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 Attention:
Hr. Leon D. White, Jr.
Vice President Electric and Steam Production 89 East Avenue Rochester, New York 14649 License No. 'DPR-18 Inspection No. 75-13 Docket No. 50-244 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, 1975.
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 of. procedures and represent'ative
'records, interviews with personnel, and observations by the inspector.
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 date.
Within the scope of this inspection, no items oi noncompliance were observed.
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 Room. 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 claimed that the information is proprietary, and should be prepared so that p'xo-r ~
pLUTIPy e~
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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 Room.
I No reply to this letter is required; however, if you should have any questions concerning this inspection, we will be pleased to discuss them with you.
Sincerely, Eldon J. Brunner, Chief Reactor Operations Branch
Enclosuxe:
XE:I Inspection Report No. 50-244/75-13 cc:
C. Platt, Plant Superintendent A, E, Upton, Esquire J.
M. Hitte, QC Engineer
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Regulatory Standards (1 cy ltr, 3 cys report)
PDR 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)
U.
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 Location:
Ontario New York Category:
Safeguards
~
Group:
Type of Licensee:
T e of Inspection:
Dates of Inspection:
PWR, 1520 Nft (w)
Routine, Unannounced 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:
E.
. HcCabe, Senior Reactor Inspector Nuclear Support Section, Reactor Operations Branch-DATE
SUMMARY OF FINDINGS Enforcement Action Hone Licensee Action on Previousl Identified Enforcement Action A.
Items of Noncom liance NRC Ins ection Re ort 50-244/75-10 Details 4.c.3 and 5.d The licensee's corrective actions with respect to these items were reviewed by the inspector and found complete.
(Details 3, 4)
B.
Items of Noncomnliance NRC Ins ection Re ort 50-244/75-10 Details 5c 9.a.(3)
and 9.c.
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)
C.
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 complete.
(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 A.
Current Findin s l.
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 a.
Administrative controls for authorized management review and approval of facility, procedures.
{Detail 13)
b.
Review and approval of temporary changes to facility procedures in accordance with requirements of the Technical Specifications.
(Detail 14)
c.
Format and content of facility procedures.
{Detail 15)
d.=
Scope of procedural coverage.
(Detail 16)
e.
Frequency of calibrations as required by the Technical
"'pecifications.
(Detail 17)
2.
Unresolved Items These are items for which more information is required in order to determine whether the items are acceptable or Items of Noncompliance.
a.
Periodic review of facility procedures.
(Detail 12)
B.
Previous Unresolved Items 1.
The following items have been resolved.
a.
Generic inadequacies in procedure format and content.
(Detail 10)
b.
Inadequate scope of procedural coverage.
{Detail 9)
c.
Facility requirements for fire prevention.
(Detail ll)
Management Interview A management interview was held at the Ginna site on September 12, 1975 at the conclusion of the inspection.
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:
A.
Purpose of the. inspection.
(Detail 2)
B.
Review of previous Items of Noncompliance and Deviations.
(Details, 3, 4, 5, 6, 7 and 8)
C.
Review of previous Unresolved Items.
(Details 9, 10 and ll)
D.
Administrative controls for review and approval of facility procedures.
(Detail 13)
E.
Review and approval of temporary changes to procedures.
(Detail 5)
F.
Format and content of facility procedures.
(Detail 15)
G.
Scope of procedural coverage.
(Detail 16)
H.
Procedure changes resulting from Technical Specification revisions.
(Detail 14)
I.
Periodic review of facility procedures.
(Detail.12)
J.
Technical Specification calibration requirements.
(Detail 17)
K.
Contxol rod deviation alarm setpoint.
(Detail 18)
DETAXLS 1.
Persons Contacted'r.
S.
V. Bullock, Quality Control Technician Mr. C.
E. Edgar, Instrumentation Technician Mr. D. Horning, Head Control Operator Mr. R. B. Junot, Shift Foreman Mr. L.'. Lang, Assistant Plant Superintendent Mr. A. E. McNamara, Operations Aid Mr. J.
