IR 05000244/1975013
| ML18142B685 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 10/01/1975 |
| From: | Brunner E NRC/IE, NRC Region 1 |
| To: | White L Rochester Gas & Electric Corp |
| References | |
| IR 19750013 | |
| Download: ML18142B685 (24) | |
Text
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UNITED STATES NUCLEAR REGULATORY COMMlSSlON
REGION I
SSI I ARK AVENUE KING OF PRUSSIA, PENNSYLVANIA 19406 "OCT g ]g75 Rochester Gas and Electric Corporation License No. 'DPR-18 Attention: Hr. Leon D. White, J Inspection No. 75-13 Vice President Docket No. 50-244 Electric and Steam Production 89 East Avenue Rochester, New York 14649 Gentlemen:
This refers to the inspection conducted by Hr. D. Johnson of this office on September 8-12, 1975 at the Ginna facility, Ontario, New Yor'k of activities authorized by NRC License No. DPR-18 and to the discussions of our findings held by Hr. D. Johnson with Hr. Platt of your staff at the conclusion of the inspection, and to a subsequent telephone discuss-ion between Hr. D. Johnson and Hr. Snow of your staff on September 17, 197 Areas examined during this inspection are described in the Office of Inspection and Enforcement Inspection Report which is enclosed with this letter. Within these areas, the inspection consisted of selective exam-
"
inations o procedures and represent'ative 'records, interviews with personnel, and observations by the inspecto Our inspector also verified the steps you had taken to correct the Items of Noncompliance and Deviation brought to your attention in our letter dated August 7, 1975 and your response dated August 29, 1975. We- have no further questions regarding items A, B, and C of. Appendix A and the item of Deviation in Appendix B. Your actions on items D, E, and F Appendix A have not yet been completed and will be inspected at a later dat Within the scope of this inspection, no items oi noncompliance were observe In accordance with Section 2.790 of the NRC's "Rules of Practice", Part 2, Title 10, Code oi Federal Regulations, a copy of this letter and the enclosed inspection report will be placed in the NRC's Public Document Roo If this repoxt contains any inioxmation that you (ox your con-tractor) believe to be proprietary, it is necessary that you make a written application within 20 days to this office to withhold such information from public disclosure. Any such application must include a tull statement of the reasons on the basis of -which it is claimedp'xo-that the information is proprietary, and should be prepared so that r ~
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Rochester Gas and Electric Corporation prietary infoxmation identified in the application is contained in a separate part of the document. Xf we do not hear fx'om you in this regard within the specified period, the report will be placed in the Public Document Roo No reply to this letter is required; however, this I
will if you should have any questions concerning inspection, we be pleased to discuss them with yo
Sincerely, Eldon J. Brunner, Chief Reactor Operations Branch
Enclosuxe:
XE:I Inspection Report No. 50-244/75-13 cc: Platt, Plant Superintendent A, E, Upton, Esquire J. M. Hitte, QC Engineer
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IE Chief, FS&EB IE:HQ (4 cys XE Files ltr, 5 cys report)
Central I'fail & Files Directorate of Licensing (4 cys ltr, 13 cys repoxt)
Regulatory Standards PDR (1 cy ltr, 3 cys report)
Local PDR NSIC TXC REG:I Reading Room Region Directors (II, IXX, IV) (Report Only)
ELD State of New York
IE:I Form 12 (Jan 75) (Rev) ST NUCLEAR REGULATORY COHMISSION OFFICE OF INSPECTION AND ENFORCEHENT I 'I
. REGION I IE Inspection Report No: 50-244 75-13 Docket No: 50-244 Licensee: Rochester Gas and Electric Cor oration License No: DPR-18 89 Hast Avenue Priority:
Rochester New York Category:
Safeguards
~
Group:
Location: Ontario New York Type of Licensee: PWR, 1520 Nft (w)
T e of Inspection: Routine, Unannounced Dates of Inspection: September 8-12, 1975 Dates of Previous Inspection: August
,;:)7 Reporting Tnopootor: /J .~
D. F. Johnson, Reactor Inspector DATE Accompanying Inspectors: None DATE DATE DATE Other Accompanying Personnel: None DATE Reviewed By: . HcCabe, Senior Reactor Inspector DATE Nuclear Support Section, Reactor Operations Branch-
SUMMARY OF FINDINGS Enforcement Action Hone Licensee Action on Previousl Identified Enforcement Action Items of Noncom liance NRC Ins ection Re ort 50-244/75-10 Details 4.c.3 and The licensee's corrective actions with respect to these items were reviewed by the inspector and found complet (Details 3, 4) Items of Noncomnliance NRC Ins ection Re ort 50-244/75-10 Details 5c 9.a.(3) and The licensee's corrective actions with respect to these items were reviewed by the inspector and found to be in progress and not yet complete. (Details 6, 7 and 8) Item of Deviation, NRC Ins ection Re ort 50-244/75-10 Detail 9e The licensee's corrective actions with'espect to this item were reviewed by the inspector and found complet (Detail 5)
Desi n Chan es None identified.
