IR 05000244/1980014
| ML17258A928 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 02/07/1981 |
| From: | Kister H, Zimmerman R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML17258A927 | List: |
| References | |
| 50-244-80-14, NUDOCS 8104030821 | |
| Download: ML17258A928 (62) | |
Text
Qs U.S.
NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT Region I Report No=.
50-244 80-14 Docket No.
50-244 License No.
DPR-18 Priority Category C
Licensee:
'ochester Gas and Electric Cor oration 89 East Avenue Rochester New York 14649 Facility Name:
R. E. Ginna Nuclear Power Plant Inspection at:
Ontario, New York Inspection conducted:
October 29 thru November 30, 1980 Inspectors:
.
~~nn~
R.
P. Zi erman, Senior Resident Inspector da e
s gned I
date signed Approved by:
H. B. Kister, Chief, Reactor Projects Section No. 4, RO
NS Branch date signed te signed Ins ection Summar
Ins ection on October 29 thru November 30, 1980 Re ort No. 50-244/80-14)
reas Ins ecte
Routine;; onsite regu ar and bac shift inspection by t e resi ent inspector (112.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />).
Areas inspected included plant operating records; maintenance and modifications; IE Bulletin response; Three Mile Island Lessons Learned implementation; periodic and special reports; Licensee Event Reports, and accessible portions of the facility during plant tours.
Results:
No, item of noncompliance were identified during this inspection.
Region I Form 12 (Rev. April 77)
81 0408 0X(
DETAILS Persons Contacted The below listed technical and supervisory level personnel were among those contacted:
W. Backus, Operations Supervisor J. Bodine, gC Engineer L. Boutwell, Maintenance Supervisor W. Dillion, Supervisor of Nuclear Security C. Edgar, I
C Supervisor D. Filk'ens, Supervisor.Health Physics and Chemistry D. Gent, Results and Test Supervisor A. Harhay, Plant Chemist G. Larizza, Technical Engineer R. Morrill, Training Coordinator T. Meyer, Nuclear Engineer J.
C, Noon, Assistant Plant Superintendent C. Peck, Operations Engineer B. guinn, Health Physicist B. A. Snow, Plant Superintendent S. Spector, Maintenance Engineer J. Straight, Fire Protection and Safety Coordinator The inspector also interviewed arid talked with other licensee personnel during the course of the inspection.
Review of Plant 0 erations a 0 General b.
The inspector reviewed plant operations through direct inspection and observat'ion throughout the reporting period.
A plant shutdown commenced on 11/1 for a scheduled maintenance outage.
Activities accomplished included performance of tube sleeving in the "B" Steam Generator, crevice cleaning of both steam generators, fire protection modifications, and seismic support modifications.
Plant startup was in progress at the close of 'the reporting period.
Shift Lo s and 0 eratin Records Operating logs and records were reviewed against Technical Specifications
and Administrative Control Procedure requirements.
Included in the review were:
Control Room Log Daily Surveillance Log Shift Supervisor's Log Plant Recorder Traces Plant Process Computer Station Event Reports daily during control room surveillance daily during control room surveillance daily during control room surveillance daily during control room surveillance daily during control room surveillance ll/1 through 11/31'he logs and records were reviewed to verify that entries are properly made; entries involving abnormal conditions provide sufficient detail to communicate equipment status, deficiencies, corrective. action restoration and testing; records are being reviewed by management; operating orders do not conflict with the Technical Specifications; logs and incident reports detail no violations of Technical Specifi-cation or reporting requirements; logs and records are maintained in accordance with Technical Specification and Administrative Control Procedure requirements.
Several entries in these logs were the subject of additional review and discussion with licensee personnel.
No unacceptable conditions were identified.
Plant Tour 1.
During the course of the inspection, tours of the following areas were conducted:
Control Room Containment Building Auxiliary Building Intermediate Building (including control point)
Turbine Building Diesel Generator Rooms Condensate Polishing Facility Battery Rooms
C I
Relay Room Yard Area and Perimeter The following observations resulted from the tour s:
a 0 b;
Monitorin instrumentation.
Process instruments were observed or corre at>on etween c annels and for conformance with Technical Specification requirements.
Annunciator alarms.
Yar ious alarm conditions which had been receive an ac nowledged were observed.
These were discussed
'with shift personnel to verify that the reasons for the alarms were understood-and corrective action, if required, was being taken.
c.
Shift mannin
.
Control room and shift manning were observed or conformance with 10 CFR 50.54 (K), Technical Specifications and site administrative procedures.
d.
e.
g, h.
Radiation rotection controls.
Areas observed included control point operation, posting o
radiation and high radiation areas, compliance with Radiation Work Permits and Special Work Permits, personnel monitoring devices being properly worn, and personnel frisking practices.
E ui ment lineu s.
Selected valves and electrical breakers were ver>fred to be 1n the position or condition required by Technical Specifications and plant lineup procedures for the applicable plant mode.
This verification included control board indication and field observation of valve and breaker position.
E ui ment ta in
.
Selected equipment, for which tagging requests a
een instigated, was observed to verify that the tags were in place and the equipment in the condition specified.
Fire rotection.
Fire detection and fire fighting equipment was o serve or conformance with Technical Specifications and site administrative procedures.
~Securit
.
Areas observed for conformance with regu1atory requirements, the site security plan and administrative procedures, included vehicle and personnel access, protected and vital area integrity, escort and badgin Plant housekee in controls.
Plant conditions were observed for conformance with sste administrative procedures.
Storage of material and components was observed with respect to prevention of fire and safety hazards.
Housekeeping was evaluated with respect to controlling the spread of surface and airborne contamination.
No items of noncompliance were identified.
3.