C. Noon, Maintenance Engineer Mr. J.
H. Paris, Instrument and Control Foreman Mr.'. E. Platt, Plant Superintendent Mr. B. R. Quinn, Health Physicist Mr. F. Schwind, Control Operator Mr. B. A. Snow, Operations Engineer Mr. J.
Sweet, 'Technical Assistant to Plant Superintendent Mr. J.
M. 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:
(2)
(3)
Licensee letter to NRC Region I dated August 29, 1975.
NRC Region I letter to licensee dated August 7, 1975.
NRC inspection report 50-244/75-10, Detail 5d.
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 shift.
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, 1975.
(2)
NRC Region I letter to licensee dated August 7, 1975.
(3)
NRC inspection report'.50-244/75-10, Detail 4c(2),
(3),
and (4).
The inspector verified by review of Administrative Procedure A-30.1
"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 completed.
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
'2)
NRC Region I letter to licensee dated August 7, 19?5.
(3)
NRC inspection report 50-244/75-10, Detail 9e.
The inspector's review of the following procedures verified the licensee's corrective'ctions as described in reference (1).
a.
HP-27, Operational Tests of Portal and Foot Honitors.
b, CP-209-1, Calibration of Portal Honitor Hodel PHC-4b.
c.
CP-209.2, Calibration Procedure for Portal Honitor Hodel PPH8.
d.
CP-209.3, Calibration Sheet for Hand and Foot Honitor Hodel HFH3..
The inspector had no further questions in this area;
6.
AO 75-04 Hain Stcam Isolation Solenoid Test Valves Reference:
(2)
(3)
Licensee letter to NRC Region I dated August 29, 1975.
NRC letter to licensee dated August 7, 1975.
NRC inspection report 50-244/75-10, Detail 5.c.
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.
7.
Record Review and Retention Reference:
(1)
Licensee letter to-NRC Region I dated August 29, 1975.
(2)
NRC letter to licensee dated August 7, 1975.
(3)
NRC inspection report 50-244/75-10, Detail 9c.
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.
8.
Standin Orders Reference:
(1)
Licensee letter to NRC Region I dated August 29, 1975.
(2)
NRC letter to licensee dated August 7, 1975.
(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.
9.
Inade uate Procedural Covera e
Reference:
NRC inspection report 50-244/74-08, Detail 2a-e.
The inspector reviewed the following procedures:
a ~
A-52.1, Shift Relief and.Turnover, Revision 1, October 17, 1974.
b.
A-46, Bypass of Safety Function or Jumper Control,, Revision 3, March 19, 197 c.
A-30, Plant Procedure Requirements, Revision 14, June 29, 1975.
d.
A-54, Ginna Station Administrative and Engineering Staff Responsibilities, Revision 1, March 19, 1975.
e.
A-54.1, Licensed Personnel Authority, Revision 0, May 5, 1975.
f.
E-25, Loss of Containment Integrity, Revision 0, November 6, 1974.
g.
E-38, Operator actions to be taken in the Event of Loss of Service Mater, Revision 0, August.l, 1975.
h.
L-26.4, Turbine Trip Mithout Reactor Trip, Revision 0, November 1, 1974.
E-15.1, Malfunction of Pressurizer Power Relay or Safety Valves, Revision 1, April 29, 1975.
E-15.2, Malfunction of Pressurizer Heaters and Spray Valves, Revision 1, November 8, 1974.
k.
A-56, Communication Systems at Ginna Station, Revision 0, May 8, 1975.
1.
M-40-3, Inspection and Maintenance of Spring Hangers, Revision 1, June 12, 1975.
m.
A-53, Preventive Maintenance Program, Revision 2, August 14, 1975.
The inspector's review verified that the licensee has provided procedural coverage for the areas indicated above.
This item is resolved.
10.
Inade uate Procedure Eormat and Content Reference:
NRC inspection.reports:
(1) 50-244/74-01, Detail 4a; and (2) 50-244/74-08, Detail 2m.
The inspector reviewed the following procedures:
I a.
E-l.l, Safety Injection System Actuation, Revision 2, July ll, 197 b.
E-l.2, Loss of Reactor Coolant, Revision 3, July 24, 1975.
c.
E-l.3, Steam Line Break Accident, Revision 0, May 9, 1975.
d.