Unusual Occurrences In accordance with the requirements of Technical Specification, Appendix A, Section 6.6.2 an abnormal occurrence was reported to NRC, Region I.
The inspector onsite was notified of the abnormal occurrence involving the control rod deviation alarm setpoint. (Detail 8)
Other Si nificant Findin s Current Findin s Acce table Areas These are areas which were inspected on a'Sampling basis and findings did not involve an .Item of Noncompliance, Deviation or Unresolved Ite Administrative controls for authorized management review and approval of facility, procedure {Detail 13) Review and approval of temporary changes to facility procedures in accordance with requirements of the Technical Specification (Detail 14) Format and content of facility procedure {Detail 15)
d.= Scope of procedural coverag (Detail 16) Frequency of calibrations as required by the Technical (Detail 17) "'pecification . Unresolved Items These are items for which more information is required in order to determine whether the items are acceptable or Items of Noncomplianc Periodic review of facility procedure (Detail 12) Previous Unresolved Items The following items have been resolve Generic inadequacies in procedure format and conten (Detail 10) Inadequate scope of procedural coverag {Detail 9) Facility requirements for fire preventio (Detail ll)
Management Interview A management interview was held at the Ginna site on September 12, 1975 at the conclusion of the inspectio Personnel in Attendance Hr. L. S. Lang, Assistant Plant Superintendent Mr. C. E. Platt, Plant Superintendent Mr. B. A. Snow, Operations Engineer e ~
j The following summarizes items discussed: Purpose of the. inspectio (Detail 2)
B. Review of previous Items of Noncompliance and Deviation (Details, 3, 4, 5, 6, 7 and 8)
C. Review of previous Unresolved Item (Details 9, 10 and ll)
D. Administrative controls for review and approval of facility procedure (Detail 13)
E. Review and approval of temporary changes to procedure (Detail 5)
F. Format and content of facility procedure (Detail 15)
G. Scope of procedural coverag (Detail 16)
H. Procedure changes resulting from Technical Specification revision (Detail 14)
I. Periodic review of facility procedure (Detail.12)
J. Technical Specification calibration requirement (Detail 17)
K. Contxol rod deviation alarm setpoin (Detail 18)
DETAXLS 1. Persons Contacted' S. V. Bullock, Quality Control Technician M C. E. Edgar, Instrumentation Technician M D. Horning, Head Control Operator M R. B. Junot, Shift Foreman M L.'. Lang, Assistant Plant Superintendent M A. E. McNamara, Operations Aid M J. C. Noon, Maintenance Engineer M J. H. Paris, Instrument and Control Foreman Mr.'. E. Platt, Plant Superintendent M B. R. Quinn, Health Physicist M F. Schwind, Control Operator M B. A. Snow, Operations Engineer M J. Sweet, 'Technical Assistant to Plant Superintendent Mr. J. Bitte, Quality Control Engineer 2. Pur ose of the Ins ection The inspector stated that the pur'pose of the inspection was to: '(1) review licensee's actions on previously identified unresolved items; (2) review licensee's corrective actions on previously identified Items of Noncompliance; (3) verify that approved procedures and procedure changes are available for safety-related activities as required by the facility Technical Specifications; and (4) verify that the calibration of components and equipment associated with safety-related systems and/or functions is in conformance with the require-ments of the Technical Specifications and approved guides and standards.