Plant Maintenance and Modificatio'ns.
a ~
During the inspection period, the inspector observed various maintenance and problem investigation activities.
The
.inspector.
reviewed these activities to verify compliance with regulatory requirements, including those stated in the Technical Specifications; compliance with administrative and maintenance procedures; compliance with applicable codes and standards; required gA/gC involvement; proper use of safety tags; proper equipment alignment and use of jumpers; personnel qualifications; radiological controls for worker protection; fire protection; retest requirements and ascertain reportability as required by
. Technical Specifications.
In a similar manner the implementation of design changes and modifications were reviewed.
Compliance with requirements to update procedures and drawings were verified and post modification acceptance testing was evaluated.
b.
The. following activities were included during this review.
Installation of five tube sleeves in the "B" Steam Generator.
(Procedure ST-80-2714F.1, Steam Generator Test Sleeve Installation, Revision 0, November 15, 1980)
Repair of weld area for local temperature instrument (TI 627)
located on RHR return line.
No items of noncompliance were identified.
Followu on IE Bulletins (IEB)
The inspector reviewed facility-records, interviewed licensee personnel and observed facility equipment/components to verify that:
licensee management received and reviewed the bulletins in, accordance with administrative procedures;
information discussed in the licensee's bulletin response
.
was accurate:
corrective action was taken as discussed in the reply; and, the licensee s response was within the time period required.
IEB 79-21, Temperature Effects on Level Heasurements.
Reference:
IE Inspection Report 79-15 (79-BU-21)
The inspector reviewed licensee records documenting that the required operator training was performed from September 17, 1979 through October 15, 1979.
This bulletin is closed.
The inspector noted that in some instances the training lesson subject was documented under a broad category on the class attendance sheets, which made it difficult to determine the specific subject taught.
Further, the inspector observed that retrieval of the above records appeared hampered due to the lack of specificity in documenting the training lesson subject.
The licensee representative stated that the need for more specific documentation associated with the subject of a training class will be reviewed.
This area will be addressed in future inspections.
5.
Im lementation of Three Nile Island TNI Lessons Learned - Cate or Items a.
The inspector reviewed the licensee's actions on "short term" requirements resulting from the NRC staff investigations of the THI accident.
b.
Each item is categorized by the number assigned in NUREG 0578 unless otherwise stated.
2.2.l.b Shift Technical Advisor (STA)
~R References:
a)
NUREG 0578 b) H. Denton (NRC) letter to All Operating Nuclear Power Plants, dated October 30, 1979.
(c) D. Crutchfield (NRC) letter to L. White, Jr.
(RGSE), dated July 7, 1980.
Each licensee shall provide an STA to the Shift Superviso The STA shall have either (1)
a bachelor's degree or equivalent in a scientific or engineering discipline or, (2) hold an NRC Senior Reactor Operator (SRO) license.
Licensee Commitments References:
(a)
L. White, Jr.
(RGSE) letter to D. Ziemann (NRC), dated December 28, 1979.
(b)
L. White, Jr.
(RG8E) letter to D. Crutchfield (NRC), dated August 5, 1980.
The licensee representative agreed to augment the shift operating organization with the assignment of an STA.
Further, only those individuals who hold an SRO license or who have a bachelor's degree in a scientific or engineering discipline would be assigned as the STA.
Ins ection Findin s
Based on the inspector 's review of the following documentation, the licensee appears to have satisfied the above requirement and associated commitment.-
The inspector also interviewed licensee personnel current-ly serving as STA's regarding their educational background.
A-52.1, Shift Organization and Responsibilities, Revision 9, October 30, 1980.
A-201, Ginna Station Administrative and Engineering Staff Responsibilities, Revision 8, August 1, 1980.
Shift Manning Schedule In general, the licensee does not actively seek final college tran-scripts for those individuals who receive employment offers prior to their graduation.
In the case of those personnel serving as STA's, completed college transcripts are currently being sought by the licensee, and will be maintained as part of the individual's personal record.
2.2.l.a.
Shift Su ervisor Res onsibilities S ecified Re uirements References:
(a)
(b) H. Denton (NRC) letter to All Operating Nuclear Power Plants, dated October 30, 1979.
(c) D. Crutchfield (NRC) letter to L. White, Jr.
(RG&E), dated July 7, 1980.
The highest level of corporate management of each licensee shall issue a management directive that emphasizes the primary management responsibility of the Shift Supervisor for safe operation of the plant under all conditions on his shift and that clearly establishes his command duties.
Plant procedures shall be reviewed to assure that the duties, responsibilities, and authority of the Shift Supervisor and control room operators are properly defined.
Emphasis shall be placed on the following:
a;-
The Shift Supervisor shall maintain the broadest perspective of operational conditions effecting the safety of the plant.
b.
-The Shift Supervisor, until properly relieved, shall remain in the control. room at all times during accident situations to direct. the activities of control room operators
, Persons authorized to relieve the Shift Supervisor shall be specified.
c.
Assumption of authoritative duties by the lead control room operator upon temporary absence of the Shift Supervisor from the control room.
Training programs for Shift Supervisors shall emphasize safe operation.
Administrative duties of the Shift Supervisor shall be reviewed by the senior officer of each utility responsible for plant operations.
Licensee Commitments References:
(a) L. White, Jr.
(RGSE) letter to D. Ziemann (NRC), dated October 17, 1979.
(b) L. White, Jr.
(RGSE) letter to D. Ziemann (NRC), dated December 28, 1979.
The licensee representative agreed to revise, as necessary, the responsibilities'f the Shift Foreman such that command oversight of. operations is provided and management review of ongoing operations important to safety is performe F
J
Ins ection Findin s Based on the inspector's review of the following documentation, the licensee appears to have satisfied the above requirement and associat-ed commitment.
