ML19345B959
| ML19345B959 | |
| Person / Time | |
|---|---|
| Site: | Zimmer |
| Issue date: | 11/26/1980 |
| From: | Borgmann E CINCINNATI GAS & ELECTRIC CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0694, RTR-NUREG-694 NUDOCS 8012030232 | |
| Download: ML19345B959 (50) | |
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y Waw THE CINCINNATI GAS &: ELECTRIC COMPANY eW CINCINN AT3.OHto 45201 .
E. A. BORO M AN N seweoR vtCE PRE $1 DENT Docket No. 50-358 November;46, 1980 3
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Mr. Harold Denton,. Director ;
Office of Nuclear Reactor Regulation .Qi W ,{:. ,
U.S.cNuclear Regulatory' Commission E v '_:
Washington, D.C. 20555 isuru F' E HE B -. =1L=
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RE: WM. H. ZIMMER NUCLEAR POWER STATIOd!-
. UNIT 1 - NUREG-0694 RESPONSE "
Dear Mr. Denton:
Enclosed are six copies of The Cincinnati Gas & Electric Company Plans for meeting NUREG-0694, "TMI-Related Requirements for New' Operating Licenses". In the near future, we will submit a. supplement. addressing these additional requirements listed in NUREG-0737 which have not been previously addressed in this current enclosure.
'Very truly-yours, .
O sn 1 57 d- THE CINCINNATI GAS & ELECTRIC COMPANY p- u : 5w.
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y e ?* x BY ea N <* E. A. BORGMANN N $ Senior Vice President 4
i ENB!:dewG h i Ericlosufti G i cd: Chirles Bechho$fer State of Ohio )
b Glegn O. Bright County of Hamilton)ss Frank F Hooper Troy L.. Conner, Jr. Sworn to and subscribed before James P. Fenstermaker me this c2.4 M day of November, 1980.
Steven G. Smith William J.'Moran
.J. ~ Robert Newlin William G. Porter, Jr. 3f .
James D. Flynn U.,ca t U L 8, h w F. T. Daniels 6/ Notary Public v W. Peter'Heile VIRGINIA P. MUHLHOFER James H. Feldman,-Jr. mtsy twc. state or GMo John'D. Woliver My commission Dpires My 28,1982 Mary.Reder David:K.' Martin
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Robert'A. Jones 1801203:
M ndrew-B._)Dennison .
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., ._a' : g_. - e e' .g 4 e-November-26,L1980 Docket No. 50-358 4
F WM. H. ZIMMER NUCLEAR. POWER STATION - UNIT - RESPONSE'TO NUREG-0694 "TMI-RELATED REQUIREMENTS FOR NEW OPERATING-LICENSES" k '
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- .e I.A.l.l. SHIFT TECHNICAL ADVISOR
- I. -POSITION 4 Aftechnical advisor to the Shift Supervisor shall'be
-present on all shifts and available to the Control-
. Room within:10: minutes. Although minimum training requirements have not been specified, Shift Technical Advisors should~ enhance the accident assessment function at the plant.
The Shift Technical Advisor.(STA) shall have a technical education, which isLtaught at the college level.and is equivalent to about 60 semester hours in basic subjects of engineering and science, and specific
'_. training in the design, function, arrangement and operation of plant systems and in the expected response
.of the plant'and instruments to normal operation, transients and accidents including multiple failures of
- equipment and operator errors.
II. IMPLEMENTATION CRITERIA 1 Shift Technical Advisors to the Shift Supervisors shall be present on all shifts and shall be available to-1 the control room within 10 minutes.
Shlft-Technical Advisors shall have a bachelors degree or equivalent in~a scientific or engineering discipline.
Education and training will be provided to all STA's as may be required, in accordance with the position established by the Institute of Nuclear Power Operations (INPO).
1 III. SCHEDULE Sufficient numbers of qualified STA's shall be available prior to fuel' loading to adequately cover all operating shifts.
IV. STATUS l ,
STA candidates have been identified and selected.
College level technical education is currently underway
-for STA candidates. The long term schedule for comple-tion of required college level work will be available i January 1,' 1981..
STA simulatoritraining schedules are finalized.
The complete'STA~ education and training schedule will be i
- available March:1, 1981.
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.I.A'.l.2 SHIFT SUPERVISOR ADMINISTRATIVE DUTIES
, I. POSITION Review the administrative duties of the shift super-visor and delegate functions that detract from or are subordinate to-the management responsibility.for.
assuring safe operation of the plant to other personnel not on duty in the control room.
II. IMPLEMENTATION CRITERIA The NRC-position will be utilized.
! III. SCFEDULE The implementation criteria shall be met prior to fuel loading.
IV. STATUS The administrative duties of the Shift Supervisor have been identified and documented.
These administrative duties are being reviewed and evaluated for delegation to appropriate personnel not on duty.in the control room and will be appropriately reflected in Station Administrative Directives.
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4 I.A.l.3 SHIFT ~ MANNING I. POSITION This position is set forth in NRC letter of July 31, 1980, (D. G. Eisenhut to All Licensees of Operating-Plants and Applicants for Operating Licenses and Holders of Construction Permits) .
II. IMPLEMENTATION CRITERIA Shift manning shall be as shown in the table below and shall be appropriately reflected in the Technical Specifications.
Minimum Shift Crew Composition License Applicable Operational Condition Category Mode 1,2,3 Mode 4,5 Senior Licensed Reactor Operator 2* 1**
Licensed Reactor Operator 2*** 1 Non-Licensed 2 1
- 0ne (1) senior licensed individual shall be designated as the Shift Supervisor;
- 0ne (1) senior licensed individual shall be in the control at all times; the Shift Supervisor may from time to time act as relief operator for the licensed senior reactor operator;
- This complement does not include the senior licensed individual assigned to directly supervise core alterations or fuel handling duties during those periods;
- * *One (1) licensed reactor operator shall be in the con. trol room at all times; the additional licensed operator may serve as relief operator for the licensed operator assigned to the control room.
Station Administrative Directives (SAD) shall be revised to I set forth the following overtime restrictions for senior licensed reactor operators, and licensed reactor operators performing a safety related function.
- 1) an individual shall not be permitted to work more than 12 straight hours (not including shift turnover time)
- 2) an individual shall not be permitted to work more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in any 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> period.
l 3) an individual shall 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 any 7 day period.
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I.A.l.3 SHIFT MANNING (cont'd) *
- 4) an individual shall not werk more than 14 consecutive days without having 2 consecutive days off.
- 5) deviations'frcm the above restrictions may be authorized by the Station Superintendent or higher levels of management with appropriate documentation of cause.
III. SCHEDULE The implementation requirements shall be met prior to fuel loading.
IV. - STATUS The present operating complement consists of: 18 personnel certified as senior reactor operator candi-dates; 5 personnel certified as reactor operator candidates; 8 personnel in cold license training with anticipated certification as reactor operator candidates on March 31, 1981; and 3 non-licensed personnel.
Hiring and training of additional personnel to meet upgraded staffing requirements is underway. Five non-licensed plant operators have been added.
Approximately five additional plant operators should Ebe available for operating responsibilities by March 31, 1981.
The SAD addressing overtime will be revised and approved by December 31, 1980.
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a I.A.2.1 IMMEDIATE UPGRADING OF OPERATOR AND SENIOR OPERATOR TRAINING AND QUALIFICATIONS I. POSITION Applicants for SRO license shall have 4 years of responsible power plant experience, of which at least 2 years shall be nuclear power plant experience (including 6 months at the specific plant) and no more than 2 years shall be academic or related techni-cal training.
Certifications that operator license a>plicants have learned to operate the controls snall be signed by the highest level of corporate management for plant operation.
Revise training programs to include training in heat transfer, fluid flow, thermodynamics, and plant transients.
II. IMPLEMENTATION CRITERIA Unique experience requirements will be established and met that accommodate the fact that the facility has not yet been in operation. Guidance of ANS 3.1, draft June, 1980, is being utilized.