3. Verification of Status Li ht on Safe uards E uiument Reference: Licensee letter to NRC Region I dated August 29, 197 (2) NRC Region I letter to licensee dated August 7, 197 (3) NRC inspection report 50-244/75-10, Detail 5 The inspector verified the licensee's corrective actions as stated in reference (1) by review of procedure 0-6.2 "Main Control Board System Status Verification" (approved for'mplementation on September 4, 1975). The inspector further verified .by direct observation that color, coding for proper valve .3m>eup has been
added to the safeguard control board section for operator verifi-cation of valve status. In addition the licensee included on the
"Daily Surveillance Check Sheet" a requirement to visually check the safeguard valve status once per shif The inspector had no further questions in this area.
4. Administrative Procedure Authorized Review and A royal Reference: (1) Licensee letter to NRC Region I dated August 29, 197 (2) NRC Region I letter to licensee dated August 7, 197 (3) NRC inspection report'.50-244/75-10, Detail 4c(2), (3), and (4).
The inspector verified by review of Administrative Procedure A-3 "1fealth Physics and Chemistry Procedure Requirements" and Adminis-trative Procedure A-52.6 "Operations Standing Orders" that the licensee's corrective actions described in reference (1) have been complete The inspector had no further questi'ons in this area.
5. Portal Radiation Honitor Calibration Reference: (1) Licensee letter to NRC Region I dated August 29s 1975 NRC Region I letter to licensee
'2)
dated August 7, 19? (3) NRC inspection report 50-244/75-10, Detail 9 The inspector's review of the following procedures verified the licensee's corrective'ctions as described in reference (1). HP-27, Operational Tests of Portal and Foot Honitor b, CP-209-1, Calibration of Portal Honitor Hodel PHC-4 CP-209.2, Calibration Procedure for Portal Honitor Hodel PPH CP-209.3, Calibration Sheet for Hand and Foot Honitor Hodel HFH The inspector had no further questions in this area; AO 75-04 Hain Stcam Isolation Solenoid Test Valves Reference: Licensee letter to NRC Region I dated August 29, 197 (2) NRC letter to licensee dated August 7, 197 (3) NRC inspection report 50-244/75-10, Detail The licensee's corrective actions described in reference (1) were reviewed by the inspector and not yet completed. This item re-mains unresolved pending completion of the stockroom inventory. Record Review and Retention Reference: (1) Licensee letter to- NRC Region I dated August 29, 197 (2) NRC letter to licensee dated August 7, 197 (3) NRC inspection report 50-244/75-10, Detail 9 The licensee's corrective actions described in reference (1) were reviewed by the inspector and found to be in progress and not yet completed. This item remains unresolved pending licensee actions. Standin Orders Ref erence: (1) Licensee letter to NRC Region I dated August 29, 197 (2) NRC letter to licensee dated August 7, 197 (3) NRC inspection report 50-244/75-10, Detail 9a(3).
'The licensee's corrective actions described in reference (1) were reviewed by the inspector and found to be in progress and not yet completed. This item remains unresolved pending completion of the licensee's actions. Inade uate Procedural Covera e Reference: NRC inspection report 50-244/74-08, Detail 2a- The inspector reviewed the following procedures:
a ~ A-52.1, Shift Relief and .Turnover, Revision 1, October 17, 197 A-46, Bypass of Safety Function or Jumper Control,, Revision 3, March 19, 197 A-30, Plant Procedure Requirements, Revision 14, June 29, 197 A-54, Ginna Station Administrative and Engineering Staff Responsibilities, Revision 1, March 19, 197 A-54.1, Licensed Personnel Authority, Revision 0, May 5, 197 E-25, Loss of Containment Integrity, Revision 0, November 6, 197 E-38, Operator actions to be taken in the Event of Loss of Service Mater, Revision 0, August.l, 197 L-26.4, Turbine Trip Mithout Reactor Trip, Revision 0, November 1, 197 E-15.1, Malfunction of Pressurizer Power Relay or Safety Valves, Revision 1, April 29, 197 E-15.2, Malfunction of Pressurizer Heaters and Spray Valves, Revision 1, November 8, 197 A-56, Communication Systems at Ginna Station, Revision 0, May 8, 197 . M-40-3, Inspection and Maintenance of Spring Hangers, Revision 1, June 12, 197 A-53, Preventive Maintenance Program, Revision 2, August 14, 197 The inspector's review verified that the licensee has provided procedural coverage for the areas indicated abov This item is resolved.
