ML19323A541
| ML19323A541 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 04/01/1980 |
| From: | NRC COMMISSION (OCM) |
| To: | |
| Shared Package | |
| ML19323A539 | List: |
| References | |
| NUDOCS 8004210317 | |
| Download: ML19323A541 (38) | |
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i ACTION PLAN ITEMS REVIEWED BY JOINT NRR/IE TEAM Shift Manning Organization and Management Criteria Safety Engineering Group Resident inspector Licensee Dissemination of Operating Experience Applicant /NRC Communications STATUS Review Team Visited VEPCO February 20-22 1
i Final Staff Position is Presented in SER Supplement No.10 4
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NTOL REO.UIREMENTS FOR FUEL LOAD AND LOW POWER TESTING i
4 l.A.1.1 Shift Technical Advisor I.A.1.2 Shift Supervisor Duties l.A.1.3 Shift Manning I. A.3.1 Revised Scope and Criteria for Licensing Examinations I.B.1.1 Organization and Management Criteria I
l.B.1.2 Safety Engineering Group and Onsite Evaluation Capability 1
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f.B.1.4 Licensee Onsite Operating Experience Evaluation Capability I.B.2.2 Resident NRC inspector I.C Procedures I.C.1 Short-Term Effort I.C.2 Shift and Relief Turnover Procedures I.C.3 Shift Personnel Responsibilities I.C.4 Control Room Access I.C.5 Licensee Dissemination of Operating Experiences I.C.7 NSSS Vendor Review of Low Power Test Procedures I.G Training During Low Power Testing II.B.4 Degraded Core-Training
NTOL REO.UIREMENTS FOR FUEL LOAD AND LOW POWER TESTING (CONT'D)
II.D.2 Relief and Safety Valve Test ll.D.5 Relief and Safety Valve Position -
li.E.1.2 Auxiliary Feedwater initiation and Indication ll.E.4.1 Containment Penetrations ll.F.1 Additional Accident Monitoring instrumentation ll.F.2 Inadequate Cora Cooling II.G Emergency Power for Pressurizer Equipment ll.K.1 IE Bulletins on Measures to Mitigate Small Break Loca's and Loss of Feedwater ll.K.3 Generic Review Matter. - Small Break Loca's and Loss of Feedwater Transients Ill.A.1.2(a)
Technical Support Center Ill.A.1.2(b)
Onsite Operational Support Center lil.A.3 Improving NRC Emergency Preparedness Ill.A.3.3 Communications lil.B Emergency Preparedness of State and Local Governments lil.B.1 Near Term Actions Ill.D.3.3 in-Plant Radiation Monitoring IV Recommendations of NRC Special Inquiry Group ITEM 1 Control Room Design Review ITEM 2 Power Ascension Test Schedule
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f O.UALIFICATIONS OF REACTOR OPERATORS SECY 79-330E l
NOT APPLICABLE TO NORTH ANNA COLD APPLICANTS REQUIREMENTS IMPLEMENTATION NORTH ANNA
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- 1. Hold Operator's License ASAP 7 of 10 Meet l
for One Year Prior to Requirements l
Senior
- 2. Simulator Training for Expeditiously Planned Hot Applicants
- 3. NRC Audit Training Planned
- 4. Retraining on Simulator ASAP Planned
- 5. NRC Requalification Examinations Planned
- 6. Revise ANS 3.5 Expeditiously In Process (Simulator Standard)
- 7. NRC Examiner Selection and ASAP in Process Training Study i
QUALIFICATIONS OF REACTOR OPERATORS SECY 79-330E
' APPLICABLE TO NORTH ANNA UNIT 2 REQUIREMENT IMPLEMENTATION NORTH ANNA
- 1. Previous Experience for ASAP Meet Requirements Senior Operators
- 2. Shift Training Expeditious Meet Requirements Partial
- 3. Instructor Qualifications Expeditious All Instructors Have, or Had Senior License
- 4. Specific Exercises on Expeditious Meet Requiremertas Simulator
- 5. Increase Scope of ASAP Meet Requirements Partial Examinations l
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- 6. Simulator Examination ASAP Do Not Meet l
Requirements
- 7. Passing Grade Increased ASAP Meet Requirements
- 8. Inform Management of ASAP Meet Requirements l
l Examination Results
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i ISSUES l
1 o Turbine Inspection e Underclad Cracking l
l e Crystal River Syndrome i
PRINCIPAL ELEMENTS OF AN OPERATING LICENSE Contains Findings in Accordance With 10 CFR 50.57 Relative to:
l e Construction e Operation e Public Health & Safety e Common Defense & Security e Technical & Financial Qualifications e Balancing Benefits Against Environmental & Other Costs License Authorizes e Possession, Use & Operation of Facility e Possession, Use & Storage of Spocial Nuclear (Byproduct
& Source) Materials License Conditions e Maximum Power Level e Technical Specifications e Special Test Program ( < 5% Power)
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jo, UNITED STATES J
y,g NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 jj k*v,/
March 26, 1980 M
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. Chairman dhearne/,-
MEMORANDUM FOR:
Comissioner Gilinsky Comissioner Kennedy Comissioner Hendrie Comissioner Bradford FROM:
Harold R. Denton, Director i
Office of Nuclear Reactor Regulation THRU:
William J. Dircks QQp I
Y Acting Executive Director for Operations
SUBJECT:
CONSIDERATION OF OPERATING LICENSE FOR NORTH ANNA P0h'ER STATION, UNIT NO. 2 The Staff has nearly completed its review of North Anna Power Station, Unit j
No. 2.