Licensed Operator training programs shall be modified to address the " Guidelines for Heat Transfer, Fluid Flow and Thermodynamics Training" developed by the INPO.
Certifications pursuant to 10CFR55, Sections 55.10 (a) (6) and 55.33 (a) (4) and (5) shall be signed by the Manager, Electric Production Department.
III. SCHEDULE Experience requirements for license candidates will be met by June, 1981.
Training programs are presently bei'ng revised. Target date for. complete implementation is June, 1981.
Applications for NRC operator license examination will be submitted 90 days prior to scheduled fuel' loading.
IV. STATUS Experience training for senior reactor and reactor operators is underway and is approximately 15% complete.
Training programs in heat transfer,' fluid flow, and thermodynamics have been developed and are presently being taught.
I.A.2.3 ADMINISTRATION OF TRAINING PROGRAMS FOR LICENSTD OPERATORS I. POSITION Training instructors who teach systems, integrated responses, transient and simulator courses shall successfully complete ?n SRO examination.
Instructors shall attend appropriate retraining programs that address, as a minimum,. current operating history, problems and changes to-procedures and administrative limitations. In the event an instructor is a licensed SRO, his retraining shall be the SRO requalification program.
II. IMPLEMENTATION CRITERIA Certification of permanent Training. Center instructors or vendor instructors utilized on-site for teaching 4
systems, integrated responses, transient, and simulator courses shall require appropriate SRO qualifications.
~ Permanent facility training staff personnel teaching the above subjects shall also demonstrate SRO qualifi-cations which shall be documented.
Retraining certification shall be required.
III. SCHEDULE Full' compliance will be attained by fuel loading.
IV. STATUS Personnel presently performing the above functions meet the criteria.
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I.B.l.2 EVALUATION OF ORGANIZATION AND MANAGEMENT IMPROVEMENTS OF NEAR-TERM OPERATING LICENSE APPLICANTS I. POSITION The licensee organization shall comply with the findings and requirements generated in an interoffice NRC review of licensee organization and management, Establish an onsite group, independent of the plant staff, to perform' independent reviews of plant operational activities and a capability for evaluation of operating experiences at nuclear power plants.
II. IMPLEMENTATION CRITERIA Cincinnati Gas & Electric is continuing to review its technical and management capabilities in an effort to determine what organizational structure best meets the intent of NUREG-0731 for a single unit nuclear site.
The following steps have been taken toward meeting the guidelines of NUREG-0731.
- 1. Personnel for the Operations Review Committee (ORC) have been selected, including two non-Coupany members. A proposed operating charter is being reviewed for approval.
- 2. The technical and management review of the company is underway.
- 3. A charter for a permanent full-time Independent Safety Review Group is currently being reviewed. At present, it is our intention to have the ISRG conduct all s&fety reviews required by the Technical Specifications other than some limited ones assigned to the SRB.
- 4. Authorization has been given for increasing the number of operators and for expanded operator training.
- 5. Arrangements have been made with the University of Cincinnati to conduct on-site courses for qualifying STA candidates to meet the requirements of ANSI /ANS 3.1 III. SCHEDULE The first formal meeting of the ORC will be held in early 1981.
The technical and management review of the company will be l complete by the first quarter of 1981.- The formal organization, meeting the intent of NUREG-0731, will be in place by mid-1981.
The ISRG will be in place and functional by mid-1981.
IV. STATUS All activities described above are currently on schedule.
I.C.1 SHORT TERM ACCIDENT ANALYSIS AND PROCEDURE REVISION I. POSITION
' Analyze the design basis transients and accidents including single active failures and considering additional equipment failures and operator errors to identify appropriate and inappropriate operator actions.
Based on these anlyses, revise, as necessary, emergency procedures and training.
II. IMPLEMENTATION CRITERIA Emergency Operating Procedures (EOP's) - will be . revised to conform to the guidelines developed by-the BWR Owners' Group. These guidelines were submitted to NRC by the BWR Owners' Group on June 30,'1980.
III. SCHEDULE All EOP's will be revised by fuel loading or December 31, 1981, whichever is later.
IV. STATUS Zimmer EOP's are being redrafted. Because of the pre-liminary clarification contained in the September 5, 1980, Eisenhut letter, additional work must be performed in this area.
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I.C.2 SHIFT RELIEF AND TURNOVER PROCEDURES I. POSITION' Revise plant procedures for shift relief and turnover to require signed checklists and logs to assure that the operating staff (including auxiliary operators and maintenance personnel) possess adequate knowledge of critical plant parameter status, system _ status, availability and alignment.
II. IMPLEMENTING CRITERIA Revise station shift turnover procedures to require:
recording of major parameters acknowledgement of appropriate operating modes identification of major equipment out-of-service recording of electrical system status and by-passed instrumentation reviewing of log books and log sheets from previous shifts appropriate operator / supervisor acknowledgement that the checklist is completed and the on-coming shift crew assumes responsibility.
III. SCHEDULE Revised station shift turnover procedures will be com-pleted prior to fuel loading.
IV. STATUS This item is complete. .
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I.C.3 SHIFT SUPERVISOR RESPONSIBILITIES I. POSITION Issue a corporate management directive that clearly establishes the command dutier of the Shift Supervisor and emphasizes the. primary management responsibility for safe operation of the plant. Revise plant proce-dures to-clearly define the duties, responsibilities and authority of the Shift Supervisor and the control room operators.
II. IMPLEMENTING CRITERIA The NRC position will be utilized.
III. SCHEDULE This item will be completed prior to fuel loading.
IV. STATUS Corporate procedure will be drafted by March 31, 1981.
Station' Directives will be revised by April 30, 1981.
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I.C.4 CONTROL ROOM ACCESS I. POSITION Revise plant procedures to limit access to the control room to those individuals responsible for the direct operation of the plant, technical advisors, specified NRC personnel, and to establish a clear line of authority, responsibility, and succession in the control room.
II. IMPLEMENTING CRITERIA Assure that Station Directives adequately describe the line of authority, responsibility and command succession in the control roam.
Revise Station Directives to limit control room access to those persons responsible for direct plant operation, advisors, and specified NRC personnel.
Revise emergency plan procedures as may be necessary to assure compatability with these criteria and the emergency response organization.
III. SCHEDULE This item will be completed prior to fuel loading, j IV. STATUS The SAD revision pertaining to authority, responsibility and succession of command is complete.
The acces's procedure and emergency plan procedures will be complete June 30, 1981.
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I.C.5 PROCEDURES'FOR FEEDBACK OF OPERATING EXPERIENCE TO PLANT STAFF I. POSITION.
Review and revise, as necessary, procedures to assure that operating experiences _are fed back to operators and other peruonnel.
II. IMPLEMENTATTCN CRITERIA It is our intention'to have the Independent Safety Review Group _ (ISRG) review each licensee event. report and manufacturers' safety notices for sienificant nuclear and radiation safety impact en Zimmer Station.
Where significant impact is found in the judgement of the ISRG, it will formally report such in a prompt manner to the Station Review Board, the Station Superintendent and the Operations Feview Committee.
In addition, the Shift Technical Advisors will also review licensee event reports and manufacturers' safety notices frcm an operating standpoint and report items of significant impact on Zimmer promotly.to the Station Superintendent.
III. SCHEDULE Procedure s implementing the above will be in place by mid-1981.
IV. STATUS This effort is currently on schedule.
I.C.7 NSSS VENDOR REVIEW OF PROCEDURES I. POSITION Obtain nuclear steam supply system (NSSS) vendor review of lcw-pcwer testing procedures to further verify their adequacy. Obtain NSSS vendor review of power-ascension test and emergency procedures to further verify their adequacy.
II. IMPLEMENTING CRITERIA Procedures will be reviewed in'accordance with the NRC position.
III. SCHEDULE Low power testing procedures will be reviewed prior to fuel loading.
Power-ascension test and emergency procedures will be reviewed prior to achieving full power.
