG step only gives guidance for no RCPs running. The SHNPP EOP step contains additional guidance to account for possibility that RCPs are running.
(1674GAS/ccc)


ENCLOSURE 2: RVLIS CROSS-REFERENCE l.2.3.5.6.7.8.9.10.11.12.13.10.15.16.17.18.19.20.21.22.23.20.25.26.SHNPP EOPs EPP-6/la EPP-6/15 EPP-6/2 1c EPP-12/13a EPP-12/18a EPP-21/15 EPP-21/19b EPP-22/8c EPP-22/12 Foldout I/a 2)Foldout 3/a 2)FRP-C.1/6a FRP-C.1/16c FRP-C.1/23 FR P-C.2/5a FRP-C.2/7a FRP-C.2/18 FR P-P.I/5 FRP-P.1/12 FRP-I.3/8b FRP-I.3/10 FRP-I.3/18 FRP-I.3/20 PATH-2/J-9 CSF-2 CSF-6 WOG ERGs ES-0.3/la ES-0.3/5 ES-0.3/11c ECA-l.1/13a E CA-l.1/18a ECA-3.2/16 ECA-3.2/20a ECA-3.3/7c ECA-3.3/11 ECA-3.2 Foldout/1 ECA-3.3 Foldout/1 FR-C.I/6a FR-C.1/16c FR-C.1/23 FR-C.2/5a FR-C.2-7a FR-C.2/18 FR-P.1/5 FR-P.1/12 FR-I.3/8b FR-I.3/10 FR-I.3/18 FR-I.3/20 E-313S F-0.2 F-0.6 Differences None None None None None See Note 1 See Note 1 None None See Note 1 See Note 2 None None None None None None None None None None None None None None None NOTE 1: NOTE 2: The generic guidelines only allow the operator to run one RCP as part of ERG recovery actions.The plant specific EOPs allow the operator to run more than one RCP and therefore the plant specific steps account for more than one RCP running.The ERG step only gives guidance for no RCPs running.The SHNPP EOP step contains additional guidance to account for possibility that RCPs are running.(1674GAS/ccc)
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Latest revision as of 04:11, 22 October 2019

Forwards Addl Info Re Reactor Vessel Level Indicating Sys (Rvlis),Per Sser License Condition.Info Documents Comparison of Emergency Operating Procedures (EOP) That Incorporate Westinghouse RVLIS Sys W/Generic EOP Guidelines
ML18019A286
Person / Time
Site: Harris Duke Energy icon.png
Issue date: 07/16/1985
From: Zimmerman S
CAROLINA POWER & LIGHT CO.
To: Harold Denton
Office of Nuclear Reactor Regulation
References
NLS-85-240, NUDOCS 8507260041
Download: ML18019A286 (10)
v • d • e


Text

REGUL ATO~INFORMAT ION DISTRIBUTION TEH (R IDS)

>~

.AC~ESSION NBR: 8507260041 DOC ~ DATE: 85/07/16 NOTARIZED: NO DOCKET-'

FACIL:50-400 BYNAME Shearon Har r is Nuclear Power Plantg Unit 1~ Carolina 05000000 AUTH AUTHOR AFFILIATION ZIMMERMANrS.R, Car ol ina Power L Light Co ~

RECIP ~ NAME. RECIPIENT AFFILIATION DENTON~H.RE Office of Nuclear Reactor Regulationg Director

SUBJECT:

Forwards addi info re reactor vessel level indicating.

SSER license condition+Info documents- comparison sys'RVLIS)<per of emergency operating procedures (EOP) that incorporate>>

Westinghouse RVLIS sys w/generic EOP guidelines.

DISTRIBUTION CODE: B001D COPIES RECEIVED:LTR gENCL L SIZE:=

,Lt Licensing Submittal: PSAR/FSAR Amdts 8, Related Correspondence'OTES:

'ITLE:

RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME, LTTR ENCL ID CODE/NAME LTTR ENCL NRR/OL/AOL 1 0 NRR LB3 BC 1 0 NRR LB3. LA 1 0, BUCKLEY'S 01 1 1 INTERNAL: ACRS 41 6 ADM/LFHB 1 0 ELD/HDS1 1 0 IE;, FILE 1 1=

IE/DEPER/EPB 36 1 IE/DQAVT/QAB21 1 1 NRR ROErM ~ L 1 1 NRR/DE/AEAB 1 0.