Our review, which is documented in the enclosed draft Supplemental Safety Evaluation Report (SSER), includes disposition of the residual pre-TMI issues (some of which were discussed in our SER, NUREG 0053 of June 1976)
'and the list of requirements for Near Term Operating Licenses (NTOL) which was based oh consideration of the actions contained in the NRC TMI Action Plan.-
1 At this time, there is one remaining outstanding issue related to issuance of a 5'.' full power license to the North Anna facility. This issue concerns the design of the control room from a human factors standpoint. The NRC Staff, ftogether with its consultants from the Essex Corporation conducted a 5 day preliminary c6-site review of the North Anna Unit 2 control room on March 17-21.
Staff members participating in the review represented the NRR Instrumentation and Control, Operator Licensing, Reactor Systems, and Analysis Branches. The objective of thi's review was to assess the degree to which the current design reflects specific human factors considerations. The review included the assessment of control and display panel layout, annunicator design, labeling and sequencing of panel components, and the usability and completeness of selected emergency procedures.
The review was perfomed by means of detailed inspection of the control panels, interviews with operators, and observation and videotaping of operators as they walked through selected emergency procedures.
Our review identified human factor deficiencies in the control room in the areas of noise level, control panel physical layout, meter displays, i
annunicators, process controllers, lamp test capability and emergency procedures.
The staff is currently in the process of categorizing these deficiencies into two groups as follows: Category 1, deficiencies which can and should be corrected imediately, and Category 2, deficiencies which 1%xs Soo yt 8' O t 54 v
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The Ccmmission are of concern, but may be corrected at a later date. Corrections of Category 1 deficiencies will be required before license issuance. Corrections of I
deficiencies in the second category will require significant control room modifications and will be impleme'ted at a specified later date after both the applicant and the Staff have carefully assessed the impact of the corrections on the operators.
i Both Category 1 and 2 will'txt discussed with the applicant on Friday of this week.
Specific changes to resolve the Categcry 1 deficiencies will be l
identified at that time.
We have concluded that, except for the issue concerning control rcom design, all necessary requirements for permitting operations up to 5% power have been satisfied. We expect to complete our review of the control room design by Friday, March 28, to the extent that the short term remedial measures needed to be implemented prior to initial criticality will have been specifically defined.
The SSER will be revised to reflect the control room matter and the revisions provided to the Commission on Monday, March 31. We will be prepared to meet
.with the Commission on the operating license for North Anna Station, Unit 2 at that time.
/76tS$
ev Harold R. Denton, Director Office of Nuclear Reactor Regulation
Enclosure:
As stated cc w/ enc 1.
ENCLOSURE 1 NORTH ANNA POWER STATION, UNIT NO. 2 CONTROL ROOM DESIGN REVIEW '
As a part of the staff actions following the TMI-2 accident, we will require
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that all licensees and applicants for operating licenses conduct a detailed control room design review. We expect those reviews to be initiated within the next several months and require over a year to complete. As an interim I
measure, the NRC staff together with consultants from the Essex Corporation, conducted a 5 day preliminary on-site review of the North Anna Unit 2 control room.
NRC staff members participating in the review respresented the Instrumentation and Control, Operator Licensing, Reactor Systems, and j
Analysis Branches. The objective of this review was to assess, prior to
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licensing, the degree to which the design reflects specific human factors considerations. The review included the assessment of control and display panel layout, annunciator design, labeling and sequencing of panel components, and the usability and completeness of selected emergency procedures. The review was perfomed by means of detailed inspection of the control panels, interviews with operators, and observation and videotaping of operators as they walked through selected emergency procedures.