IV. STATUS Current practice and procedures require NSSS site operations manager to review startup procedures which include low-power and power-ascension procedures. This review is ongoing and is approximately 50% complete.
Station Directives have been revised to require NSSS on-site operations manager to document appropriate review of the Emergency Operating Procedures. This is an on-going activity and will be completed prior to full power.
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I.C.8 ' PILOT MONITORING OF SELECTED EMERGENCY PROCEDURES
. FOR NEAR-TERM OPERATING LICENSE APPLICANTS FULL POWER LICENSE REQUIREMENT Correct emergency procedures, as necessary, based on the NRC audit of selected. plant emergency operating
- procedures - (e.g. , small-break LOCA, loss of feedwater, restart of engineered safety features following a loss of-ac power,'steamline break or steam-generator tube rupture).
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I.D.1 CONTROL ROOM REVIEW I. POSITION Perform a preliminary assessment'of the control room to identify significant human factors deficiencies and instrumentation problems and establish a schedule upproved by the NRC for correcting deficiencies.
II. IMPLEMENTING CRITERIA The guidance and requirements contained in NUREG/CR-1580
(" Human Engineering Guide to Control Room Evaluation",
July, 1980) will be used for the review.
III. SCHEDULE Short-term review; issue report; October - November, 1980.
Establish schedule for correcting deficiencies -
Decenber , 198 0.
NRC control room team review - January, 1981.
Correct deficiencies in accordance with the schedule approved by NRC.
IV. STATUS Control Room Review Committee established Human factors consultant retained INPO & EPRI workshops attended First formal committee meeting - October 7 Review start - mid October
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.l I.G.1 TRAINING ~DURING LOW - POWER TESTING I.' POSITION Define and commitDto a snar'al low-pcwer testing program approved by NRC be conducted-at power levels no greater than 5 percent for the purposes
.of providing meaningful ~ technical information beyond that obtained in the normal startup test-
-program and_to provide supplemental training.
II.- -IMPLEMENTING CRITERIA.
Additional NRC input is needed.
III.- SCHEDULE Plan Definition - by fuel load' Conduct Training - during low power test program IV.- STATUS Awaiting idditional NRC input.
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a II.B.1 {* ACTOR COOLANT SYSTEM VENT I. POSITION Each applicant and licensee shall install reactor coolant system (RCS) and reactor vessel head high point vents remotely operated from the Control Room. While the purpose of the system is to vent noncondensible gases from the-RCS which may inhibit core cooling during natural circulation, the vents must not lead to an unacceptable increase in the probability of a LOCA or a challenge to containment integrity. Since these vents fcnnn a part of the reactor coolant pressure boundary, the design of the vents shal] conform to the requirements of Appendix A to 10CFR Part 50, " General. Design Criteria". The vent system shall be designed with sufficient redundancy that assures a low probability of inadvertent or irreversible actuation.
II. IMPLEMENTATION CRITERIA
- 1. Submit a description of the design, location, size, and power supply for the vent system along with results of analyses for lost,-of-coolant accidents initiated by a break in the venc pipe. The results of the analyses should demonstrate compliance with the acceptance criteria of 10CFR50.46. ,
- 2. Submit procedures and supporting analysis for the operator's use of the vents which also includes the information available to the operator for initiating or terminating vent usage.
III. SCHEDULE Provide information requested under Section II implementation by July 1, 1981.
IV. STATUS CRITERIA This item is complete.
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4 II.B.2 DESIGN REVIEW OF PLANT SHIELDING OF SPACES FOR POST-ACCIDENT OPERATIONS.
I.- POSITION With the assumption of a post-accident release of radioactivity equivalent to that described in Regulatory Guides ~1.3 and 1.4 (i.e. , the equivalent-of 50% of the core radiciodine, 100%'of the core noble gas inventory, and 1%.of the core solids, are contained in the primary coolant), each licensee shall perform a radiation and shielding design review of the spaces around systems'that may, as a result of an accident, contain highly radioactive materials. The design review should
. identify the' location of vital areas and equipment, such as the control room, radwaste control stations, emergency power supplies, motor control centers, and instrument areas, in which-personnel occupancy may be unduly limited or safety equipment may be
--unduly degraded by the radiation fields during postaccident operations of these systems.
Each licensee shall provide for adequate access to vital areas and protection of safety equipment by design changes, increased permanent or temporary shielding, or post-amWnt procedural controls.
The design review shall determine which types of corrective actions-are needed for vital areas throughout the facility.
II. IMPLEMENTATION CRITERIA A. Method of Analysis The inventories of radionuclide releases from the reactor core used in the Zimmer study were those specified by the NRC. Codes used for the analyses included RIBD for the initial core inventory assuming 3 years at' full power. Sub-program BAFFLE was used to model the movement of radioactivity.
External dose rates due to direct radiation from the drywell,-the pressure-suppression pool, the airborne and contained sources in the Reactor
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, B. ' Radiation sources considornd
- l. Direct radiation from the airborne. and liquidborne radioactivity in the primary containment.
- 2. Direct radiation from radioactivity in the Reactor and Auxiliary Buildings-contained in pipes and in equipment such as heat exchangers, tanks, and HVAC filters.
- 3. Direct radiation from airborne radioactivity in the_ Reactor Building (secondary containment).
- 4. Immersion dose rates from airborne radio-activity due to primary containment, ECCS, or e
shutdown cooling equipment leakage.
- 5. Radiation from the effluent plume outside the plant buildings.
C. Scope of Systems Review Systems which are'requ'. red to function as a result of'a LOCA and may contain radioactivity released from the core that were included in the shield design study include:
. Low Pressure Core Spray
. Low Pressure Coolant Injection
. Automatic Depressurization System
. Standby Gas Treatment System
. Reactor Core Isolation Cooling System
. Residual Heat Removal System
. Control Room HVAC
. Combustible Gas Centrol System
. Main Steam Isolation Valve Leakage Control System
. Equipment Leakoff Lines
. Floor Drains
. Area Sumps
. Waste Collection Systems ,
. LPCS and LPCI Overpressurization Protection
. Post-Accident Liquid and Gaseous Sample Lines i
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F D. Cont.yinment of Radioactivity
. The review included verification of the retention of radioactivity in the Reactor Building (secondary containment) except for controlled releases:
- 1. All leaked liquid drains to sumps whose contents are eventually transferred to the Reactor Building floor drain tanks.
- 2. No . points from which an automatic, uncontrolled transfer of radioactivity out of the Reactor Building can occur have been discovered.
The processing of secondary containment air by the SGTS is, of course, considered to be a controlled release.
E. Vital Area-Access The following vital areas have been evaluated for post-accident access:
- 1. Areas requiring continuous occupancy
. Main Control Room The Main Control room is accessible over the entira course of the accident.
. Technical Support Center This review provided shield design para-meters for a new facility which will contain the TbC. The new TSC will meet specified habitability requirements.
. Security Center The security central alarm station is serviced by the control room habitability system.
- 2. Areas requirine infrequent access
. Radiochemical / Chemical Laboratories These facilities are accessible to perform required post-accident analysis.
. Radwaste Control Room The Radwaste Control Room is accessible to perform required post-accident radwaste operations.
. Post-Accident Sample Station The shield design review revealed that the e.sisting reacter water and containment sampling _ facilities are inaccessible in the postulated post-accideat environment.
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Special sampling facilities are being designed for procurement and installation as reported in Item II.B.3.
F. Plant Modifications Being Evaluated as a Result of the Shield Design Review.
.-Upgrade Laboratory Facility HVAC
. Add shielding near Reactor Building personnel access point on 546' level to reduce streaming into Operations Support Center.
. Relocate Main Plant Vent Stack Radiation Monitoring System Panel.
. Additional controls to prevent transfer of the Reactor Building floor drain tanks and equipmen.t drain tank to Radwaste under accident conditiens.