E-l.4, Steam Generator Tube Rupture Accident, Revision 0, May 9, 1975.
e.
O-l.l, Plant Heatup from Cold Shutdown to Hot Shutdown, Revision 21, May 8, 1975.
ge 0-1.2, Plant from Hot Shutdown to Steady Load, Revision 23, August 20, 1975.
0.2.2, Plant Shutdown from Hot Shutdown to Cold Conditions, Revision 17, May 15, 1975.
h.
L'-4, Station Blackout, Revision 6, April 30,,1975.
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 a 0 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 potential.
(1)
A-17, Plant Operations Review Committee Operating Proce-dure, Revision 10, August 20, 1975.
(2)
A-30, Plant Procedure Requirements, Revision 4, July 29, 1975.
1".
'.. Exhibit 5, Emergency Maintenance Procedures.
2.
Exhibit 10, System Modification Procedures.
(3)
A-51, "Plant System Modification Activities, Revision 2, April 29;. 1975.
b.
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 c.
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 used.
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-30.
a.
Emer enc Procedures latest revision date (1)
E-33, quadrant Power Tilt (2)
. E-34, Drop of Spent Fuel Cask 11/9/72 3/27/73 b.
S stem 0 eratin Procedures latest revision date (2)
5.3.5, Deborating and Evaporator Conden-sate Demineralizer Regeneration 5.10.1, Part Length Rod Control 12/8/69 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 (2)
Di-4, Turbine E,
H. Haintenance r ~
EH-5, Changing capacitors in Pressurizer Pressure Controller 431k 7/24/70 7/27/7O
-10-
,latest
'revision date (3)
EM-8, Check for DC Control Ground and Repair on Bus. /I14 Emergency Diesel.
Generator Tie Breaker 8/5/70 (4)
EM-9, Removing Steam Flow Fl-465 from service to weld leaky isolation valve.
(5)
EM-11, Replacement or Calibration of Components in Tavg Deviation Circuitry.
(6)
EH-23, Repair of Pressurizer Power Operated Relief Valve.
(7)
EM-30, Main Steam f1eader Pressure Trans-mitter Isolation Valve.
(8)
EM-31, Checkout and Reset E.H. System.
(9)
EM-34, Repair of Diaphragm Valve 365A (Blender to RUST Isolation Valve).
(10) EM-35, Pressurizer Spray Valve PCV-431A Zero Shift of I/P converter.
1/28/71 1/16/73 5/20/71 4/16/73 10/26/70 8/10/71 9/15/71 (ll) EM-36, Repair of Control Bistable PC 430B.
9/16/71 (12) EM-37, Replacement of Manual Trip Logic Relay HTl-1 in "B" Reactor Protection Logic Train.
2/19/72 (13) EM-40, Procedure for Changing, Loading and Unloading Hercoid for Instrument Air Compressors.
7/21/72 (14) EH-41, Procedure for Changing PS-6(100V power supply) in logic cabinet.
(15) EH-42, Main S.G.
Feedwater Pump Bearing Oil Pressure Switch Adjustment.
7/27/72 ll/6/72
(16) EH-43, Tying in Sensing Point for Instru-ment Air Compressor to Mercoid PS-2032.
latest revision date 8/7/72 (17) EM-44, Replacement of Temperature Converter TH-145.
(18) EM-46, Repair of "B" Spent Resin Tank Outlet Valve, V-1687.
(19) EM-49, Replace Gate Drive and/or Direction Detector of the Part Length Control Rod System.
(20) Back Flush of A and B Spent Resin Tank Drainage Strainers.
(21) EH-51, Backwashing Spraying Filters on 1A or lB Spent Resin Storage Tank.
(22) EH-52, Repair of Check Valve 1828.
(23) EM-55, Repair of V-256.
(24) EM-56, Reapir and Test of Boric Acid Check Valve 355.
8/22/72 9/8/72 9/15/72 10/11/72 10/12/72 10/20/72 11/4/72
'll/13/72 (25) EM-57, Unplugging Purification and Normal Letdown System.
(26) EH-58, Lovering of
$i'aste Holdup Tank Mater Level for Spent Resin Removal.