Our review identified several human factor deficiencies. The staff is currently in the process of categorizing these deficiencies into two groups as follows:
1) deficiencies which are to be corrected prior to criticality.
2) ceficiencies which are of concern, but may be corrected at a later date.
i
F It is anticipated that once categorized, corrections of those deficiencies in the second category will require significant control room modifications. They will be implemented only after both the applicant and the staff have Assessed' their impact on the operators. The acceptability of licensing North Anna Unit 2 prior to correcting these deficiencies is based on the extensive training the operators have received and their familiarity with the control room as presently equipped.
Deficiencies identified During the Control Room Review Although not a complete list at this time, some of the human factors related control room' deficiencies identified by the review team are:
1.
Control Room Noise Level The control room is fairly compact which, by itself, should enhance connunica-tions. However, the background noise is such that normal conversational volume is only effective within a short distance. A ventilation duct run-1 l
ning up one end of the Unit 2 control room supplies a continuous high noise i
background. To this is added periodic announcements over the public address system, the speaker for which is mounted in the ceiling directly over z
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the control censole. We requested that an operator don a Scott Air-Pak and i
attempt to communicate with others nearby. Given the background noise in the control room and the inherent design of the facemask, it proved a near impossibility for the operator to connunicate.
2.
Control Panel Physical Layout The control panels are confusing to the operator. Meters are laid out in strings and clusters such that identifying a particular meter is difficient.
In some instances, the indication on the vertical panel was in a different l
location from that which would correspond to the control on the control 1
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c:nsole. This prevents an easy sweep upward of the eyes from control to indicator. Yepco has made an attempt to demarcate groups of associated 3.
meters to aid the operator in locating a specific meter. However. because of the basic layout of the displays, these 1ines of denari:stion take on fairly abstract shapes.
3.
Meter Displays Extensive use of vertical meters is made, each of which has a vertical label on the meter face. Vertict1 labels are difficult to read, at best, and become more difficult when the meter of interest is one of many identical meters butted against each other in a long string.
4.
Annunciators Except for "first out" capability, there is no indication of the significance of a given annunciator.
In addition, an alarm can clear and the annunciator return to its normal state without the operator being aware of it.
5.
Process Controllers The Hagan controllers used extensively provide a readout of the demand signal only.
In most cases the operator can detennine if the control manipulation was effective by monitoring the controlled parameter.
In other cases, the operator must wait for a delayed response of the system to the variatiac in the controlled parameter.
In these latter cases, it may be prudent to provide direct indication of the controlled parameter.
t 6.
Lamp Test Capability l
Except for the annunciators, there is no capability to test the control panel lamps. Hence, there is always room for doubt when an indicating light remains unlit.
Is the bulb burnt out or did the equipment fail to activate?
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7.
Administrative Several examples of poor administrative controls exist in the control room.
Labels are peeling off of some indicators and controls, recorder pen: are filled with a different color ink from the color of tbt pen and scales on 1
some recorder chartpaper do not correspond to the scale on the recorder it-
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self.
There does not appear to be good control over control room access. While n
j the review team was present, we observed as many as fifteen others present in the comon control room.
8.
Emergency Procedures It does not appear that the emergency procedures have ever been system-atically stepped through on the control boards.
In.:me cases steps are missing, in other instances important steps and warnings are buried in the notes.
Some of the instructions are ambiguous and others call for a precision beyond the capability of the instrumentation.
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- 9..' Miscellaneous Deficiencies In addition to those items listed abcve, there are other human factors related deficiencies scattered throughout the control room. For example, it the color red had multiple meanings, depending upon the area of the control boards in which it was used. There were violations of convention noted; for example a switch marked " lower-raise" (convention) imediately above a switch marked " raise - lower" (violation of convention). Several "j" handle switches need to have guard bars installed to prevent inadvertent operation. The colors chosen for some of the labels (black writing on dark purple background) make it nearly impossible to read.
These deficiencies and others identified in the final Essex report will i
be discussed with the applicant in the near future. Categories will be assigi.ed and corrective action taken depending on the categories.
We will make follow up visits to the site to ensure that the applicant f
has corrected the deficiencies.
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UNITED STATES g
NUCLEAR REGULATORY COMMISSION o
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VIRGIN!A E!.ECTRIC AND POWEW COMPANY DOCXET NO. 50-339 i
NORTH ANilA POWER STATION, UNIT NO. 2 LICENSE FOR FUEL LOADING AND LOW POWER TESTING i
License No. NPF-7 1.