. Additional controls to prevent cperation of the. Tendon tunnel sump pumps under accident conditions. -
. Administrative controls on operation of che Reactor Water Cleanup System under accident conditions.
III. SCHEDULE Milestone Date A. Complete identification of vital 1-2-81 .
equipment.
B. Complete modifications necessary Prior to O.L.
to provide access to vital areas.
IV. STATUS Activity % Complete A. Shield design' review 100 B. Identification of vital equipment / areas 30 l
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II.3.3- POST-ACCIDENT SAMPLING CRPABILITY I. POSITION A design and operational review of the radiological spectrum analysis facilities, and reactor coolant and containment atmosphere sampling systems shall be performed to determine the capability of personnel to promptly obtain and analyze (within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> from the time a decision is made to take a sample) a sample under accident conditions without incurring a radiation exposure to any individual in excess of 3 and 18 3/4 rer to the whole body or extremities.respectively.
Accident conditions should assume a. Regulatory Guide 1.3 release of fission products. If the review indicates that personnel could not promptly and safely obtain the samples, additional design features or shielding should be provided to meet the criteria.
The design' criteria to achieve these dose limit criterien in actual operation are those of GDC-19 (Appendix A, 10CFR Part 50) (i.e. 5 rem whole body, 75 rem extremities).
II. IMPLEMENTATION CRITERIA A Post-Accident Sampling System shall be provided.
The system shall be-designed to enable personnel to promptly obtain and analyze under accident conditions samples of reactor coolant and containment atmosphere.
The-system shall be designed to achieve the design criteria specified in Section I above with the following operaticnal objectives:
A. The samples obtainable shall include reactor coolant, suppression pool water, drywell atmos-phere, wetwell atmosphere, and secondary containmant atmosphere. The atmosphere samples shall be obtainable under both positive and negative pressure.
B. Diluted samples shall be collected if dilution is needed to reduce radiation exposure to personnel during sample collection and handling. The range of nuclide concentration considered shall be from approximately 1 uCi/g to 10 Ci/g.
C. It shall be possible to further dilute a_ collected sample to obtain a sample suitable for counting on the onsite Ge(Li) gamma analysis sytem.
D. Samples suitable for chloride analysis shall be obtainable within four days of reactor shutdown.
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-primary containment atmosphere oxygen and hydrogen:in the-range O to 10 volume percent..
F. Facilities.shall be provided to determine either total dissolved gases or dissolved H2 gas.in reactor coolant.
The'vuntilation' exhaust'frcm the sampling facility G. . will be filtered'with charcoal absorbers and
-HEPA filters.
III. SCHEDULE Milestones Date A. .BM'to vendors 10-1-80 B. Vendor proposals due 11-15-80 C. Issue P.O. 1-15-81 D. System delivery 9-18-81 E. Installation 10-5-81 F. Documentation to NRR for post-a ir.lementation review 12-15-81 IV. STATUS Engineering is 20% complete.
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t II.B.4 TRAINING FOR MITIGATING CORE -DAMAGE I. POSITION
~ Develop a training program to teach the use of installed equipment and systemc to control or mitigate accidents in which the core is severly damaged.
II. ' IMPLEMENTATION CRITERIA The guidelines for instruction are as recommended by the Institute of-Nuclear Power Operations, Special Training Program, STH-01-06-80, Rev. 0, (" Training.
Guidelines for Recognizing and Mitigating the Consequences of Severe Core Damage").
III. SCHEDULE This training program shall be implemented prior to fuel load.
IV. STATUS The material will be divided into eight categories and covered in six modules. This program is being developed and to date two of the eight categories have been implemented into the Phase V 1980-81 Modular Training Program. The remainil.g six categories are expected to be implemented within the 1980-81 training year.
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. v, II.D.1-- RELIEF AND SAFETY VALVE TEST REQUIREMENTS
, I. POSITION EstablishLa test program.to qualify!the. reactor' coolant system relief and safety _ valves under expected operating conditions for design basis transients and accidents. _
II. IMPLEMENTATION CRITERIA Cincinnati Gas &' Electric has been working- closely wid1 the-BWR Owners' Group-in the analysis of transients'that may.
lead.to:two-phase ~or liquid discharge ~through the safety / relief valves. ~ Based upon these and our independent analyses, we-do not believe that-high pressure testing'is necessary. However, fu the alternate shutdown cooling mode, the operator,will 4
intentionally put low pressure water through the safety / relief valves.- Therefore, the Zimmer safety / relief valves will be tested for this-condition.
4 III. SCHEDULE Testing will be performed as part of the BWR Owners' Group j
and is scheduled for completion in July, 1981.
IV. STATUS Discussions are underway with the NRC Staff on-the exact scope of the test program. Construction of the test facility is in progross.
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II.D.3 RELIEF AND SAFETY VALVE ?OSITION INDICATION
! I '. pOS'ITION-e' Reactor System Relief and Safety Valves shall be provided with
- a. positive' indication in the Control. Room. derived from a reliable i valve position detection ~ device or-a reliable indication of
. flow'in theidischarge pipe.
II. DIMPLEMENTATION CRITERIA
- The Zimmer Safety / Relief Valves-are manufactured by Crosby
-Valve-and'are equipped with an electromechanical-lift' indicating.
i' assembly which is mounted directly on. top of the SRV. 'The-housing for this assembly mechanically mates to the valve bonnet.-
-A. reverse-spring loaded actuator rod rides'the end of the valve
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spindle rod to directly transmit valve motion relative to the
. valve seating l surface.. Fully open or fully closed actuator-rod positions are: sensed by. dual microswitches through mechanical.
Electrical output from L
contact arms that ride the actuator rod.
the microswitches are-fed to the Control Room to remotely indicate'SRV position and-annunciation. Temperature elements mounted in1thermowells on each of the SRV' discharge pipes provide a confirmatory indication to the. Control Room of'SRV actuation or j long-term leakage. 'The temperature indicators serve asca back-up confirmation to the electromechanical devices.- Cabling,
, III. ' SCHEDULE'
.The'electromechanical lift indicating assembly is being installed on.the Safety / Relief Valves. Cables have been pulled.
Installation and preoperational testing will be completed prior
- to fuel. load.
t IV. STATUS
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Environmental and seismic qualification. testing of electromechanical position ~senscrs is currently underway;by the manufacturer.
! The qualificaticn program for this item is scheduled to be completed by mid-1981. CG&E Co. is monitoring the' progress j of this~and-other qualification test programs.
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4 II . E .1. l' AUXILIARY: FEEDWATER SYSTEM RELIABILITY EVALUATION I. POSITION
- 1. Provide a simplified auxiliary feedwater system
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reliability analysis that uses event-tree and fault-tree logic techniques.to determine the-potential for AFWS failure following a main feed-water transient, with particular emphasis on potential failures resulting frcm human errors,
-common causes, single point vulnerability, and test and maintenance outage.
- 2. Provide an evluation of the AFWS using the acceptance criteria of Standard Review Plan Section 10.4.9.
- 3. Describe tne design basis accident and transients and corresponding acceptance c'iteria for the AFWS.
- 4. Based on the analyses performed modify the AFWS,-
as necessary.
(This requirement is not applicable to Boiling Water Reactors) 9 8
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II.E.1.2 ' AUXILIARY FEEDWATER INITIATION AND INDICATION I. POSITION Install a control-grade system for automatic initiation of the auxiliary feedwater1 system that~ meets the single-failure criterion, is testable, and is' powered from the emergency buses,.and control-grade indication of auxiliary feedwater flow to each steam generator that is powered from emergency buses.
Upgrade, as necessary, automatic initiation of-the auxiliary feedwater system and indication of auxiliary feedwater flow to each steam generator to safety-grade quality.
(This requirement is not applicable to Boiling Water Reactors) j i
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II.E.3.l' EMERGENCY L POWER FOR PRESSURIZER HEATERS I. POSITION Install the capability to supply from emergency power buses a sufficient number of pressurizer heaters and asscciated controls to establish and maintain natural circulation in
' hot standby conditions.