NRR/DE/CEB 11 1 1 NRR/DE/EHEB 1 1 NRR/DE/EQB 13 2 2 NRR/DE/GB 28 2 2 NRR/DE/HEB 18 1 1 NRR/DE/HTEB 17 1 1 NRR/OE/SAB 24 1 NRR/DE/SGEB 25 1 1 NRR/DHFS/HFEBPO 1 1 NRR/DHFS/LQB 32 1 1 NR R/OHF S/P SR B 1 1 I NR R D L'/S 8 P 8 1 0 NRR/OS I/AEB 26 ,1 1 NRR/DS I/ASB 1 1 NRR/DSI/CPB 10 1 1 NRR/DS I'/CSB 09 NRR/OS I/ICS8 16 1 f iVRR/DS I /HETB 12 1 1 NRR/OS I/PSB 19 1 1 RAB 22 1 1 NRRIOS I/RSB 23 1 1 REG FILE 04 1 1 RGiV2 3 3 RIB 1 0.

EXTERNAL: 24X 1 1 BNL(AMDTS ONLY) 1 1 DMB/DSS (AMDTS)

NRC PDR PNL GRUELgR 02 1

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1 1 NSIC'5 LPDR. 03- 1 1

1 1

TOTAL NU>>>>lBER OF COPIES REQUIRED LTTR 52 ENCL f I

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~ 0 Carolina Power 8 Light Company SERIAL: NLS-85-200 JUL 1 6 1985 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation United States Nuclear Regulatory Commission Washington, DC 20555 SHEARON HARRIS NUCLEAR POWER PLANT UNIT NO. I - DOCKET NO.50-000 REACTOR VESSEL LEVEL INDICATINGSYSTEM

Dear Mr. Denton:

Carolina Power R Light Company (CPRL) hereby submits additional information concerning the Shearon Harris Nuclear Power Plant (SHNPP) Reactor Vessel Level Indicating System (RVLIS). This information is submitted in response to the Safety Evaluation Report (SER) license condition concerning Inadequate Core Cooling Instrumentation. As discussed on page 0-0 of SER Supplement l, the NRC requires that prior to criticality CPRL demonstrate that the SHNPP Emergency Operating Procedures (EOP) that incorporate the generic Westinghouse RVLIS system conform to generic EOP guidelines relating to the use of the RVLIS, or that deviations be identified and explained. The attached information documents CPRL's completion of this evaluation.

If you have any questions concerning this subject or require additional information, please contact me.

Yours very truly, S . Zi merman nager Nuclear Licensing Section 3HE/ccc (1670GAS)

Attachment Cct Mr. B. C. Buckley (NRC) Mr. Wells Eddleman Mr. G. F. Maxwell (NRC-SHNPP) Mr. 3ohn D. Runkle Dr. 3. Nelson Grace (NRC-RII) Mr. G. A. Schwenk Mr. Travis Payne (KUDZU) Dr. Richard D. Wilson Mr. Daniel F. Read (CHANGE/ELP) Mr. G. O. Bright (ASLB)

Wake County Public Library Dr. 3. H. Carpenter (ASLB)

Mr. 3. L. Kelley (ASLB)

'11 b.

. 85072b004i 8507105000 PDR 'ADOCN E

Fayettevilte Street ~ P. O. Box 1551 o Raleigh, N. C. 27602

V I'

ATTACHMENT The Westinghouse Owners'roup Emergency Response Guidelines (ERG) Revision I, high pressure version incorporates a generic Reactor Vessel Level Indicating System (RVLIS) which is described in the executive volume of the ERGs under the generic issues section entitled, "Reactor Vessel Liquid Inventory System." A comparison of the generic RVLIS against the Reactor Vessel Level Instrument System Manual that describes the SHNPP RVLIS system shows that the SHNPP RVLIS is consistent with the generic RVLIS in terms of instrument readout capabilities (i.e., upper range/full range/dynamic head range) and application of RVLIS reading to different plant situations.

Enclosures l and 2 contain the necessary cross-reference information to show that the plant specific use of RVLIS in SHNPP EOPs conform to the generic ERGs relating to the use of RVLIS. Enclosure 2 lists the comparison between generic and plant specific usage of RVLIS and any deviations are identified and explained.

(1674GAS/ccc )

Q.

ENCLOSURE I: EOP TITLE CROSS REFERENCE WOG CPRL Guidelines Title EPP-I ECA-O. I Loss of AC Power to IA-SA and IB-SB Busses EPP-2 ECA-O. I Loss of All AC Power Recovery Without SI Required EPP-3 ECA-0.2 Loss of All AC Power Recovery With SI Required EPP-0 ES-O. I Reactor Trip Response EPP-5 ES-0.2 Natural Circulation Cooldown EPP-6 ES-0.3 Natural Circulation Cooldown With Steam Void in Vessel (with RVLIS)

EPP-7 ES-0.0 Natural Circulation Cooldown With Steam Void in Vessel (Without RVLIS)