The Nuclear Regulatory Commission (the Comission) having found that:
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J A.
The application for license filed by the Virginia Electric and Power Company (the licensee) complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's regulations set forth in 10 CFR Chapter I and all i
required notifications to other agencies or bodies have been duly i
made;
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B.
Construction of the North Anna Power Station, Unit No. 2 (facility) has been substantially completed in conformity with Construction Permit No. CPPR-78 and the application, as amended, the provisions j
i of the Act and the regulations of the Comission.
C.
The facility requires an exemption from certain requirements of l
Appendix J to 10 CFR Part 50. This exemption is described in the Office of Nuclear Reactor Regulation's Safety Evaluation Report, S"pplement No. 10. This exemption is authorized by law and will net endanger life or property or the comon defense and security and is otherwise in the public interest.
The exemption is, therefore, hereby granted. With the granting of this exemption, the facility will operate, to the extent authorized herein, in conformity with the 7
application, as amended, the provisions of the Act, and the rules and i
regulations of the Comission.
D.
The facility will operate in conformity with the application,.as i
amended, the provisions of the Act, and the regulations of the Comission; E.
There is reasonable assurance: (1) that the activities authorized by this license can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in i
compliance with the regulations of the Comission; i
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Virginia Electric and Power Company is technically and financially F.
i qualified to engage in the activities authorized by this license in accordance with the regulations of the Commission; The Virginia Electric and Power Company has satisfied the applicable G.
provisions of 10 CFR Part 140, " Financial Protection Requirements and Indemnity Agreements", of the Commission's regulations; i
The issuance of this license will not be inimical to the common-H.
defense and security ur tu one nealth and safety of tne public; I.
After weighing the environmental, economic, technical, and other 9
benefits of the facility against environmental and other costs and considering available alternatives, the issuance of License No. NPF-7 I
subject to a condition for protection of the environment set forth,
herein is in accordance with 10 CFR Part 51, of the Commission's regulations and all applicable requirements have been satisfied; and J.
The receipt, possession, and use of source, byproduct and special nuclear material as authorized by this license will be in accordance with the Commission's regulations in 10 CFR Part 30, 40, and 70, I
including 10 CFR Sections 30.33, 40.32, 70.23 and 70.31.
2 Pursuant to approval by the Nuclear Regulatory Commission at a meeting i
on April
,1980 License No. NPF-7 is hereby issued to the Virginia Electric and Power Company to read as follows:
A.
This license applies to the North Anna Power Station, Unit No. 2, a pressurized water nuclear reactor and associated eoutpment (the facility),
owned by the Virginia Electric and Power Company (VEPCO). The facility is located near Mineral in Louisa County, Virginia and is described in 4.f.':, '.
the " Final Safety Analysis Report" as supplemented and amended (Amendments l
17 through 68) and the Environmental Report ds supplemented and amended U
!"Y.o (Supplements 1 through 4).
r Subject to the conditions and requirements incorporated herein', the B.
Commission hereby licenses VEPC0-(1) Pursuant to Section 103 of the Act and 10 CFR Part 50, to possess, use, and operate the facility at the designated location in Louisa County, Virginia, in accordance with the limitations set forth in this license; 9
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(2) Pursuar,t to the Act and 10 CFR Part 70, to receive, possess and use at any time special nuclear material as raactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Y
l Safety Ana'ysis Report, as supplemented and amended
- I (3) Pursuant to the Act and 10 CFR Parts 30, 40 and 70 to receive, possess and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation
- monitoring equipment calibration, and as fission detectors in j
amounts as required; i
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(4) Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, i
possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical j
form, for sample analysis or instrument calibration or associated, with radioactive apparatus or components; i
(5) Pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials.as may be produced by the operation of the facility.
j C.
VEPCO is authorized to perform steam generator moisture carryover studies at the North Anna Power Station. These studies involve the j
use of an aqueous tracer solution of two (2) curies of sodium-24.
VEPCO's personnel will be in charge of conducting these studies and be knowledgeable in the procedures. VEPCO will impose personnel exposure limits, posting, and survey requirements in conformance with those in 10 CFR Part 20 to minimize personnel exposure and contamination during the studies. Radiological controls will be established in the i,,
areas of the chemical feed, feedwater. steam, condensate and :agling systems where the presence of the radioactive. tracer is expected to
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warrant such controls. VEPCO will take special precautions to minimize j6. '. '
radiation exposure and contamination during both the handling of the radioactive tracer prior to injection and the taking of. system samples following injection of the tracer. VEPCO will insure that all regulatory requirements-for liquid discharge are met during disposal of all sampling effluents and when reestablishing continuous blowdown from the steam generators after completion of the studies.