L. D. White, Jr.
(RG&E) memorandum to Ginna Station Personnel, dated December 13, 1979.
Subject-Shift Supervisor's Respon-sibilities.
Procedure A-52.1, Shift Organization and Responsibilities, Revision 9, October 30, 1980.
Procedure A-201, Ginna Station Administrative and Engineering Staff Responsibilities, Revision 8, August 1, 1980.
The licensee representative has addressed the requalification train-ing program for all personnel which are qualified to assume the Shift Foreman position, and will further stress the responsibilities and management function for safe operation of the facility during this training.
NUREG 0660 TAP No. I.C.I.
~R References:
Short Term Accident and Procedures Review (a)
NUREG 0660 b)
NUREG 0578 (c H. Denton (NRC) letter to All Operating Nuclear Power Plants, dated October 30, 1979.
Develop emergency operating instructions for handling small-break loss-of-coolant accidents and conduct operator retraining.
Licensee Commitments References:
(a) L. White, Jr.
(RG8E) letter to D. Ziemann (NRC), dated October 17, 1979.
The licensee representative agreed to develop procedures and conduct operator training:
Ins ection Findin s Inspection results are documented in IE Inspection Report 80-0 'I
Licensee corrective action for deviation 80-04-01 and unresolved item 80-04-02, identified in the above inspection report,. were reviewed by the inspector and found acceptable.
2.2.1.c Shift and Relief Turnover Procedures Re uirements References:
(a NUREG 0578 (b H; Denton (NRC) letter to All Operating Nucl'ear Power Plants, dated October 30, 1979.
(c)
D. Crutchfield (NRC) letter to L. White, Jr.
(RG&E), dated July 7, 1980.
A checklist which includes the following items shall be provid-ed for the oncoming and offgoing control room operators and shift-supervisor:
a.
critical plant parameters are within allowable limits b.
assurance of proper system alignment c.
identification of components in a degraded mode of operation Checklist or logs shall be provided for completion by the off-going and oncoming auxiliary operators and technicians.
A system shall be established to evaluate the effectiveness of the shift and relief turnover procedures.
Licensee Commitments References:
(a) L. White, Jr.
(RG&E) letter to D. Ziemann (NRC), dated October 17, 1979.
(b)
L'. White, Jr.
(RG&E) letter to D. Ziemann (NRC), dated December 28, 1979.
The licensee representative agreed to formalize the transfer of operational information between shifts in accordance with the above requirements.
t Ins ection Findin s Based on the inspector's review of the following documentation, the licensee appears to have satisfied the above requirement and associat-ed commitment.
The inspector also interviewed operating shift personnel to determine the adequacy of the exchange of information during relief turnove Procedure A-52.4, Control of Limiting Conditions for Operating Equipment, Revision 23, August 7, 1980.
Procedure A'-52.5, Control of Limiting Conditions for System Specifications, Revision 6, March 3, 1980.
Procedure A-52.1.4, Health Physics and Chemistry Shift Change-over, Revision 1, January 14, 1980.
Procedure 0-6.13, Daily Surveillance Log, Revision 9, December 11, 1980.
e Procedure 0-9, Shift Relief Turnover-Control Room, Revision 2, June 26, 1980.
'rocedure 0-9;--1, Shift Relief Turnover-Auxiliary Operator, Revision 4, December 11, 1980.
The inspector noted that Procedure 0-9 allowed the oncoming Shift Foreman to review and sign the Daily Surveillance Log following shift
.turnover.
The inspector stated that the Shift Foreman should be aware of all information contained on the Daily Surveillance Log prior to shift relief in order to ensure plant status is fully understood.
The licensee r'6presentative acknowledged the inspector's comment, and issued,a procedure change on October 30, 1980 requiring Shift Foreman review and signature prior to relieving the shift.
Periodic independent verification of valve lineups serves as a means of evaluating the effectiveness of the shift turnover procedures.
2.2.2.a Control Room Access Re uirements References:
(a)
NUREG 0578 (b) H. Denton (NRC) letter to All Operating Nuclear Power Plants, dated October 30, 1979.
(c)
D. Crutchfield (NRC) letter to L. White, Jr.
(RGSE), dated July 7, 1980.
Develop and implement an administrative procedure that establishes the authority and responsibility of the person in charge of the control room to limit access.
Develop and implement procedures that establish a clear line of authority and responsibility in the control room in the event of any emergenc Licensee Commitments References:
(a) L. White, Jr.
(RGIIE) letter to D. Ziemann (NRC), dated October 17, 1979.
(b) L. White, Jr.
(RGIIE) letter to D. Ziemann (NRC), dated December 28, 1979.
The licensee representative agreed to make provisions for limiting access to the control room.
Further, it was stated that the authority and responsibilities of plant personnel reporting to the Control Room, Technical Support Center and the Operational Support Center would be described in plant procedures.
Ins ection Findin s Based on the inspector's-review of the following documentation, the licensee appears to have satisfied the above requirement and associated commitment.
- A-52.1, Shift Organization and Responsibilities, Revision 9, October 29, 1980.
SC-1.3D, Manning the Technical Support Center, Revision 3, October 14, 1979.
SC-1~3E; Manning the Operational Support Center, Revision 0, December 27, 1979.
SC-1. 3A, Site Radiation Emergency (Shift Supervisor and Control Room), Revision 12, May 10, 1980.
SC-1.3B, Site Radiation Emergency (Emergency Coordinator and Survey Center Assignees),
Revision 13, January 24, 1980.
2.1.8.a Post Accident Sam lin Re uirements References:
(a)
NUREG 0578 (b) H. Denton (NRC) letter to All Operating Nuclear Power Plants, dated October 30, 1979.
(c)
D. Crutchfield (NRC) letter to L. White, Jr.
(RGRE), dated July 7, 1980.