(This' requirement is.not applicable to Boiling Water Reactors)
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II.E.4.1- CCNTAINMENT DEDICATED PENETRATIONS 3 I. POSITION The objective of the containment design is'to-improve the reliability and capability of the nuclear power plant containment structure to reduce the radiological consequences and risks to the public from design basis events and degraded-core-melt accidents.
II. IMPLEMENTATION CRITERIA Plants that have external hydrogen recombiners, such as Zimmer, are required to have redundant dedicated containment penetrations available such that the recombiner system is connected to the containment atmosphere without'the need to open large containment purging ducts or otherwise jeopardizing the containment function.
Zimmer will-provide a redundant dedicated penetration for the drywell atmosphere since that is where all of the sources of ignition are located and there is a spare penetration available.
A single dedicated penetration will be provided for the wetwell where.there is presently no penetration dedicated to the flammability control system. There is.one suitable penetration available for this use. Since there are virtually no ignition sources in the wetwell, the single penetration is considered to be sufficient.
III. SCHEDULE The criteria will be implemented by fuel load. Engineering will be completed and material ordered by January 1, 1981. Material delivery will cc=mence in April and end in June. Installation will be complete in August, 1981.
IV. STATUS Penetrations are being assigned and construction drawings will be prepared to pipe the penetrations to the recombiner.
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. .s II.E.4.2 CONTAINMENT ISOLATION DEPENDABILITY I. POSITION Provide (1) containment isolation on diverse signals, (2) automatic isolation of non-essential systems (including the basis for specifying the non-essential systems), (3) no automatic reopening of containment isolation valves when the isolation signal is reset.
II. IMPLEMENTATION CRITERIA Diverse containment isolation signals that comply with the recomnendations of SRP 6.2.4 are provided as indicated in Table 5.2-8 of the Wm. H. Zimmer FSAR. A review of the designation of essential and non-essential systems has been ccmpleted and the results will be submitted to the NRC. Non-essential systems identified by.this review will be automatically isclated by containment isolation signals. The isolation logic will be modified to prevent automatic reopening of the isolation valves when the isclation signal is reset. Each individual valve will require operator action to reopen.
III. SCHEDULE Will submit the documentation required by NUREG-0694 to the NRC by January 1, 1981. This documentation will include the basis for designating non-essential systems and will specify the modifications to the containment isolation system.
The required modifications will be complete prior to receiving a full power license.
IV. STATUS A review of the containment isolation system has been completed as required by NUREG-0694 with the following results.
A.
- 1. It was discovered that certain air operated and solenoid operated isolation valves would automatically reopen when the containment isolation logic is reset. Engineering to correct this deficiency has been ccmpleted.
- 2. Four (4) containment purge and relief valves IVQ005A&B and IVQOO8A&B do not close on _ solation signals. Engineering to add these valves to the containment isolation system is complete.
B. All other containment isolation functions / valves meet the requirements of SRP 6.2.4.
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9 II.F.1 ADDITIONAL ACCIDENT MONITORING INSTRUMENTATION I. POSITION
- 1. Fuel loading and low power testing requirements:
Provide proccdures for estimating noble gas, radiciodine, and particulate release rates if the existing effluent instrumentation goes off scale.
- 2. Dated requirements:
-Install continuous indication in the control room of the following parameters:
a) Containment pressure (range : -5 psig to 3x design pressure).
b) Containment water level (range - bottom to 5' above normal) .
c) Centainment hydrogen level (rance 10 percent volume).
d) Containment radiation (range 108 R/hr.).
e) Noble gas efflucnt (range 105 uci/cc (Xe-133)).
Provide capability to continuously sample and perform onsite analysis of radionuclide and particulate effluent samples.
This instrumentation shall meet the qualification, redundancy, testability and other requirements of the proposed revision to Regulatory Guide 1.97.
II. IMPLEMENTATION CRITERIA Procedures for estimating effluent release rates will be developed if it becomec apparent that the new effluent monitoring equipment is not operational prior to fuel loading.
The instrumentation required by NUREG 0694 will be installed.
This instrumentation will have the ranges defined in the forthccming clarification letter and will meet the requirements of the latest draf t of the proposed revision to Regulatory Guide 1.97. The capability to sample and perform onsite analysis of radionuclide and particulate effluent will be provided.
III. SCHEDULE Documentation describing the additional post-accident monitoring instrumentation will be submitted by June 1, 1981. Equipment installation is dependent on equipment availability and delivery time. The following installation schedule is therefore tentative and subject to revision:
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II.F.1 ADDITIONAL ACCIDENT MONITORING INSTRUMENTATION (Cont'd) a)[ Containment pressure Oct. ,81 b) Containment water level Oct.,81 c) Containment hydrogen ccncentration Jan.,82 d) :High range containment monitors Jan. ,82 e) Noble gas effluent mc1:itors Feb.,82 f) Effluent sampling Feb.,6'.
F IV. STATUS Engineering is approximately 20% complete and.is continuing i,
on an accelerated schedule.
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EII.F.2 INADEGUATE CORE COOLING INSTRUMENTS I. POSITION
- 1. Fuel loading and low power testing requirements.
Develop procedures to be used by operators to recognize inadequate core cooling with currently installed instrumentation in PWR's. Install a primary coolant saturation meter. Provide a description of any additional instruments or controls needed to supplement installed equipment to provide unambiguous, easy-to-interpret indication of inadequate core cooling, p rocedures for use of-this equipment, analyses used to develop these procedures, and a schedule for installing this equipment.
- 2. Dated requirements.
Install, if required, additional instruments or controls needed to supplement installed equipment in order to provide unambiguous, easy-to-interpret indication of inadequate core cooling.
II. IMPLEMENTATION CRITERIA Not applicable.to boiling water reactors.
III. SCHEDULE This item is complete.
IV. STATUS A subcooling meter is not needed in the BWR because the BWR operates in all power modes sith-liquid and steam in the reactor pressure vessel; thus, saturation conditions are always maintained irrespective.of system pressure.
The BWR Owners Group,of which the Cincinnati Gas & Electric Co.
is a member, submitted GE document NEDO 24708 (Aug., 1979) to the NRC staff in response to requests for information on-vessel level, transient analyses, etc. Section 2.3.2, Reactor Water Level Instrumentation contains an indepth analysis of BWR level instrumentation. The conclusion of that analysis is that'no additional hardware is needed to identify inadequate
-core cooling in the BWR.
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II.G. EMERGENCY POWER FOR PRESSURIZER EQUIPMENT FUEL LOAD & LOW' POWER TEST REQUIREMENT Motive and control components of the power-operated relief. valves and associated block valves and the pressurizer-level indication shall be capable of being supplied; frcm 'the off-site power. source or from the emergency ~ power buses when off-site power is not available.
(This. requirement is not applicable to Boiling Water Reactors)
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.II.K.1 IE' BULLETINS ON MEASURES TO MITIGATE SMALL-BREAK LOCAS AND LOSS OF FEEDWATER ACCIDENTS I. POSITION C.1.5 Review all valve positions, positioning requirements, positive controls and related test and maintenance procedures to assure proper ESF functioning.
C.l.10 Review and modify procedures for removing (restoring) safety related systems _frcm (to) service to assure operability status is known.
C.l.22 Describe the automatic and manual actions-necessary for auxiliary heat removal systems used when the main feedwater system is not operable.
C.l.23 Describe all uses and types of reactor vestal level indication for both automatic and manual initiation of safety systems. Describe other instrumentation that might give the same information on plant status.
II. IMPLEMENTATION CPITERIA Per IE Bulletin 79-08 response (copy attached) .
III. SCHEDULE Per IE Bulletin 79-08 response.
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IV. STATUS Per IE Bulletin 79-08 response.
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II.K.3 FINAL RECOMMENDATIONS OF E&O TASK FORCE I. POSITION C.3.9 For Westinghouse-designed reactors, modify the pressure integral derivative controller, if installed on the
'PORV, to eliminate spurious openings of the PORV.