D.
This license shall be deemed to contain and is subject to the conditions specified in the following Comnission regulations in 10 CFR Chapter 1:
Part 20, Section 30.34 of Part 30. Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; and is subject to all j,
applicable provisions of the Act and to the rules, regulations, and orders of the Connission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
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l 4-(1) Maximum Power Level VEPCO is authorized to (c) load fuel, (b) proceed to initial criticality (c) perform startup testing at zero power in Operational Mode 2, and (d) after prior written approval by the Director of Nuclear Reactor Regulation, operate the facility for testing at reactor core power levels not in excess of 139 megawatts thermal (five percent of rated power).
(2). Technical Specifications The Technical Specifications contained in Appendices A and B attached hereto are hereby incorporated into this license.
VEPCO shall operate the facility in accordance with the Technical i
Specifications.
(3) Secondary Water Chemistry
-EPC0 shall perform secondary water chemistry program to inhibit
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steam generator tube degradation. This program shall include:
a.
Identification of a sampling schedule for the critical parameters and control points for these parameters; b.
Identification of the procedures used to quantify parameters that are critical to control points; L
c.
Identification of process sampling points; d.
Procedure for the recording and management of da'ta;
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procedures defining corrective actions for off control point chemistry conditions; and j
f.
A procedure identifying the authority responsible for the data and the sequence and timing of administrative events required to initiate corrective, action.
g.
Monitor the condensate at the effluent on the condensate pump. When condenser leakage is confirmed repair or plug the leak in accordance with Branch Technical Position MTES 5-3.
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(4)
Initial Test Program VEPCO shall conduct the post-fuel-loading initial test ~ program that has been reviewed and approved by the Commission at the time of issuance of this license without making any major J
I modification of this program. Major modifications are deemed to involve unreviewed safety questions under 10 CFR Section 50.59 and are defined as:
Elimination of any test identified in Section 14 of the Final a.
Safety Analysis Report as essential.
b.
Modification of test objectives, methods or acceptance criteria i
e for any test identified in Section 14 of the Final Safety i
Analysis Report as essential.
c.
Performance of any test at a power level different from there described.
d.
Failure to complete any tests included in the described program (planned or scheduled for power levels up to the authorized i
i power level).
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(5) Fuel Load and Zero Power (i.e., power recuired to perform physics i
testing) Testing Conditions l
s The following conditions shall be completed to the satisfaction of i
the Commission prior to fuel loading. The following conditions are related to matters specified in the TMI Action Plan, Near Term Operating License (NTOL) Requirements, dated February 6,1980, and i
applicable to fuel load and zero power testing. Each of the following conditions references the appropriate section of Part II 1s of Supplement No.10 to the Safety Evaluation Report (NUREG-0053 i
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Suppl. No.10) for the North Anna Power Station, Units I and 2 and r
follows the numbering sequence utilized in the February 6,1980
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NTOL Requirements list.
a.
Shift Technical Advisor (I. A.l.1)
VEPCO shall provide on each shift a Shift Technical Advisor (STA) whose principal duty shall be to act as an advisor to r
the shift supervisor primarily in assessing accident / transient occurrences.
During 1980, at least one Senior Reactor Operator (SRO) or an experienced engineer who is a member of the site Safety.
s Engineering Staff shall be designated as the Shift Technical Advisor.
. The STA'shall interface with the Safety Engineering Staff whose j
function is to perform operating experience assessments, for I
the purpose of disseminating operating experience to the other members of the plant staff.
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At no time shall the SRO/STA, who is Lnder the for.:tional supervision of the Shift Supervisor, assume command or control 1
function nor function as the Shift Supervisor or Assistant Supervisor.
b.
Shift Supervisor Duties (I.A.1.2) and Shift Personnel Responsib111 ties (.I.C.3)
VEPCO shall issue a management directive, signed by the highest level of corporate management, which emphasizes the assignment of primary management responsibility to the Shift Supervisor.
l This directive shall be reissued on an annual basis.
The responsibility of the Shift Supervisor shall be that defined in VEPCO Administrative Procedure ADM-1.0, " Station j
Organization and Responsibility."