10. Inade uate Procedure Eormat and Content Ref erence: NRC inspection.reports: (1) 50-244/74-01, Detail 4a; and (2) 50-244/74-08, Detail 2 The inspector reviewed the following procedures:
I E-l.l, Safety Injection System Actuation, Revision 2, July ll, 197 E-l.2, Loss of Reactor Coolant, Revision 3, July 24, 197 E-l.3, Steam Line Break Accident, Revision 0, May 9, 197 E-l.4, Steam Generator Tube Rupture Accident, Revision 0, May 9, 197 O-l.l, Plant Heatup from Cold Shutdown to Hot Shutdown, Revision 21, May 8, 197 .2, Plant from Hot Shutdown to Steady Load, Revision 23, August 20, 197 ge 0.2.2, Plant Shutdown from Hot Shutdown to Cold Conditions, Revision 17, May 15, 197 L'-4, Station Blackout, Revision 6, April 30,,197 The inspector's review verified that the licensee has revised the above procedures and included previous comments identified in references 1 and 2.
ll. Facilit Re uirements for Fire Prevention "
Reference: NRC inspection Report 50-244/75-07, Detail 4 a0 The licensee has revised the following procedures to provide that materials and equipment. utilized during any system modi-fication shall be reviewed for flameability and ignition source potentia (1) A-17, Plant Operations Review Committee Operating Proce-dure, Revision 10, August 20, 197 (2) A-30, Plant Procedure Requirements, Revision 4, July 29, 197 ". '.. Exhibit 5, Emergency Maintenance Procedure . Exhibit 10, System Modification Procedure (3) A-51, "Plant System Modification Activities, Revision 2, April 29;. 197 The licensee has initiated procedure A-54.2, "Ginna Station Staff Responsibilities for Fire Prevention, Revision 0, May 2X, 1975, that provides requirements for training sessions on fire prevention and instructions for maintenance of combus-tible materials and ignition'source The licensee has initiated procedure A-54.3, Cutting and
'Welding Permit, Revision 0, August'8, 1975, that establishes the position of a Fire lfarshal whose responsibilities will be to review permfts and specify fixe equipment to be use This item is resolved.
12. Periodic Review of Facilit =Procedures A-30, Plant Procedure Requirements, Section 3.5 states in part:
"Procedural Periodic Review Requirements and Responsibilities" require that "certain procedures defined in this procedure shall be reviewed on a periodic basis. The review shall not exceed 2g years."
The inspectors examination of facility procedures identified the following procedures that appear not to have been reviewed within the time frequency specified in AP-3 Emer enc Procedures latest revision date (1) E-33, quadrant Power Tilt 11/9/72 (2) . E-34, Drop of Spent Fuel Cask 3/27/73 S stem 0 eratin Procedures latest revision date 5.3.5, Deborating and Evaporator Conden-sate Demineralizer Regeneration 12/8/69 (2) 5.10.1, Part Length Rod Control 3/27/72 (3) 5.10.2, Part L'ength Rod Insertion and Removal 3/30/72 co Emer enc Maintenance Procedures latest revision date Di-4, Turbine E, H. Haintenance 7/24/70 r ~
(2) EH-5, Changing capacitors in Pressurizer Pressure Controller 431k 7/27/7O
-10-
,latest
'revision date (3) EM-8, Check for DC Control Ground and Repair on Bus. /I14 Emergency Diese Generator Tie Breaker 8/5/70 (4) EM-9, Removing Steam Flow Fl-465 from service to weld leaky isolation valv /28/71 (5) EM-11, Replacement or Calibration of Components in Tavg Deviation Circuitr /16/73 (6) EH-23, Repair of Pressurizer Power Operated Relief Valv /20/71 (7) EM-30, Main Steam f1eader Pressure Trans-mitter Isolation Valv /16/73 (8) EM-31, Checkout and Reset E.H. Syste /26/70 (9) EM-34, Repair of Diaphragm Valve 365A (Blender to RUST Isolation Valve). 8/10/71 (10) EM-35, Pressurizer Spray Valve PCV-431A Zero Shift of I/P converte /15/71 (ll) EM-36, Repair of Control Bistable PC 430 /16/71 (12) EM-37, Replacement of Manual Trip Logic Relay HTl-1 in "B" Reactor Protection Logic Trai /19/72 (13) EM-40, Procedure for Changing, Loading and Unloading Hercoid for Instrument Air Compressor /21/72 (14) EH-41, Procedure for Changing PS-6(100V power supply) in logic cabine /27/72 (15) EH-42, Main S.G. Feedwater Pump Bearing Oil Pressure Switch Adjustmen ll/6/72