Perform a review of the reactor coolant and containment atmos-phere sampling systems to determine the capability of promptly obtaining a sample under accident condition Perform a review to determine the capability to promptly quantify various radioisotopes that are indicators of the degree of core damage.
Procedures shall be provided to perform boron and chloride chemical analysis assuming a highly radioactive initial sample.
Licensee Commitments References:
(a) L. White, Jr.
(RGRE) letter to D. Ziemann (NRC), dated October 17, 1979.
(b) L. White, Jr.
(RGIIE) letter to D. Ziemann (NRC),, dated November 19, 1979.
(c) L. White, Jr.
(RGSE) letter to D. Ziemann (NRC), dated December 28, 1979.
The licensee representative agreed to perform the required review and based'on the results make equipment and procedural improvements as necessary in order to minimize personnel radiation exposure.
Ins ection Findin s References:
(a) Design Review of Plant Shielding and Environmental gualification of Equipment For Spaces/Systems which may be used in Post Accident Operations Outside Contain-ment, December 1979.
(b) Procedure PC-23.1, Emergency Sampling of Primary Coolant, Revision 2, May 10, 1980.
(c) Procedure PC-23.2, Containment Atmosphere Sampling and Analysis, Revision 2, October 13, 1980.
(d) Procedure PC-4, Hydrogen Concentration and Radiogas Activity in Primary Coolant Sampling and Analysis, Revision 4, February 15, 1979.
A review was performed by the licensee to determine the improvements necessary for the prompt collection, handling and analysis of post accident samples with the intent of minimizing personnel radiation exposure.
Plant procedures were developed for determining the dis-solved hydrogen concentration, dissolved noble gas activity, presence of'amma emitting isotopes, boron concentration and chloride concen-
~tration in a highly radioactive reactor coolant sample.
PC-23.1, Emergency Sampling of Primary Coolant, references additional
procedures'n the body of the sampling procedure which are necessary in order to perform various emergency sampling analyses.
Although the licensee appears to have satisfied the above requirement and associated commitment, concerns on the inconsistencies encountered between PC-23.1 and the referenced procedures were identified and will be addressed in IE Inspection Report 80-16
-'.:..-
2.1.3.a Direct Indication of Power-0 crated Relief Valve and Safet Va ve Position Re uirements References:
a)
NUREG OS78
')
H. Denton (NRC) letter to All Operating Nuclear Power Plants, dated October 30, 1979.
(c)
D. Crutchfield (NRC) letter to L. Wh'ite,
~
Jr.
(RG8E), dated July 7, 1980.
Reactor system relief and safety valves shall be provided with a positive indication in the control room derived from a reliable valve position detection device or'
reliable indication of flow in the discharge pipe.
Valve position should be seismically and environmentally qualified.
Licensee Commitments References:
(a) L. White, Jr.
(RGSE) letter to D. Ziemann (NRC), dated October 17, 1979.
(b) L. White, Jr.
(RG8E) letter to D. Ziemann (NRC), dated November 19, 1979.
'c) L. White, Jr.
(RGSE) letter to D. Ziemann (NRC), dated December 28, 1979.
The licensee representative stated that the following actions had been taken to satisfy the above requirement.
A control room annunciator has been added to the existing power-operated relief valve stem mounted limit switch in-dication.
Linear variable displacement transducers (LVDTs) with control room position alarm were installed on the pressurizer safety valve Environmental qualification has been reviewed with respect
'o accidents associated with safety and PORV lifting.
The materials and components were found suitable for such en-vironments.
Ins ection Findin s
Based on the inspector's observations in the control room and review of the following documentation,'he licensee appears to have satisfied the above requirement and associated commitment.
EWR 2603, Environmental gualification Report on Safety:
Valve Position Indication, Revision 0, October 31, 1980.
Certificate of Compliance for 4 NAMCO limit switches, Purchase Order Number 001,968.
2.1.7.a Automatic Initiation of the Auxiliar Feedwater AFW
" ~Sstem
~R References:
(a)
NUREG 0578 (b) H. Denton-(NRC).letter to All Operating Nuclear Power Plants, dated October 30, 1979.
(c) D. Crutchfield (NRC) letter to L. White, Jr.
(RG&E), dated July 7, 1980.
Provide automatic/manual initiation of AFWS.
Provide testability of the initiating signals and circuits.
Initiating signals and circuits shall be powered from the emergency buses.
The system shall be designed so that a single failure will not result in the loss of AFW function.
Licensee Commitments References:
(a) L. White, Jr.
(RG&E) letter to D. Ziemann (NRC), dated October 17, 1979.
(b) L. White, Jr.
(RG&E) letter to D. Ziemann (NRC), dated December 14, 197 (c) L. White, Jr.
(RGSE) letter to D. Ziemann.
(NRC), dated December 28, 1979.
The licensee representative stated that the design configuration and operation of the AFW system satisfied the above requirements.
Ins ection Findin s Based on the inspector's observations in the control room and review of the following documentation, the licensee appears to have satisfied the above requirement and associated commitment.
Ginna FSAR Westinghouse logic drawings 882D612 - SH 6 Safeguards Actuation Signals
-
SH 7 Safeguards Sequence
- SH 8 Feedwater Isolation and Auxiliary Feed-water Pump The AFW system consists of one turbine-driven and two motor-driven pumps.
Each motor-driven pump feeds one steam generator.
The turbine-driven pump feeds both steam generators.
The motor-driven pumps, powered from separate redundant 480 v emergency buses, a. e started by developing the automatic initiation logic for either loss of main feedwater, steam generator low-low level, or a safety injection signal.
The turbine-driven pump is started automatically on both steam generators at low-low level or loss of voltage on both 4160 volt buses.