(27) EH-59, Spent Resin Removal from Haste Holdup.
(28) EM-61, Location and Repair of Condenser Tube Leak.
(29) EH-62, Isolation of Reactor Coolant Pump Seal Injectipn Line.
12/8/72 12/ll/72 2/9/73 1/12/73 1/12/73
-12-latest revision date (30) EM-64, Diesel Generator Breaker Overcurrent Relay Replacement.
(31) EM-66, Charging Line Controller HCV-142.
1/12/73 4/16/73 (32) EH-67, Regeneration Heat Exchanger Charging
, Line Outlet Temperature.
(33)
EM-68", Charging Pump Controllers.
(34) EH-69, Summing Amplifier TM-4018.
4/23/73 5/1/73 5/2/73 (35) EM-70, Pressurizer Sprayline Temperature Loop
"A"'36)
EM-71, Source Range Procedure.
5/29/73 5/7/73 d.
Maintenance Procedures latest revision date (1)
M-2H, Haste Condensate Polishing Demineralizer and Resin Replacement.
4/15/73 (2)
M-3.2, Reactor Cavity Cleanup and Contamination.
2/12/73 (5)
(6)
H-14, Use of Freeze Plugs or Fluid Systems.
H-19, Annunication of Loss of Instrument Busses.
(3)
M-7.6, Haste Condensate Polishing-Demineralizer Outlet Filter Replacement.
(4)
M-11.2, Isolation of Containment Spray Pump for Maintenance or. Modification.
2/20/73 5/29/73 4/22/71 1/19/72 (7)
M-20, Replacement of Undervoltage Attachments or Breakers. in Reactor Trip Switchgear.
1/20/72
-13-latest revision date (8)
M-21, Purging H2 System for Maintenance Purposes.
1/26/72 (9)
M-22, Securing and Restoring to Service the N2 Supply Syst: em for Maintenance.
2/4/72 (10) M-25, Adjustment Procedures for Thermally Compensating Supports or Main Steam Header Safety Valves and Power Operated Relief Valves.
'/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 System.
(13) M-33, Removal of Motor Control Center 1K from Service.
1/8/73 3/5/73 Periodic Test Procedures latest revision date (1)
PT-18, Instrument Air Compressor Air Capacity Check.
1/5/72 f.
" Turbine Plant 0 eratin Procedures latest revision date (1)
T-17L, Turbine Lube Oil Reservoir Backup Vapor Entractor.
12/4/72 g.
Refuelin Procedures latest revision date (1)
RP-21, Shipping Spent Puel.
w ~
2/21/73
-14-latest'evision date (2)
RF-22, Removing a Fuel Assembly from a Fuel Shipping Cask.
(3)
RF-24, Shipping Damaged Rod Control Cluster.
h.
Refuelin Shutdown Surveillance Procedures (1)
6.0, 'Integrated Leakage Rate Test.
(2)
8.0, RTD Cross Calibration.
S stem Modification Procedures 3/2/73 5/23/73 10/3/72 11/7/72 latest revision date (1)
SH-71-1, RHR Valve Modification HOV 850A, B.
(2)
Shf-71-2, Pressurizer Spray Hinivalve
, Replacement.
3/25/71 3/1/71 (3)
SM-71-3, High Head Safety Injection Check Valve Installation.
3/1/71 (4)
SH-71-4, Charging Pump Installation.
4/5/71 (5)
SH-71-5, 'A'oop Sample Line Installation.
4/6/71 (6)
SM-72-53, Piping Modifications to Spent Resin Storage Tanks.
(7)
SH-73-1, B.A. Tank Low Level'Lockout Reset Switch.
11/9/72 3/19/73 (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 Building.
2/16/73
latest revision date (10) SM-73-5B, Relocation of Gas Release Line.
2/16/73 (ll) SM-73-6, Modification of Steam Generator (Loop "A" and "B") Bypass Valve Control System.
3/9/73 (12) SM-73-12, Emergency Diesel Air Compressor Control Switch.
5/23/73 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-30.
The number of procedures identified as potentially overdue fox review (73) represents about 7% of the list of 1038 pxocedures checked.