latest revision date (16) EH-43, Tying in Sensing Point for Instru-ment Air Compressor to Mercoid PS-203 /7/72 (17) EM-44, Replacement of Temperature Converter TH-14 /22/72 (18) EM-46, Repair of "B" Spent Resin Tank Outlet Valve, V-168 /8/72 (19) EM-49, Replace Gate Drive and/or Direction Detector of the Part Length Control Rod Syste /15/72 (20) Back Flush of A and B Spent Resin Tank Drainage Strainer /11/72 (21) EH-51, Backwashing Spraying Filters on 1A or lB Spent Resin Storage Tan /12/72 (22) EH-52, Repair of Check Valve 182 /20/72 (23) EM-55, Repair of V-25 /4/72 (24) EM-56, Reapir and Test of Boric Acid Check Valve 35 'll/13/72 (25) EM-57, Unplugging Purification and Normal Letdown Syste /8/72 (26) EH-58, Lovering of $i'aste Holdup Tank Mater Level for Spent Resin Remova /ll/72 (27) EH-59, Spent Resin Removal from Haste Holdu /9/73 (28) EM-61, Location and Repair of Condenser Tube Lea /12/73 (29) EH-62, Isolation of Reactor Coolant Pump Seal Injectipn Lin /12/73
-12-latest revision date (30) EM-64, Diesel Generator Breaker Overcurrent Relay Replacemen /12/73 (31) EM-66, Charging Line Controller HCV-14 /16/73 (32) EH-67, Regeneration Heat Exchanger Charging
, Line Outlet Temperatur /23/73 (33) EM-68", Charging Pump Controller /1/73 (34) EH-69, Summing Amplifier TM-401 /2/73 (35) EM-70, Pressurizer Sprayline Temperature Loop 5/29/73
"A"'36)
EM-71, Source Range Procedur /7/73 d. Maintenance Procedures latest revision date (1) M-2H, Haste Condensate Polishing Demineralizer and Resin Replacemen /15/73 (2) M-3.2, Reactor Cavity Cleanup and Contaminatio /12/73 (3) M-7.6, Haste Condensate Polishing-Demineralizer Outlet Filter Replacemen /20/73 (4) M-11.2, Isolation of Containment Spray Pump for Maintenance or. Modificatio /29/73 (5) H-14, Use of Freeze Plugs or Fluid System /22/71 (6) H-19, Annunication of Loss of Instrument Busse /19/72 (7) M-20, Replacement of Undervoltage Attachments or Breakers. in Reactor Trip Switchgea /20/72
-13-latest revision date (8) M-21, Purging H2 System for Maintenance Purpose /26/72 (9) M-22, Securing and Restoring to Service the N2 Supply Syst: em for Maintenanc /4/72 (10) M-25, Adjustment Procedures for Thermally Compensating Supports or Main Steam Header Safety Valves and Power Operated Relief Valve '/15/72 (ll) M-29, Decontamination of R-18 Liquid Samplers ll/12/72 (12) M-31, Stator Replacement for Part Length Control Rod Drive Syste /8/73 (13) M-33, Removal of Motor Control Center 1K from Servic /5/73 Periodic Test Procedures latest revision date (1) PT-18, Instrument Air Compressor Air Capacity Chec /5/72 f. "
Turbine Plant 0 eratin Procedures latest revision date (1) T-17L, Turbine Lube Oil Reservoir Backup Vapor Entracto /4/72 Refuelin Procedures latest revision date (1) RP-21, Shipping Spent Pue /21/73 w ~
-14-latest'evision date (2) RF-22, Removing a Fuel Assembly from a Fuel Shipping Cas /2/73 (3) RF-24, Shipping Damaged Rod Control Cluste /23/73 h. Refuelin Shutdown Surveillance Procedures (1) 6.0, 'Integrated Leakage Rate Tes /3/72 (2) 8.0, RTD Cross Calibratio /7/72 S stem Modification Procedures latest revision date (1) SH-71-1, RHR Valve Modification HOV 850A, /25/71 (2) Shf-71-2, Pressurizer Spray Hinivalve
, Replacemen /1/71 (3) SM-71-3, High Head Safety Injection Check Valve Installatio /1/71 (4) SH-71-4, Charging Pump Installatio /5/71 (5) SH-71-5, 'A'oop Sample Line Installatio /6/71 (6) SM-72-53, Piping Modifications to Spent Resin Storage Tank /9/72 (7) SH-73-1, B.A. Tank Low Level'Lockout Reset Switc /19/73 (8) SM-73-2, Modification of Emergency Gen-erator "A" or "B" to Facilitate Trip Testing. 1/16/73 (9) SM-73-5A, Spent Fuel Cask }1andling Auxiliary Buildin /16/73
latest revision date (10) SM-73-5B, Relocation of Gas Release Lin /16/73 (ll) SM-73-6,"A"Modification
"B")
of Steam Generator (Loop and Bypass Valve Control Syste /9/73 (12) SM-73-12, Emergency Diesel Air Compressor Control Switc /23/73 The inspector stated this audit was performed on a sampling basis and did not constitute a complete review of all facility procedures requiring periodic review in accordance with A-3 The number of procedures identified as potentially overdue fox review (73) represents about 7% of the list of 1038 pxocedures checke Xn addition, 80% of the identified pro-