Initiating signals are demonstrated functional at refueling intervals through performance of surveillance testing.
2.1.7.b Auxiliar Feedwater Flow Indication to Steam Generators Re uirements References:
a)
NUREG 0578 b)
H. Denton (NRC) letter to All Operating Nuclear Power Plants, dated October 30, 1979.
(c)
D. Crutchfield (NRC) letter to L. White, Jr.
(RGSE), dated July 7, 1980.
Auxiliary feedwater flow indication should satisfy the single failure criteria for each steam generator.
One auxiliary
y
feedwater flow channel may be backed up by a steam generator level channel.
Auxiliary feedwater flow instrument channels shall be powered from the vital instrument buses.
Testability shall be a feature of the design.
Licensee Commitments References:
(a) L. White, Jr.
(RGSE) letter to D. Ziemann (NRC), dated October 17, 1979.
(b) L. White, Jr.
(RGSE) letter to D. Ziemann (NRC), dated November 19, 1979.
(c) L. White, Jr.
(RGSE) letter to D. Ziemann (NRC), dated December 14, 1979.
(d) L. White, Jr.
(RGSE) letter to D. Ziemann (NRC), dated December 28, 1979.
The licensee stated that the AFW flow instruments for each steam generator met the above requirements.
In order to satisfy the single failure criteiia, redundant indication of auxiliary feedwater flow is provided by steam generator level.
Ins ection Findin s Based on the inspector's observations in the control room and review of the following documentation, the licensee appears to have satisfied the above requirement and associated commitment.
RG8E drawing 10904-105, Auxiliary F.W. Flow Indication Power Supply, April 27, 1980.
EWR 2836, Design Criteria - Power Supply for Auxiliary Feedwater Flow Indicators, Revision 0, April 9, 1980.
.In May, 1980, the Auxiliary Feedwater flow indication was modified to provide redundant flow indication for each motor driven auxiliary feedwater pump and the common discharge of the turbine driven aux-iliary feedwater pump.
Prior to the modification, a steam generator level channel was used as backup to the AFW flow indication to satisfy the single failure criteri.1.1 Re uirements Emer enc Power Su
- Pressurizer Heaters References:
(a)
NUREG 0578 (b H. Denton (NRC) letter to All Operating Nuclear Power Plants, dated October 30, 1979.
(c)
D. Crutchfield (NRC) letter to L. White, Jr.
(RGSE), dated July 7, 1980.
The number of pressurizer heaters necessary to maintain natural circulation at hot standby conditions shall have
= 'he capability of being supplied from the emergency buses.
Procedures and training shall be established for reloading the pressurizer heaters onto the emergency power buses.
Licensee Commitments References:
(a) L. White, Jr.
(RGIIE) letter to D. Ziemann (NRC), dated October 17,'979.
(b) L. White, Jr.
(RGIIE) letter to D. Ziemann (NRC), dated November 19, 1979.
(c) L. White, Jr.
(RGSE) letter to D. Ziemann (NRC), dated December 28, 1979.
The licensee representative stated that the pressurizer heater power supply satisfied the above requirement.
A procedure was written in-cluding criteria to prevent overloading of~a diesel generator while providing for sufficient power to overcome heat losses and maintain natural circulation.
Operator. training in the use of the procedure was performed.
Ins ection Findin s
Based on the inspector's review of the following documentation, the licensee appears to have satisfied the above requirement and associ-ated commitment.
Ginna FSAR Procedure 0-8.1, Restoration of Pressurizer Heaters to Main-tain Natural Circulation at HSD, Revision 1, January 24, 198 t P
RGSE wiring diagram 10905-89 - Pressurizer Heater Control Group 10905-90 - Pressurizer Heater Backup Group Training records associated with walk-through and classroom instruction for Procedure E-l.l, Immediate Actions and Diag-nostics for Spurious actuation of SI, LOCA, Loss of Secondary Coolant and Steam Generator Tube Rupture.
The pressurizer proportional and backup heaters are powered from sepa'rate redundant 480 v emergency buses which can receive power from either onsite or offsite sources.
The heaters are load shed upon a safety injection signal and an undervoltage signal.
Manual reconnection of the heaters to the buses with adquate heater capacity for natural circulation operation has been incorporated in plant pro-cedures.
2.1.5.c Re uirements Recombiner Procedures Review and U
rade References:
(a)
NUREG 0578 (b) H. Denton (NRC) letter to All Operating Nuclear Power Plants, dated October 30, 1979.
(c)
D. Crutchfield (NRC) letter to L. White, Jr.
(RGSE), dated July 7, 1980.
Procedures and bases upon which recombiners would be used should be reviewed considering shielding requirements and personnel exposure limitations.
Licensee Commitments References:
(a) L. White, Jr.
(RG8E) letter to D. Ziemann (NRC), dated October 17, 1979.
(b) L. White, Jr.
(RGSE) letter to D. Ziemann (NRC), dated November 19, 1979.
(c) L. White, Jr.
(RG&E) letter to D. Ziemann (NRC), dated December 28, 1979.
The licensee representative stated that access to the hydrogen con-trol panel under post accident conditions was evaluated and found to be possible without causing undue radiation to personne I I
Ins ection Findin s Based on the inspector's review of the following documentation, the licensee appears to have satisfied the above requirement and associated commitment.
Design Review of Plant Shielding and Environmental gualif-ication of Equipment for Spaces/Systems which may be used in Post'ccident Operations Outside Containment, December, 1979.
Procedure S-21.1, 1A Hydrogen Recombiner Purging and Oper-ation, Revision 0, April 5, 1975.
Procedure S-21.2, 1B Hydrogen Recombiner Purging and Oper-ation, Revision 0, April 3, 1975.