The shift supervisor shall remain in the control room at all I
times when either unit is operating in Modes 1, 2, 3 or 4
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except that he may be allowed to be absent provided an individual, other than the shift technical advisor, who possesses a valid SR0 license assumes the control room command function during his absence. The individual who assumes the control room command function must remain in the control room until the shift supervisor returns and reassumes the command j
function.
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The shift supervisor shall only be relieved by an individual who I
possesses a valid SRO license. Individuals who do not-possess a valid SRO license, including members of station management, shall not relieve the shift supervisor, nor shall they direct the licensed activities of licensed operators.
c.
Shift Mannino (I.A.1.3) l l
The shift crew composition for the operation of North Anna, i
Unit 2, in conjunction with Unit 1 shall be in accordance with l
l Table 6.2.1 of the Technical Specifications. An additional requirement is that one senior licensed operator, licensed on North Anna, Unit 2 shall be stationed in the Unit 2 control
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room at all times during Unit 2 operation in Modes 1 through 4.
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7 jv VEPCO shall have adninistrative procedures to assure that qualified individuals to man the operational shifts are readily available in the event of an abnormal or emergency situation.
These administrative procedures shall include provisions which limit the amount of overtime worked by licensed operators.
The need for a licensed operator to exceed the limits on over-time shall be inf requent.
The limits on overtime work hours are:
(1) An individual should not be permitted to werk icre than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> straight (not inciading shift turnover time).
(2) There should be at least a 12-hour break between all work periods (shift turnover time is included in this 12-hour break).
l (3) An individual should not work more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in I
any 7. day periods (4) An individual should not work more than 14 consecutive -
days without having 2 consecutive days off.
However, for those circumstances which arise requiring i
deviation from the above, such deviation may be authorized by the plant manager or high levels of the VEPCO management in accordance with established procedures and with appropriate documentation of the cause.
J d.
Revised Scope and Criteria for Licensing Examinations (I. A.3.1) l VEPCO shall train all shift personnel (licensed and non licensed i
at North Anna Power Station, Units 1 and 2 in all the TMI related design changes which have been incorporated in the station. This training shall take place on shift and be conducted by the Shift Technical Advisors.
e.
Organization and Manacement Criteria (I.B.1.1)
With respect to offsite technical support to the plant staff in the event of an emergency, VEPCO shall develop procedures delineating the responsibility and authority of corporate office personnel in providing technical support.
The shift supervisor for Unit 2 shall report to an individual holding a senior operator license on Unit 2 -(as required by Technical Specification 6.2) at all times during Unit 2 operation in Modes 1 through 4.
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Safety Engineering Group and Onsite Evaluation Cacability, and Dissemination of Operatino Experience 1.8.1.2 (I.8.1.4 and I.C.5)
VEPCO shall establish an onsite Safety Engineering Group composed of a minimum of five engineering personnel (group shall be multidisciplined) who are not part of the plant staff and who shall be present on each day shift to act as independent observers. The duties and responsibilities of this group shall be as follows 1.
Se cognizant of the scope and intent of the test program h
2.
Be cognizant of the tests being conducted 3.
Be f amiliar with the operation of a pressurized water type rcactor 4.
Provide daily status reports to the Director. Safety i
Engineering at the corporate office.
VEPCO shall establish an onsite capability to evaluate operating history of North Anna Unit-2, in conjunction with that of North l
Anna Unit 1 and also plants of similar design.
j Shift Relief and Turnover Procedures (I.C.2)_
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i VEPCO shall implement the shift and relief turnover procedures j
in accordance with VEPCO Administrative Procedure ADM-29.3--
" Conduct of Operations."
VEPCO shall establish a system to evaluate the effectiveness g., '-
of the turnover procedures.
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h.
Control Room Access (I.C.4)
VEPCO shall implement control room access procedures in i
accordance with ADM-6.0. " Control Room Access."
1.
Degraded Core - Training (II.B.4)
VEPCO shall prepare a program to ensure that all operating i
personnel are trained in the use of installed plant systems j
to control or mitigate an accident in which the core is severely damaged. The training program shall include the i
following topics
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9 Incore Instrume1tation 1.
Use of fixed or movable incore detectors :o determine extent of core damage and geometry changen i
2.
Use of thermocouples in determining peak temperatures; l
methods for extended range readingst methods for direct readings at terminal junctions.
Excore Nuclear Instrumentation (NIS) f 1.
Use of NIS for determination of void formation; void i
location basis for NIS response as a function of core i
temperatures and density changes.
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Vital Instrumentation 1.
Instrumentation response in an accident environment; failure sequence (time to failure, method of failure); indication reliability (actual vs. indicated level).
2.