. cedures involved system or component maintenance or modifica-tion of a one-time natur The inspector stated that all safety-related procedures should be periodically reviewed to determine their current adequacy and applicabilit This item is unresolved pending verification of completion of the licensee's review of procedures in accordance with the requirements of A-30.
13. Administrative Controls The inspector performed an audit of the licensee's administrative controls by performing a sampling review of the following proced-Uxeso A-17, Plant: Operations Review Committee Operating Procedur A-19, Changes in Written Procedure A-30, Plant Procedure Requirement A-36, Station Holding Procedur A-50.1, R.E. Ginna Administrative Contxols Progra A-52.6, Operations Standing Orde A-54, Ginna Station Administrative and Engineering Staff Responsibilitie A-54.1, Licensed Personnel Authorit The inspector's findings, based on review of the above procedures, were that the licensee has procedures for changing, revising, re-view, approval, updating, and document contxol of facility pro-cedure No inadequacies were identified with the above procedures with respect to the requirements of Technical Specification Section 6,
"Administrative Controls," ANSI N18.7 "Administrative Controls for Nuclear Power Plants," and Regulatory Guide 1.33 "Quality Assurance Program Requirements (Operations)."
The inspector had no further questions on these items.
14. 'Tem orar Changes to Procedures a~ The inspector reviewed the Plant Operations Review Committee (PORC) meeting minutes for the year 1975 and verified that temporary changes, changes and revisions to procedures re-sulting from Technical Specification revisions, design changes, syst: em or component modifications and procedure updating have been reviewed and approved in accordance with the requirements of the Technical Specifications, the licensee's Administrative Controls and ANSI N1 The inspector verified, by examination of representative rec-ords and discussions with the licensee, that changes to pro-cedures are being made in accordance with the licensee's Administrative procedure ce The inspector verified that the control room copies of.facil-ity procedures being utilized for the operation of the facil-ity were the latest revision, were reviewed and approved by authorized management, and reflected the most recent Tech-nical Specification revisions and applicable temporaryI change ~ ~
-17-The inspector had no further questions in this area.
15. Format and Content 'of Facility Procedures The inspector conducted an audit of facility procedures by a sampling review of the following procedure a, Administrative Procedures (1) A-17, Plant Operating Review Committee Operating Procedur (2) A-19, Changes in Written Procedure (3) A-30, Plant Procedure Requirement (4) A-36, Station Holding Procedur (5) A-50.1, R.E. Ginna Administrative Controls Progra System 0 eratin Procedures (1) S-1A, Startup of Rod Drive Hotor Generator Set (2) . S-10-2, Part Length Rod Contro (3) S-16, Safety Injection Syste (4) 0-1.18, Establishing Containment Integrity for Operatio (5) T-4, Hain Feedwater Pump Operation (6) S-3.2, Charging and Volume Control Syste (7) T-32, Fire Service 4'ater Syste (8) S-6, Nuclear Instrument Syste c ~ Alarm Procedures Those associated with procedures in b.(l)-(8) abov Emer enc Procedures (1) E-1.2, Loss of Coolant Acciden (2) E-5, Control Room Inaccessabilit (3) E-22.3, Malfunction .of the Letdown Syste Maintenance Procedures (1) M-55.2, Maintenance Removal and Repair of Reactor Head Control Rod Drive cables and assemblie (2) H-37.14, Inspection and Maintenance of "B" Loop Accumu-lator Discharge check Valve V-842 (3) M-53, Containment Air Lock Inspectio (4) H-11.5D, Electrical Inspection of 1B Auxiliary Motor Driven Feed Pum (5) M-ll.4.3, Isolation of lC Charging Pump for Maintenance and Restoring to Service after Haintenanc (6) M-52.1, Installation of Detector and Cable on Miniture Detector Flux Mapping Syste Findings from review of the. above procedures indicated no discrepancies between the licensee's procedures and the format and content described in ANSI N18.7, "Administrative Controls for Nuclear Power Plants."