EPP-3 ES-I. I SI Termination EPP-9 ES-1.2 Post-LOCA Cooldown and Depressurization EPP-10 ES-1.3 Transfer to Cold Leg Recirculation EPP-I I ES-1.0 Transfer to Hot Leg Recirculation EPP-12 ECA-I. I Loss of Emergency Coolant Recirculation EPP-13 ECA-1.2 LOCA Outside Containment EPP-IO E-2 Faulted Stea'm Generator Isolation EPP-15 ECA-2.1 Uncontrolled Depressurization of All Steam Generators EPP-16 N/A SGTR Isolation EPP-17 ES-3.1 Post-SGTR Cooldown Using Backfill EPP-13 ES-3.2 Post-SGTR Cooldown Using Blowdown EPP-19 ES-3.3 Post-SGTR Cooldown Using Steam Dump EPP-20 ECA-3.1 SGTR With Loss of Reactor Coolant: Subcooled Recovery EPP-21 ECA-3.2 SGTR With Loss of Reactor Coolant: Saturated Recovery EPP-22 ECA-3.3 SGTR Without Pressurizer Pressure Control FRP-S. I FR-S. I Response to Nuclear Power Generation/ATWS FRP-S.2 FR-S.2 Response to Loss of Core Shutdown FRP-C. I FR-C. I Response to Inadequate Core Cooling FRP-C.2 FR-C.2 Response to Degraded Core Cooling FRP-C.3 FR-C.3 Response to Saturated Core Cooling FRP-H. I FR-H. I Response to Loss of Secondary Heat Sink FRP-H.2 FR-H.2 Response to Steam Generator Overpressure FRP-H.3 FR-H.3 -Response to Steam Generator High Level FRP-H.O FR-H.O Response to Loss of Normal Steam Release Capability FRP-H.5 FR-H.5 Response to Steam Generator Low Level FRP-P.I FR-P. I Response to Imminent Pressurized Thermal Shock Conditions FRP-P.2 FR-P.2 Response to Anticipated Pressurized Thermal Shock Conditions FRP-3.1 FR-Z. I Response to High Containment Pressure FRP-J.2 FR-Z.2 Response to Containment Flooding FRP-3.3 FR-Z.3 Response to High Containment Radiation Level FRP-I. I FR-I. I Response to High Pressurizer Level FRP-I.2 FR-I.2 Response to Low Pressurizer Level FRP-I.3 FR-I.3 Response to Voids in Reactor Vessel PATH-1 E-0 2 E-I Reactor Trip or Safety Injection/Loss of Reactor Secondary Coolant PATH-2 E-3 Steam Generator Tube Rupture NOTE: This cross-reference list may change depending on future WOG ERG revision and plant specific needs.

(1674GAS/ccc )

ENCLOSURE 2: RVLIS CROSS-REFERENCE SHNPP WOG EOPs ERGs Differences

l. EPP-6/la ES-0.3/la None
2. EPP-6/15 ES-0.3/5 None
3. EPP-6/2 1c ES-0.3/11c None EPP-12/13a ECA-l. 1/13a None
5. EPP-12/18a E CA-l. 1/18a None
6. EPP-21/15 ECA-3.2/16 See Note 1
7. EPP-21/19b ECA-3.2/20a See Note 1
8. EPP-22/8c ECA-3.3/7c None
9. EPP-22/12 ECA-3.3/11 None
10. Foldout I/a 2) ECA-3.2 Foldout/1 See Note 1
11. Foldout 3/a 2) ECA-3.3 Foldout/1 See Note 2
12. FRP-C. 1/6a FR-C. I/6a None
13. FRP-C.1/16c FR-C. 1/16c None
10. FRP-C.1/23 FR-C.1/23 None
15. FR P-C.2/5a FR-C.2/5a None
16. FRP-C.2/7a FR-C.2-7a None
17. FRP-C.2/18 FR-C.2/18 None
18. FR P-P. I/5 FR-P.1/5 None
19. FRP-P. 1/12 FR-P.1/12 None
20. FRP-I.3/8b FR-I.3/8b None
21. FRP-I.3/10 FR-I.3/10 None
22. FRP-I.3/18 FR-I.3/18 None
23. FRP-I.3/20 FR-I.3/20 None
20. PATH-2/J-9 E-313S None
25. CSF-2 F-0.2 None
26. CSF-6 F-0.6 None NOTE 1: The generic guidelines only allow the operator to run one RCP as part of ERG recovery actions. The plant specific EOPs allow the operator to run more than one RCP and therefore the plant specific steps account for more than one RCP running.

NOTE 2: The ERG step only gives guidance for no RCPs running. The SHNPP EOP step contains additional guidance to account for possibility that RCPs are running.

(1674GAS/ccc)

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