Alternative methods for measuring flows, pressures, levels.
I and temperatures.
i a.
Determination of pressurizer level if all level transmitters fail.
b.
Determination of letdown flow with a clogged i
filter (l'ow flow).
s c. -
ch Determination of other Reactor Coolant System
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parameters if the primary method of maasurement has failed.
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g Primary Chemistry 1.
Expected chemistry results with severe core damage-consequences of transferring small quantities of liquid j
outside containment; importance of using leak tight systems.
2.
Expected isotopic breakdown for core damage; for clad damage.
p 3.
Corrosion effects of extended inanersion in primary water; time to failure.
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t 1 Radiation Monitoring 1.
Response of Process and Area Monitors to severe damages-behavior of detectors when saturated; method for detecting radiation readings by direct measurenent at detector output (overanged detec ir); expected accuracy of detectors at i
different locatic 5; use of detectors to determine extent of core damage.
i 2.
Methods of determining dose rate inside containme'nt from l
measurements taken outside containment.
J Gas Generation 1.
Methods of H2 generation during an accident; other sources of gas (Xe, Kr); techniques for venting or disposal of
- noncondensibles.
1 2.
H2 flammability and explosive limit; sources of 02 in containment or Reactor Coolant System l
- j. Relief and Safety Valve Test (II.D.2)
VEPCO shall carry out a testing program to qualify the relief
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and safety valves under expected operating conditions for design basis transients and accidents in accordance with the requirements and schedule provided in NUREG-0578, Section 2.1.2, as clarified in NRC letter to operating license applicants.
l dated November 9, 1979.
k.
Relief and Safety Valve Position (II.D.5)
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VEPC0 shall provide reactor system relief and safety valves with a positive indication in the control room derived from a reliable valve position detection device or a reliable indication of flow in the discharge pipe. The indication shall comply with the requirements contained'in NUREG10578 Section 2.1.3.a. as clarified in NRC letter to operating license opplicants dated November 9,1979.
1.
Auxiliary Feedwater Initiation and' Indication (!!.E.1.2) i VEPCO shall provide for automatic initiation of the auxiliary feedwater system and shall provide for indication. in the control room, of auxiliary feedwater flow to.each steam generator. These requirements shall comply with NUREG-0578, i
Sections 2.1.7.a and 2.1.7.b. as clarified in NRC letter to operating license applicants dated November 9,1979.
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Inadeouate Core Cooling - Subcooling Meter (ll.F.2) m.
VEPCC shall provide a subcooling meter to provide on-line indication of coolant saturation condition. The meter shall comply with the requirements of NUREG-0578. Section 2.1.3.b, l
as clarified in NRC letter to operating license applicants dated November 9,1979.
Inadeouate Core Cooling - Additional Instrumentation (II.F.2) i n.
VEPCO shall provide a design of additional instruments to provide an unambiguous indication of inadequate core cooling.
This requirement shall comply with NUREG-0578, Section 2.1.3.b, as clarified in NRC letter to operating license applicants I
dated Noveaber 9,1979.
Emergency Power for Pressurizer Equipment (ll.G) o.
VEPC0 shall provide emergency power for the power-operated relief valves (PORVs), the PORV block valves and pressurizer level instrurent channels. This requirement shall comply with NUREG-0578, Section 2.1.1, sub-section 3.2, as clarified in NRC letter to operating license applicants dated November 9,1979.
IE Bulletins on Measures to' Mitigate Small Break LOCAs and p.
Loss of Feedwater Accidents (ll.K.1)
VEPCO shall review the operating procedures and training instructions requested by Item 7(a) of IE Bulletin 79-06A.
This review should ensure that operators have been instructed not to override automatic operations of the engineered safety features, unless their continued operation would result in unsafe plant conditions, or until the plant is clearly in a stable, controlled state, and engineered safeguards are no longer required.
q.
Improve Licensee Facilities for Resconding to Emergencies -
Onsite Technical Support Center (III. A.I.2)a VEPCO shall establish a technical support center to meet the January 1,1980 requirements of NUREG-0578, Section 2.2.2.b.
as clarified by NRC letter to operating license applicants dated November 9, 1979.
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I Improve Licensee Facilities for Resoonding to Emergencies -
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Onsite Operational Suoport Center (llI. A.l.2)O VEPC0 shall provide an onsite operational support center to I
meet the requirements of NUREG-0578, Section 2.2.2.c, as clarified by NRC letter to operating license applicants dated November 9,1979.