The inspector had no further questions in this area.
16. Sco e of Procedural Covera e Based upon review of procedure indices, examination of representa-tive records and discussions with the licensee, the inspector iden-tified no inadequacies in the procedure scope and content require-ments set forth in Regulatory Guide 1.33 Appendix A and in section 6 of the Technical Specification The inspector had no further questions in this area.
17. Calibration Fre uencies The inspector conducted an audit on a sampling basis of the follow-ing calibration procedure Reactor Coolant S stem (1) CP-401, Tavg Channel (2) CP-411, Reactor Coolant Flo Reactivit and Power Control (1) CP-35, Nuclear Instrumentation Intermediate Rang (2) CP-31, Nuclear Instrumentation Source Rang '(3) CP-41, Nuclear Instrumentation Power Range, 'Auxiliary S stems (1) .CP-476, Feedwater Flo d, Power Conversion S stems (1) CP-485, Turbine First Stage Pressur ~Emer eoc Cora Coolie S stems (1) CP-938, Accumulator Leve (2) CP-945, Containment Pressur Electrical System (1) PT-9, Undervoltage and Underfrequency 4KV Busse e ~
e
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e
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-20- Other En ineered Safet Feature (1) CP-209, Containment Area Monitor. Emergenc Power (1) M-60.5, Protective Relay Calibration and Trip Tes (1A Diesel)
(2) M-60.6, Protective Relay Calibration and Trip Tes (1B Diesel) Rod .Control (1) CP-2, Analog Rod Position.
The inspector check'ed, by review of the above procedures and exam-ination of representative records, to see whether the required fre-quency of calibrations have been met and are in accordance with the requirements of the Technical Specifications. The one discrepancy identified is described below.
Technical Specification Section 4, table 4.1-1, item 20, requires that the containment sump level be calibrated once each refueling outage. The licensee stated to the inspector that the arrangement of the existing hardware for the float switch in the containment sump is such that it cannot be calibrated, but than an operational check is performed every refueling outage to verify proper opera-tion of the float and its associated alarm circuitry.
The inspector examined the design of the float switch and concurred with the licensee that the operational checks performed appear suf-ficient to ensure proper operation of the containment sump level indication and alarm. The inspector recommended that the licensee submit to the Division of Reactor Licensing a request for a change to the Technical Specifications to reflect an"operational check in lieu of the present calibration requirement in table 4.4.1 item 2 <<Ai The licensee concurred with the inspector's comment and stated he would propose a change to the Technical Specifications in reference to the calibration requirement for the containment sump leve r
. r ~
-21-This is an unresolved item pending completion of action on a change to the Technical Specifications.
18. Rod Control Deviation Alarm Set oint In. accordance with the requirements of the Technical Specifications, Appendix A, Section 6.6.2, an observed occux'rence was x'eported to NRC, Region I'on September 8, 197 The inspector onsite at this time was notified as to the nature of the abnormal occurrence as summarized belo During routine scan of the control board, the operator'noted that there was an apparent discrepancy between the rod position indi-cators (RPI's) and the bank counte This led to the investigation of the alarm set point for the rod deviation monitor, that subsequently revealed the deviation set point was not within the +7.5" for which the alarm had been, historically, se The Operator commenced hand logs as required by Technical Specifica-tion 3.5 table 3.51 item 1 Upon further investigation if: was found that the alarm set point had been changed to +20".
This monitoring alarm is part of the plant computer and can be changed by any person using the console in the control roo This item is to be reported, under Technical Specification 6.6. (3), as a thirty day report which will be subsequently evaluated by NRC Region I.