The above referenced design review considered an individual to spend approximately. 30 minutes at the control panel for starting and controlling the hydrogen recombiner system.
.The estimated dose to an individual in this area for the necessary occupancy would be 1.6 rem.
2.1.3.b Instrumentation for Detection of Inade uate
~C Re uirements References:
(a)
NUREG 0578 (b H. Denton (NRC) letter to All Operating Nuclear Power Plants, dated October 30, 1979.
(c) D. Crutchfield (NRC) letter to L. White, Jr.
(RG8E), dated July 7, 1980.
Install a primary coolant saturation meter to provide on-line indication of coolant saturation condition.
Develop procedures to be used by the operator to recognize inadequate core cooling.
Licensee Commitments References:
(a) L. White, Jr.
(RGSE) letter to D. Ziemann (NRC),.dated October 17, 1979.
(b) L. White, Jr.
(RGSE) letter to D. Ziemann (NRC), dated November 19, 1979.
(c) L. White, Jr.
(RG&E) letter to D. Ziemann (NRC), dated December 28, 197 The licensee representative stated that two independent analog subcooling meters were installed to provide a continuous display of the temperature margin to saturation.
Procedures were revised to include the subcooling meters as a possible means of recogniz-ing inadequate core cooling.
Ins ection Findin s Based on the inspector's observations in the control room and re-view of the following documentation, the licensee appears to have satisfied the above requirement and associated commitment.
EWR 2604, Safety Analysis - Detection of Inadequate Core Cool-ing, Reactor Coolant System Subcooling Margin Monitoring System, Revision 0, December 5, 1979.
EWR 2604, Design Criteria - Detection of Inadequate Core Cool-ing Reactor Coolant System Subcooling Margin Monitoring System, Revision o, December 4, 1979.
Procedure E-1.5, Void Formation in the Reactor Coolant System, Revision 7, June 26, 1980; The existing hot leg temperature and pressurizer pressure inputs to each meter have a range of 500 - 700 F and 1750 - 2500 psig, respec-tively.
The temperature and pressure inputs are scheduled to be up-graded to 330 - 700 F and 15 - 2500 psia, respectively during the 1981 spring refueling outage.
2.1.1 Emer enc Power Su
- Pressurizer Level and Re se B oc a ves Re uirements References:
(a)
NUREG 0578 (b H. Denton (NRC) letter to All Operating Nuclear Power Plants, dated October 30, 1979.
(c) D. Crutchfield (NRC) letter to L. White, Jr.
(RGEE), dated July 7, 1980.
Pressurizer level indication instrument channels shall be powered from vital instrument buses.
Motive and control components of the power-operated relief valves (PORVs)
and associated block valves shall be capable of being supplied fr'om offsite or onsite emergency source Licensee Commitments
'References:
(a) L. White, Jr.
(RGSE) letter to D. Ziemann (NRC), dated October 17, 1979.
(b) L. White, Jr.
(RGIIE) letter to D. Ziemann (NRC), dated November 19, 1979.
The licensee representative stated that emergency power supplies for the PORVs, block valves and pressurizer level indicators satisfied the above requirement.
Ins ection-Findin s
Based on the inspector's review of. the following documentation, the licensee appears to have satisfied the above requirement and associ-ated commitment.
Ginna FSAR RGIIE Drawing 21489-299, Revision 0, Pressurizer PORY/Block Valve.
The PORVs are air operated valves which fail closed on a loss of air.
The air compressors, the source of supply air, are not supplied from the emergency buses.
Nitrogen accumulators located in containment, whi.ch provide the motive source to the low temperature overpressure
'rotection system would be used as a backup in the event of a loss of air.
The power for the solenoid valves associated with the PORVs is supplied from the emergency buses.
Each block valve is powered from a different emergency bus than the respective PORV solenoid valve.
The pressurizer level indicating instrument channels are powered from vital instrument buses.
2.2.2.b 'nsite Technical Su ort Center TSC Re uirements References:
a)
NUREG 0578 b)
H. Denton (NRC) letter to All Operating Nuclear Power Plants, dated October 30, 1979.
(c)
D. Crutchfield letter to L. White, Jr.
(RGIIE), dated July 7, 198 Establish a
TSC and provide a complete description..
Install dedicated communications.
Provide monitoring for direct radiation and airborne con-taminants.
Display of plant parameters for a'ssessing occurrence.
Submit a long range plan for upgrading TSC.
Licensee Commitments References:
(a) L. White, Jr.
(RGSE) letter to D. Ziemann (NRC); dated October 17, 1979.
(b) L. White, Jr.
(RGSE) letter to D. Ziemann (NRC), dated November 19, 1979.
(c) L. White, Jr.
(RG5E) letter to D. Ziemann (NRC), dated December 28, 1979.
The licensee representative agreed to establish a
TSC where plant-status information would be available to responsible engineering and management personnel in the event of an'ccident.
The follow-ing actions were said to have been taken to satisfy the above re-quirements.
Plans and procedures for support and staffing of the TSC were developed.
The below listed equipment was made available in the TSC.
a.
controlled drawings required to assess an incident.
b.
an NRC direct line extension (ENS)
c.
plant computer data link d.
radiation monitor e.
storage cabinet containing charcoal cannisters, DC battery lighting packs and potassium iodide tablets.
f.
hard wired intercom system connecting TSC, control room, emergency center, standby emergency center and onsite operational support center.
Ins ection Findin s
Based on the inspector's observations at the TSC, the licensee appears to have satisfied the above requirement and associated commitment.
The inspector also reviewed the following documenta-tio I
Procedure SC-1.3D, Manning the Technical Support Center Revision 3, April 14, 1980.
Procedure SC-1.18, Administration of Potassium Iodide, Revision 0, February 26, 1980.