Xn addition, 80% of the identified pro-
. cedures involved system or component maintenance or modifica-tion of a one-time nature.
The inspector stated that all safety-related procedures should be periodically reviewed to determine their current adequacy and applicability.
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.
A-17, Plant: Operations Review Committee Operating Procedure.
b.
A-19, Changes in Written Procedures.
c.
A-30, Plant Procedure Requirement d.
A-36, Station Holding Procedure.
e.
A-50.1, R.E. Ginna Administrative Contxols Program.
f.
A-52.6, Operations Standing Order.
g.
A-54, Ginna Station Administrative and Engineering Staff Responsibilities.
h.
A-54.1, Licensed Personnel Authority.
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-cedures.
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 N18.7.
b.
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 procedures.
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 temporary changes.
I
~
~
-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 procedures.
a, Administrative Procedures (1)
A-17, Plant Operating Review Committee Operating Procedure.
(2)
A-19, Changes in Written Procedures.
(3)
A-30, Plant Procedure Requirements.
(4)
A-36, Station Holding Procedure.
(5)
A-50.1, R.E. Ginna Administrative Controls Program.
b.
System 0 eratin Procedures c ~
(1)
S-1A, Startup of Rod Drive Hotor Generator Sets.
(2)
. S-10-2, Part Length Rod Control.
(3)
S-16, Safety Injection System.
(4)
0-1.18, Establishing Containment Integrity for Operation.
(5)
T-4, Hain Feedwater Pump Operations.
(6)
S-3.2, Charging and Volume Control System.
(7)
T-32, Fire Service 4'ater System.
(8)
S-6, Nuclear Instrument System.
Alarm Procedures Those associated with procedures in b.(l)-(8) abov d.
Emer enc Procedures (1)
E-1.2, Loss of Coolant Accident.
(2)
E-5, Control Room Inaccessability.
(3)
E-22.3, Malfunction.of the Letdown System.
e.
Maintenance Procedures (1)
M-55.2, Maintenance Removal and Repair of Reactor Head Control Rod Drive cables and assemblies.
(2)
H-37.14, Inspection and Maintenance of "B" Loop Accumu-lator Discharge check Valve V-842A.
(3)
M-53, Containment Air Lock Inspection.
(4)
H-11.5D, Electrical Inspection of 1B Auxiliary Motor Driven Feed Pump.
(5)
M-ll.4.3, Isolation of lC Charging Pump for Maintenance and Restoring to Service after Haintenance.
(6)
M-52.1, Installation of Detector and Cable on Miniture Detector Flux Mapping System.
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 procedures.
a.
Reactor Coolant S stem (1)
CP-401, Tavg Channel 1.
(2)
CP-411, Reactor Coolant Flow.
b.
Reactivit and Power Control (1)
CP-35, Nuclear Instrumentation Intermediate Range.
(2)
CP-31, Nuclear Instrumentation Source Range.
'(3)
CP-41, Nuclear Instrumentation Power Range, c.
'Auxiliary S stems (1)
d, Power Conversion S stems (1)
CP-485, Turbine First Stage Pressure.
e.
~Emer eoc Cora Coolie S stems (1)
CP-938, Accumulator Level.
(2)
CP-945, Containment Pressure.
Electrical System (1)
PT-9, Undervoltage and Underfrequency 4KV Busses.
e
~
~e
~ ~
e
~,
-20-g.
Other En ineered Safet Feature (1)
CP-209, Containment Area Monitor.
h.
Emergenc Power (1)
M-60.5, Protective Relay Calibration and Trip Test.
(1A Diesel)
(2)
M-60.6, Protective Relay Calibration and Trip Test.
(1B Diesel)
i.
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 20.
<<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 level.
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, 1975.
The inspector onsite at this time was notified as to the nature of the abnormal occurrence as summarized below.
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 counter.
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, set.
The Operator commenced hand logs as required by Technical Specifica-tion 3.5 table 3.51 item 16.a.
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 room.
This item is to be reported, under Technical Specification 6.6.2.b.
(3), as a thirty day report which will be subsequently evaluated by NRC Region I.