C.3.10 For Westinghouse-designed reactors, if the anticipatory reactor trip upon turbine trip is to be modified to be bypassed at power levels less than 50 percent, rather than below 10 percent as in current designs, demonstrate that the probability of a small-break LOCA resulting frcm a stuck-open PORV is not significantly changed by this mcdification.
C.3.11 Demonstrate that the PORV installed in the plant has a failure rate equivalent to or less than the valves for which there is an operating history.
C.3.12 For Westinghouse-designed reactors, confirm that there is an anticipatory reactor trip on turbine trip.
(These items are not applicable to Boiling Water Reactors) l
III.A.l.l. UPGRADE EMERGENCY PREPAREDNESS I. POSITION (1) Provide an emergency response plan in substantial compliance with NUREG-0654, " Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plant" (which may be modified as a result of public comments solicited in early 1980) . (NRC will give substantial weight to FEMA's findings on offsite plans in judging their adequacy against NUREG-0654) (2) Perform an emergency response exercise to test the integrated capability and a ma;3r portion of the basic elements existing within emergency preparedness plans and organizations.
II. IMPLEMENTATION CRITERIA Emergency response plans for ZPS-1, the Sta of Ohio, the Commonwealth of Kentucky, the State of Indiana, and Clermont, Campbell, Pendleton, and Bracken counties will be revised to be in substantial compliance with the guidance set forth in NUREG-065 4. Procedures necessary to implement the revised plans will be developed. Plans will be submitted to NRC and/or FEMA for approval, as appropriate.
Public and media information/ education programs and emergency response training programs ' .11 be developed and implemented.
Communication and radiation r.onitoring capabilities of respon-sible offsite agencies will be upgraded. A system to provide prompt notification to the population will be designed and installed. ,
An emergency response exercise will be conducted which involves both onsite and offsite response to a simulated accident.
III. SCHEDULE Plans complete - December, 1980 FEMA /NRC Approval -
February, 19 81 Procedures complete - March, 1981 Exercise - Prior to fuel loading Prompt Notification System Functional - January, 1982
~IV. STATUS Plan development is proceeding in accordance with the above schedule.
III.D.3.3. IMPROVED IN-PLANT IODINE INSTRUMENTATION UNCER ACCIDENT CONDITIONS I. POSITION _
Providc equipment and associated training and procedures for accurately deterrining the airborne icdine concentration in areas within the facility where plant personnel may be present during an accident.
i' II. IMPLEMENTATION CRITERIA i The following equipment will be used for in-plant iodine sampling and analysis:
(
l A. Fixed Airborne Activity Monitoring System (FAAM) consisting '
i of six microprocessor based Continuous Air Particulate and Iodine .(P/I) Channels monitoring the six major HVAC ducts leading to the plant vent stack. The six iodine channels consist of 2 inch X 2 inch NaI(Tl) crystals, AM-241 seed imbedded for automatic gain stabilization.- The single channel analyzer (SCA) chassis are equipped with automatic s background subtraction via a.second SCA and NaI(Tl) crystal set on a window adjacent to the 364 Kev I-131 photopeak.
- Each FAAM has local and remote alarm and. readout capability.
4 RedIndant remote terminals are provided, one in the control
- Room and one in the Health Physics office.
B. Three air sample-panels provide the capability to obtain remote grab air P/I samples from twenty-five rooms throughout the plant.
j C. Eleven portable air samplers are available for obtaining grab air P/I samples throughout the plant.
D. Three portable Eberline Instrument Corp. Model SAM-2 SCA's are available for analyzing grab samples.
E. Silver Zeolite Iodine sample Cartridges are available ,
- for use'with any of the sample systems described above.
F. The existing fixed counting facilities are located well below grade level in the Auxiliary Building. Direct
! ' radiation' levels under Regulatory Guide 1.3 conditions are 0.015 RAD /hr during the 30 day post-accident period.
Existing fixed single and multi-channel analyzer shields
! are sufficient to reduce. direct radiation effects to levels
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III.D.3.3 IMPROVED IN-PLANT IODINE INSTRUMENTATION UNDER ACCIDENT CONDITIONS (cont'd)
G. Under the accident conditions described above and
- with extremely adverse meterology resulting in increased airborne levels in the. counting room, again direct airborne radiation effects are acceptabaly l
ameliorated with detector shielding and the detector i
caves are purged with bottled nitrogen to preclude the admission of airborne contaminents therein.
4 H. Should it become necessary, space can be made available in 'che habitable TSC to set up SAM-2 SCA's for iodine counting.
I. Bottled nitrogen is also available to purge iodine cartridges for Noble gas removal. ,
i III. SCHEDULE t
- All equipment will be in place, procedures developed and tested, and training conducted prior to fuel load.
IV. STATUS i
With the exceptions listed below, equipment is currently in place.
A. SAM-2 Portable SCA's have been ordered and are scheduled for delivery during 1980.
B. Silver Zeolite Cartridges will be on hand by June
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1981.
C. Bottled ~ nitrogen is availabic in the laboratory, however, supply lines to the gamma analysis caves have not been installed. . This will be completed by June, 1981. A supply af bottled nitrogen will be available in the TSC by September 1, 1981.
D. Procedures for post' accident iodine sampling and analysis, and training in their use will be completed i by. June 1, 1981.
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'ill.A.I.2 EMERGENCY SUPPORT FA'CTLITIES:
OPERATIONAL SUPPORT CENTER (OSC)
I. Position Establish an onsite operational support center (OSC), I
, . separate from, but with communications to the control room for use by operations support personnel during an accident.
i II. Implementation Criteria Three onsite operational support centers have been desig-
. nated to satisfy the OSC requirements and are fully described in the ZF2-1 Emergency Plan. The operational support centers are identified as the Maintenance OSC, the Rad / Chem OSC and l the Operations OSC. The operational support centers are equipped wi;h communications to the cor. trol room. ;
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, III. Schedule i
This item is complete.
IV. Status
, This item is complete 6
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- III.A.I.2. EMERGENCY RESPONSE FACILITIES:
i EMERGENCY OPERATIONS FACILITY (EOF)
I. Position Designate a near-site emergency operations facility (EOF). with adequate communications sufficient to provide i evaluation of radiation releases and coordination of all onsite and offsite activities during an accident. The EOF shall provide shielding against direct radiation,
. ventilation isolation capaoility, dedicated communication with the TSC and direct display of radiological and'
- meteorological parameters.
II. Implementation Criteria An interim EOF will be designated at the W. C. Beckjord ,
Station, approximately 9.5 miles north of ZPS-1 on U.S. Route
- 52. The interim EOF will be located in the second floor Ser-vice Building Conference Room and will meet the NUREG-0696 requirements for EOF location and staffing.
instrumentation and power supplies, data availability, and records.
III. Schedule The interim EOF will be functional by fuel loading. The permanent EOF will be functional by April 1,1982.
IV. Status Preliminary design and layout of the interim EOF have been initiated.
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III.A.1.2 EMERGENCY SUPPORT FACILITIES:
TECHNICAL SUPPORT CENTER (T5C)
I, Position Establish an onsite technical support center (TSC)
U separate from, but close 20, the control room for engine' ring and management support of reactor operations during an accident.
The TSC shall be large enough for the necessary utility person-nel and five NRC personnel, have direct display or callups of plant parameters, and dedicated communications with the control room, emergency operations facility and the NRC. The TSC shall have radiation monitoring and ventilation systems, including particulate and charcoal filters, and otherwise increase the radiation protection to assure that TSC personnel will not re-ceive doses in excess of 5 rems to the whole body and 30 rem to the thyroid for the duration of the accident.
II. Implementation The Technical Support Center (TSC) will be r ocated on the'first floor of a two (2) floor building addition to the Service Building at elevation 520 feet. The TSC has 2,050 sq.
ft. area.for 25 people and nearby is an NRC office, both located in structure which is habitable to the same degree as the con-trol room for postulated accident conditions. The TSC is approximately a two (2) minute walk from the control room.