During the tour of the TSC, the inspector noted that although port-able air samplers were available in close proximity to the TSC, a
sampler was not located in the TSC.
The licensee representative acknowledged the inspector's observation and placed a portable air sampler in the TSC.
2.2.2.c
~ti t
'eferences:
Onsite 0 erational Su ort Center a
H. Denton (NRC) letter to All Operating Nuclear Power Plants, dated October 30, 1979.
(c) D. Crutchfield (NRC) letter to L. White, Jr.
(RG&E), dated July 7, 1980.
Establish an onsite center where operations support personnel will report in an emergency situation.
The emergency plan shall reflect the existence of the center.
Communications with the Control Room shall be provided.
Licensee Commitments References:
(a) L. White, Jr.
(RG&E) letter to D. Ziemann (NRC), dated October 17, 1979.
(b) L. White, Jr.
(RG&E) letter to D. Ziemann (NRC), dated December 28, 1979.
The licensee representative agreed to establish on Onsite Opera-tional Support Center and to incorporate the center into the emergency plan.
Communications with the Control Room were *provid-ed by telephone and station intercom.
Ins ection Findin s Based on the inspector's observations at the Onsite Operational Support Center, the licensee appears to have satisfied the above requirement and associated commitment.
The inspector also review-ed the following documentatio Procedure SC 1.3E, Manning the Operational Support Center,
.
Revision 0, December 27, 1979.
Direct station intercom communication between the Onsite Opera-tional Support Center and the Control Room is not possible in the current configuration.
Communications with the Control Room using the intercom system requires relaying the message through the Tech-nical Support Center.
In this manner the licensee hopes to red'uce the number of different conversing parties with the Control Room.
Telephone communication is still provided.
2.1.6.a Inte rit of S stems Outside Containment Likel to Contain Radioactive Materials Re uirements References:
(a)
NUREG 0578 (b H. Denton (NRC) letter to All Operating Nuclear Power Plants, dated October 30, 1979.
(c)
D. Crutchfield (NRC) letter to L. White, Jr.
(RGSE), dated July 7, 1980.
Implement leak reduction measures for all systems that could
.
carry radioactive fluid outside of containment.
Measure actual leakage rates with system in operation and report them to the NRC.
Licensee Commitments References (a) L. White, Jr.
(RG&E) letter to D. Ziemann (NRC), dated October 17, 1979.
(b) L. White, Jr.
(RGSE) letter to D. Ziemann (NRC), dated November 19, 1979.
(c) L. White', Jr.
(RGSE) letter to D. Ziemann (NRC), dated December 28, 1979.
The licensee representative agreed to continue leak reduction meas-ures already established and to develop a preventive maintenance program which would include regularly scheduled maintenance on maintainable valves and annual external inspection for leakage and/
or deterioratio Ins ection Findin s Based on the inspector's review of the following documentation, the licensee appears to have satisfied the above requirement and associated commitments, Procedure PT-39; Primary Systems Leakage Evaluation Inservice Inspection, Kevision 0, August 26, 1980.
S. Spector (RGSE)
memorandum to G. Larizza (RGSE), dated June 6, 1980.
Subject Primary System Leakage Evaluation Inservice Inspection Discrepancies.
Systems identified in the licensee's October 17, 1979 submittal were tested in December, 1979 and leakage rates were reported to the NRC in the licensee's December 28, 1979 submittal.
2.1.8.b Re uirements Increased Ran e of Radiation Monitors References:
H. Denton (NRC) letter to All Operating Nuclear Power Plants, dated October 30, 1979.
An interim method for quantifying release rates of up to 10,000 Ci/sec for noble gases from all potential release points of the existing effluent instrumentation goes off scale.
Special procedures must be developed for the removal and analysis o'f the radioiodine/particulate sampling media under accident conditions.
Provide information describing the system/method to be used in quantifying noble gas, radioiodine and particulate.releases.
Procedures shall contain all aspects of the measurement analysis for noble gas, radioiodine and particulate effluent releases.
Licensee Commitments References:
L. White, Jr.
(RGSE) letter to D. Ziemann (NRC), dated December 28, 1979.
The licensee representative stated that procedures have been written
Ii r
l
27" or revised, as necessary, to provide for quantifying noble gas efflu-ents released from the plant vent if the existing instrumentation goes offscale.
Additionally, a Trapelo MAP 63, samples the plant vent ef-fluent iodine concentration.
The charcoal cartridges would be trans-ported from the sample location to the laboratory in a shielded con-tainer, purged of noble gases, and counted on a Germanium - Lithium system.
Ins ection Findin s Based on the inspector 's observation of the preselected monitoring locations, availability of the designated portable detectors in the Health Physics Office, and the review of the following documentation, the licensee appears to have implemented the above requirement and associated commitment.
References:
(a) Procedure PC-23.3, Estimation of Noble Gas Release Rate from the Plant Vent During Accident Conditions, Revision 2, May 7, 1980.
(b) Procedure PC-23.4, Radioactivity Release Thru Steam Vents, Revision 0, May 21, 1980.
(c) Procedure HP-11.2, Iodine in Air-Charcoal Cartridge Method.
Procedure PC-23.3 describes the location where the portable radiation monitor would be placed in relation to the plant vent during an acci-dent situation.
Review of the proximity of the monitor to the plant vent as well as to other possible sources of high radiation in an ac-cident environment, indicated that a reliable representation of the actual release rate would be questionable.
Further inspection effort on this matter will be addressed in IE Inspection. Report 80-16.
2.1.4 Re uirements References:
Containment Isolation (a)
NUREG 0578 (b H. Denton (NRC) letter to All Operating Nuclear Power Plants, dated October 30, 1979.
(c)
D. Crutchfield (NRC) letter to L. White, Jr.
(RGLE), dated July 7, 1980.