Emergency Preparedness of State and Local Governments -
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Near-Term Actions (III.B and III.B.1)
During the period of this license. VEPCO'shall maintain in f
effect an emergency plan that meets:
(1) Regulatory requirements of 10 CFR Part 50, Appendix E (2) Regulatory Position Statement in Regulatory Guide 1.101 (March 1977) i (3) The Essential Planning Elements in NUREG-75/lli and j
Supplement 1 thereto defined by NRR as significant for ~
fuel load and low power testing.
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In-plant Radiation Monitoring - Partial (III.O.3.3)
VEPCO shall provide equipment and associated training and procedures for accurately determining the airborne iodine concentration throughout the plant under accident conditions, Communications (III.A.3.3) u.
VEPCO:shall install and maintain direct dedicated telephone lines between the North Anna plant control room, Site Technical Support Center and NRC Incident Response Center in Bethesda, Maryland.
v.
Prior to operation in Mode 2 VEPCO shall implement the-following corrective actions related to the control room:
(1) Significantly improve labeling and instrumentation and control demarcation (2) Color code annunicator w% dows as an aid in identifying high priority alarms (3) Review, and improve where necessary, administrative controls used to casure proper safety injection status l
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(4) Correct problems associated with general maintenance and implement measures to prevent their recurrence l
t (5)
Institute controls to limit control room access i
(6) Test the adequacy of emergency lighting.
i (6) Conditions on Operations Beyond Zero Power Testing The following conditions shall be completed to the satisfaction of the Commission prior to proceeding beyond zero power testing.
The following conditions are related to matters specified in the i
TMI Action Plan, Near Term Operating License (NTOL) Requirements dated February 6,1980, and applicable to operation beyond zero i
power testing. Each of the following conditions references the s
appropriate section of Part II of Supplement No.10 to the Safety Evaluation Report (N" REG-0053 Suppl. No.10) for the North Anna Power Station, Units 1 and 2 and follows the list.
The below designated sections of Part II of Supplement No.10 to the Safety Evaluation Report (NUREG-0053 Suppl. No.10) are i
hereby incorporated by reference.
In case of any inconsistency between the license and Supplement No.10 to the Safety
)
Evaluation Report, the terms of the license shall govern.
l Short-Term Effort-Analysis and procedure Modifications (I.C.1)'
a.
i VEPCO shall revise its emergency operating. instructions I
for dealing with small break LOCAs and inadequate core cooling based on its analysis of these events and the vendor guidelines derived from these analyses.
b.
NSSS Vendor Review of low Power Test Procedures (I.C.7)
VEPCO's low power test procedures shall be reviewed by the nuclear steam supply system vendor, Westinghouse, and documentation of the review submitted to NRC.
c.
Low Power Test Program (I.G.1)
As set forth in Section I.G.1, VEPC0 shall obtain staff approval of a low power test program.
d.
Additional' Accident Monitoring Instrumentation (II.F.1)
VEPCO shall comply with the requirements contained in' NUREG-0578, Section 2.1.8.b as clarified in NRC letter to operating license applicants dated November 9,1979.
_ E.
VEPCO shall maintain and fully implement the physical security plan 6
entitled, " Security Program, North Anna Power Station, Units 1 and 2,"
dated May 25, 1977, as amended, November 30, 1977, and revised September 25, 1978, Oct ober. 25, 1978 J anuary 12,1979, March 23, 1979, April 30,1979, and December 1,1979, and as amended in accordance with the provisions of 10 CFR 50.54(p).
Pursuant. to 10 CFR Section 2.790(d), the security plan is being withheld from public disclosure because it is deemed to be commercial or financial information within the meaning of 10 CFR Section 9.5(a)(4) and subject to disclosure only in accordance with 10 CFR Section 9.12.
F.
This license is subject to the following additional condition for the protection of the environment:
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(1) Before engaging in additional construction or operational activities which may result in an environmental impact that was not evaluated by the Commission, the licensee will prepare and record an environ-mental evaluation of such activity. When the evaluation indicates that such activity may result in a significant adverse environmental impact that was not evaluated, or that is significantly greater than that evaluated, in the Final Envircnmental Statement er any addendum thereto, VEPCO shall provide a written evaluation of such activities and obtain prior approval from the Director, Office of Nuclear Raactor Regulation.
G.
This license is effective as of the date of issuance and shall expire one year after that date.
FOR THE NUCLEAR REGULATORY COMMISSION j
Harold R. Denton, Director, i
Office of Nuclear Reactor Regulation
Attachment:
Appendices A and B Technical Specifications Date of Issuance:
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