The TSC will have dedicated communications with the control room, EOF, and NRC emergency centers. 'The TSC will have CRT displays of the Reg. Guide 1.97 parameters, inter-active mode, and those for the Safety Parameter Display System.
Power is supplied from offsite transmission system backed up by a diesel engine generator for lighting and HVAC loads, and in-cludes batteries and uninterruptible power supplies which will supply display systems.
Our objective is to obtain a data acquisition system which has an availability of .01 per NUREG 0696. We are negotiating with vendors to meet this objective. For the SPDS the object'ives are for qualification'to an operating
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basis earthquake (0.B.E.)' ' and ava11 ability of .001, also per NUREG 0696.
III. Schedule Construction of the TSC is scheduled to start November 1, 1980 with completion date of April 1, 1982. New centers to support- the TSC, a Power and a Computer Center, are scheduled to support the TSC completion and operation. Construction of the Computer Center has begun and Power Center is to begin November 1, 1980.
III.A.1.2. EMERGENCY SUPPORT FACILITIES:
- TECHNICAL SUPPORT CENTER (ISC) (cont'd)
IV. Status Engineering progress is as follows:
TSC 30% complete Power Center 50% complete Computer Center 100% complete Engineering has begun on control and instrumentation and is 5% complete. No formal preoperational testing is required.
Power supply, HVAC, display and associated equipment in bid evaluation stage. Deliveries are requested to meet the schedules of the assoicated installation.
As stated under Schedule, construction of the Computer Center has begun.
~III.D.1.1 INTEGRITY OF-SYSTEMS OUTSIDE CONTAINMENT LIKELY TO CONTAIN RADIOACTIVE MATcRIAL FOR PWRs AND BKRs I. POSITION' Applicants shall implement a program to reduce ~ leakage f:om systems outside containment that would or could contain highly radioactive fluids during a serious transient or accident to as-low-as-practical levels.
This program shall include the following:
- 1. Immediate Leak ~ Reduction
- a. Implement all practical leak reduction measures for all systems that-could carry radioactive fluid outside of containment.
b.-. Measure actual leakage rates with system.in operation and report them to the NRC.
- 2. Continuing Leak Reduction Establish and implement a program of preventative maintenance to reduce leakage to as-low-as-practical levels. This program shall include periodic integrated leak tests at intervals not to exceed each refueling cycle.
II. IMPLEMENTATION CRITERIA.
Inspection criteria for operations personnel who provide shiftly . inspection of the operating plant t
shall. include criteria for identification of unusual
{ or abnormal leakage.
- Thv training and retraining programs till emphasize the significance of leakage, the need.for continual personnel awareness-and-the need to identify and take prompt measures to correct leakage problems.
Leakage reduction to as low as practicable levels will be implemented during the preoperational testing program. Components from which leakage can be measured or estimated will be identified and leakage values documented. Practical leakage re-duction _ measures will be. applied and corrective actions documented.
A formal operational leakage assessment program will be developed and implemented during each refueling
-cycle that include the systems assessed during the preoperational program. Gaseous systems testing will include helium leak detection or other equivalent methods, t
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l The forma'=ASME Code Section XI requirements for Class I I and III systems as well as the installed Leak De .ccion System are utilized in conjunction with and in support of the leakage reduction pro-grams.
III. S{ f.LE The preoperational leakage reduction program will be implemented in accordance with the project preop test schedule.
The information requested in the clarification section of the enclosure to D.G. Eisenhut's letter of September 5, 1980 will be provided 4 months prior to issuance of an operating license.
IV. STATUS A listing of systems to be included and excluded in the preoperational program and the formal, periodic program (each refueling) has been developed.
Procedures and criteria..for those programs are partially drafted.
e III.D.3.4 CONTROL ROOM HABITABILITY-4 I. Position Confirm that control rocm operators will be adequately protected against the effects of accidental releases of toxic and radioactive gases and that the station can be safely operated or shut down under design basis accident conditions.
II. Implementation Criteria-The ZPS-1 control avou will be evaluated to assure com-
, . pliance with the criteria set forth in Standard Review Plan sections 2.2.1-2.2.2, 2.2.3, and 6.4 using the guidance set forth in Regulatory . Guides 1.78 and 1.95, and in accordance l
with the additional clarification provided in D. G. Eisenhut's letter dated September 5, 1980.
III. Schedule 4
Results of the analysis of the ZPS-1 control room will be submitted by January 1, 1981.
IV. Status
. Analysis of the ZPS-1 control room in accordance with the September 5, 1980 clarification letter has been initiated.
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ATTACHMENT 1 l NRC IE BULLETIN 79-08 l
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THE CINCINNATI GAS & ELECTRIC CO.N!PANY CtNCINN ATl OHto 4 520i August 7, 1979 C. A. BORG MAe. U.S. Nuclear Regulatory Commission Region III 799 Roosevelt Road Glen Ellyn, Illinois 60137 ATTN: Mr. James G. Keppler, Director RE: W)!. H. ZIMMER NUCLEAR POWER STATION - UNIT 1 NRC IE BULLETIN 79-08 EVENTS RELEVANT TO BOILING WATER F0WER REACTORS IDENTIFIED DURING THREE MILE ISLAND INCIDENT W.O. 57300, JOB E-5590, FILE # 956, DOCKET # 50-358 Gentlemen: The attached document is furnished in response to IE Bulletin 79-03. We believe this information provides a complete response to NRC IE 3ulletin 79-08. Very truly yours, THE CINCINNATI GAS & ELECTRIC COMPANY p .. .- f {sh >ty W ~ 0 E.A. BORG!tANN, SR. VICE PRESIDENT HCB /kj.d cc: U.S. Regulatory Conmission Office of Inspection and Enforcement Division of Reactor Operations Inspection Washington, D.C. 20555 W.W. Schwiers S.G. Salay J.R. Schott W.D. Waymire J.D. Flynn H.C. Brinkmann K.K. Chitkara l 2 R. usk General File k .. B (IY 7Q t N.
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. 4 . ,9 RESPONSES TO IE BULLETIN 79-03 i ITEM 1. Review the description of. circumstances described in Enclosure 1 .
j of IE' Bulletin 79-05 and the preliminary chronology of the THI-2 l 3/28/79 accident included in Enclosure 1 to IE Bulletin 79-05A.
- a. This review should be' directed toward understanding:
(1) the extreme se'riousness and consequences of the
! simultaneous blocking of'both trains of a safety system at ,
the Three Mile Island Unit 2. plant and other actions taken ; during the early phases of the accident; (2) the apparent
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operational errors which led to the eventual core damage; (3) the necessity to syste=atically analyze plant conditions and parameters and take appropriate corrective action. $ b. Operational personnel should be instructed to (1) not. over-ride automatic action of engineered safety features unless continued operation of engineered safety features will re-sult in unsafe plant conditons (see Section'Sa of this ^ bulletin); and ~ (2) not make operational decisions based solely on a single plant parameter indication when one or more confirmatory indications are available,
- c. All licensed operators and plant management and supervisors with operational responsibilities shall participate in this review and such participation shall be documented in plant records.
3 RESPONSE TO ITEM 1 1 All ' cold license candidates including operators, plant osanagement and super-visors with operational responsibilities have participated in a review of the
.Three Mile Island Accident. This included the preliminary chronology of the TMI-2, 3/23/79 accident included in I&E Bulletin 79-05A, Enclosure 1.
RESPONSE TO' ITEM la_ ^ The circumst: aces described in I&E Bulletin 79-05, Enclosure I and the understanding of subjects discussed in I&E Bulletin 79-08 Item la are being reviewed as follows: I a. A cold-license candidate s.mulator refresher course was , , conducted in July and August, 1979. .The course rein-forced and demonstrated BWR level instrumentation de-
. sign, interpretation, minor transients and upset con- , .ditions degrading to loss _of coolant conditions. Also j, covered operator decisions to preclude ~ emergency system i component. operation.