All containment isolation system designs will have diversity in the parameters sensed for the initiation of containment isolatio C
Identify each system as either essential or nonessential.
All nonessential systems shall be automatically isolated by the containment isolation signal.
Reopening of containment isolation valves following an iso-lation signal: shall require deliberate operator action.
Licensee Commitments References:
(a) L. White, Jr.
(RGSE) letter to D. Ziemann (NRC), dated October 17, 1979.
(b) L. White, Jr.
(RGSE) letter to D. Ziemann (NRC), dated November 19, 1979.
(c) L. White, Jr.
(RG8E) letter to D. Ziemann (NRC), dated December 28, 1979.
The licensee representative stated that there were certain valves which could reopen upon reset of the containment isolation or con-tainment ventilation isolation if the controllers were set in the open position.
Following an isolation signal procedural controls required the operator to place all valve position controllers in the closed position to prevent any valves from opening on initiation of the reset.
A modification to prevent the isolation valve from opening on reset without operator action was scheduled for imple-mentation during the spring 1980 refueling outage.
Ins ection Findin s
Based on the inspector's observations in the control room and review of the following documentation, the licensee appears to have satis-fied the above requirement and associated commitment.
Westinghouse drawing 882D612 sheet 6, Safeguard Actuation Signals, Revision 6, August 25, 1980.
RG8E responses to IE Bulletin 79-06A, Review of Operational Errors and System Misalignments Identified During the Three Mile Island Incident, dated April 28, 1979 and June 22, 1979.
Station Modification (SM)-2605, Diverse Containment Isolation, Design Criteria, Revision 1, March 28, 1980.
SM-2605, Diverse Containment Isolation Safety Analysis, Re-vision 0, February 1, 1980.
The licensee modified the containment isolation system reset logic to inhibit the reopening of a containment isolation valve following reset of the isolation signal until deliberate operator action is
C r I
taken.
Each isolation valve now has a pushbutton which when acti-vated provides the logic to allow reopening the valves as necessary.
2..1.8.c Im roved In-Plant Iodine Instrumentation Un er Aces ent Conditions
'e uirements References:
(a)
NUREG 0578 (b H. Denton (NRC) letter to All Operating Nuclear Power Plants, dated October 30,'979.
(c)
D. Crutchfield (NRC) letter to L. White, Jr.
(RGSE), dated 'July 7, 1980.
. Each licensee shall have the capability to accurately deter-mine the airborne iodine concentration with portable monitor-ing equipment in areas within the facility where plant personnel may be present during an accident.
Licensee Commitments References:
(a) L. White, Jr.
(RGEE) letter to D. Ziemann (NRC), dated October 17, 1979.
(b) L. White, Jr.
(RGSE) letter to D. Ziemann (NRC), dated November 19, 1979.
The licensee representative stated that mobile air monitors having a
single channel analyser calibrated to the I-131 energy are located in various areas throughout the plant to detect the presence of iodine.
Further, portable air samplers with charcoal and silver zeolite cart-ridges were said to be available in the Health Physics Office and at the Emergency Survey Center.
Ins ection Findin s
Based on the inspector's observations of mobile instrumentation and portable air sampler (including cartridge) locations, and review of the following doucmentation, the licensee appears to have satisfied the above requirement and associated commitment.
Procedure SC-1.7B, Determination of Iodine or Particulate, Revision 7, April 22, 1980.
Site Radiation Emergency (Shift Supervisor and Control Room),
Revision 12, May 10, 1980.
Manning the Technical Support Center, Revision 3, April 14, 198 ~
t C
Licensee Event Re orts LER's)
The inspector reviewed the following LER's to verify that the details of the event were clearly reported, including the accuracy of the description of cause and adequacy of corrective action.
The inspector determined whether further information was required, and whether generic implications were involved, The inspector also verified that the reporting requirements of Technical Specifications and Station Administrative and Operating Procedures had been met, that appropriate corrective action had been taken, that the event was reviewed by the Plant Operations Review Committee, and that the continued operation of the facility was conducted within the Technical Specification limits.
80-07:
RPS Channel EI Flux Tilt Controller (TC-402R) inoperable-Xugust 29, 1980 -'Licensee investigation deternined that a faulty main amplifier bridge assembly and low limit transistor attributed to the malfunction.
Three redundant channels remained operable.
The inspector verified the effected controller was replaced with a quali-fied space.
Post installation calibration was performed August 29, in accordance with calibration Procedure 20.0, Delta Flux Controllers, Revision 6, September 15, 1978.
80-08:
'B'iesel Generator (DG) Breaker tie-in to Bus 16 would not
~c ose - Septenber 10, 1980.
Licensee investigation deternined that the closing relay anti-pump release lever operated i'ntermittently due to overtightening of the guide pin.
The A DG was run continously as required by TS 3.7.2.b.
The inspector verified that the closing relay was replaced with a qualified space.
Post installation functional test-ing was performed through repeated exercising of the breaker.-
No items of noncompliance were identified.
Review of Periodic and S ecial Re orts Upon receipt, periodic and special reports submitted by the licensee pursuant to Technical Specification 6.9.1 and 6.9.3 were reviewed by the inspector.
This review included the following considerations:
the report includes the information required to be reported by NRC requirements; test results and/or supporting information are con-sistent with design predictions and performance specifications; planned corrective action is adequate for resolution of identified problems; determination whether any information in the report should be classified as an abnormal occurrence; and the validity of reported information.
Within the scope of the above, the following periodic reports were reviewed by the inspecto I I
~I
Monthly Operating Report for October, 1980.
No items of noncompliance were identified.
Exit Interview
-'t periodic intervals during the course of the inspection, meetings were held with senior facility management to discuss the inspection scope and finding ~
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