- b. A formal. presentation of the events leading to and chronology- !
as now kno'wn; with lessons learned be complete by Nov. 1, 1979 l I t ' l i <.
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_ fM %~ ; RESPONSE TO ITEM _la CONT'b 4 c.- The continuing onsite training program, Phase 'I, will . provide additional review with opernting licensed sup-crvisory and management personnel as further ir. formation is made available. RESPONSE TO ITEM lb O Station Administrative Directives (SAD's) have been revised to instruct operational personnel that automatic action of engineered safety features and isolation' signals shall not be manually overridden unless: l
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- a. Continued operation of the engineered safety features or isolation signals will result in unsafe plant con-ditons, or
- b. It is known or positively determined that the automatic action .was initiated by a spurious or erroneous signal and it is verified that operaton of the engineered safety feature or isolation is not required, or j
- c. Approved procedures specifically allow manual override
- under specific conditions, and those conditions are veri-4 fled to be satisfied.
4 Additionally, these points will be periodically restressed during the op- ] erator requalification' training program. SAD's haveLbeen revised to instruct operational personnel that when one or ! more confirmatory indicators are available, operational decisions shall not
.be made' based solely on a single plant. parameter indication.. Additionally, the SAD provides instructions to operational personnel that all available . ] info,rmation should be considered in decisions to' manually initiate, term-inate, or control. operation of safety systems.
RESPONSE TO' ITEM ic. t ~ Attendance during the review described in la, above was documented. ITEM 2 Review the containment isolation initiation assign and procedures, and prepare and implement all changes necessary.to initiate containment is ' olation, whether manual or automatic, of all lir.9s whose isolation does
, not degrade needed safety features or cooling capability, upon automatic initiation of safety injection.
RESPONSE TO ITEM 2 .. a i Containment isolation design and procedures have been reviewed. Containment-l - isolation _of all lines'whose isolation does not degrade needed safety features . .or cooling capability is initiated either automatically or~ manually upon au'to-z matic initiation of safety injection.
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A .. . . ~ . . . ~ . - . . . - ~ ~ ~ - 3~ *- ITEM 3 i Describe the actions, bott automatic and manual, necessary for proper functioning of the auxiliary heat rc= oval systems (e.g., RCIC) that are used when the. main feedwater system is not operable. For t.ny manual action necessary, describe in su==ary form the procedure, by which this action is taken itt a timely. sense. RESPONSE TO ITEY ,3 The auxiliary heat removal systems provided to remove decay heat from the reactor core and containment following loss of the feedwatcr system are: High Pressure Core Spray (HPCS) System Reactor Core Isolation Cooling (RCIC) System-
. ' Safety Relief Valves (SRV) and Automatic De-pressurization System (ADS)
Low Pressure Core Spray (LPCS) System Lov. Pressure Coolant Injection (LPCI) Mode of the Residual Heat Removal (RER) System The operation of systems needed to achieve initial core cooling, containment cooling, and extended core cooling for long term plant shutcown is decribed below. 1 4
- a. INITIAL CORE COOLING Following loss of feedwater and subsequent reactor scram, a low reactor water level signal will auto-matically_ initiate main steam line isolation valve closure'.
The safety relief valves (SRV's) will automatically. actuate to maintain reactor pressure. At the same time,
, the low water level signal automatically initiates the HFCS and RCIC Systems. These systems will continue to inject water into the reactor vessel until a high water 4
level signal closes the HPCS injective valve and trips the RCIC system. Following a high reactor water level trip, the HPCS injection valve will again reopen when reactor water level decreases to the low water level setpoint. _ The RCIC System must be manually reset before it will reinitiate after a high water level trip. The HPCS and RCIC Systems. have redundant supplies of water, cormally taking suction from the cycled condensate storage tanks (CST's). The HPCS and RCIC Systems suction will auto-I
=atically transfer from the CST to the suppression pool if the CST water is depleted or if the suppression pool water level increases to a'high-level.
The licensed operator can manually initiate the HPCS and
- RCIC Systems.from the main control room before the low reactor water level automatic initiation level is reachcd.
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- The operator has the option of manual control af ter auto--
matic initiation and can maintain reactor water level by 4 throttling' system flow rates. -This would prevent a trip of the systems due to high water level. The operator can verify that these systems are delivering water to the re-actor vessel b.v any or all of the listed methods. a, . a, , , , . - - -- - n a
- - .* * . 3 ITEM 3a CONT'D
- 1. Verifying reactor water level increases when systems _ initiate using redundant icvel-indi-Cators.
- 2. Verifying system flow rates u-ing flow indicators in the control room.
- 3. Verifying system flow is to the reactor by checking contrcl_ room position indication of motor-operated valves. This assures no dive Tion of system flow from the reactor.
The HPCS and RCIC Systems can maintain. reactor wate level at full reactor pressure and until pressure decreases to where low pressure systems such as the LPCI Mode of the RHR' or Low Pressure Care Spray (LPCS) can maintain reactor water level,
- b. CONTAINMENT COOLING-After reactor scram and isolation and establishment of sat-isfactory core cooling, the operator would initiate the suppression pool cooling mode of RHR. This mode of opera-tion removes heat resulting from safety relief valve (SRV) discharge and/or RCIC exhaust to the suppression pool. This is accomplished by placing one loop of the RHR System in the suppression pool cooling mode; (RHR suction from and discharge to the suppression pool through one RHR heat exchangcr.)
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The operator verifies proper operation of the RBR System containment cooling function from the main control room by:
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- 1. Verifying RHR and' Service Water (WS) System flow using system control room flow indicators.
- 2. Verifying correct RHR and Service Water System flow paths using control room position indication of motor-operated valves.
- 3. Monitoring suppression pool water temperature.
Even though one loop of the RHR is in the Suppression pool cooling mode, core cooling is its primary function. Thus, if a high drywell pressure or low water _ level signal is re- b ceived at any time during the period when the RHR is in the suppression pool cooling mode, the RER system will automat-ically revert to the LPCI injection mode. In addition, the > HPCS and LPCS Systems would automatically start upon receipt of all ECCS initiation-signals (s). The HPCS System functions as described in response 3.a and immediately injects water into the reactor vessel to maintain reactor water inventory. The LPCS and LPCI Mode of RHR would injec', water into be re-actor vessel if. reactor pressure is below the respective
*.. b ITEM 3b CONT'D
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system discharge pressures. Upon receipt of conincident low reactor water level and high drywell pressure signals, and af ter a two minute time delay, the Automatic Depressur-ization System (AUS) would relieve reactor pressure to allow the low pressure systems (LPCS & LPCI) to inject water into the vessel. (Also see responses 5.a and 5.b)
- c. EXTENDED CORE COOLING
'a' hen the reactor has been depressurized, the RHR System can be placed in the long term shutdown cooling mode. The op-erator manually terminates the LPCI made of one RHR loop and places that loop in the shutdown cooling mode as follows:
- 1. Trip the selected RHR pump
- 11. Close motor operated valves (MOV's) in the suppression pool suction and discharge lines of the selected loop.
iii. Open the RHR shutdown cooling suction and dishcarge MOV's iv. Restart the selected RHR pump In this operating mode, the RUR System can cool the reactor to cold shutdown. Proper operation and flow paths in this mode can be verified by methods similar to those described for ine containment cooling mode. 1 TEM 4 Describe all uses and types of vessel level indication for both automatic and manual initiation of safety systems. Describe other redundant instrumentation which the operator might have to give the same information regarding plant status. Instruct operators to utilize other available information to initiate safe.ty systems. RESPONSE TO ITEM 4 Description of the reactor vessel level automatic initiation for the safety systems is provided in Chapter 7.3 of the FSAR. The description of the vessel level indication for manual initiation of the safety systems is de - scribed in Chapter 7.5 of the FSAR. As described in response to item 1, .D's have been revised to provide instructions to operating personnel ta util!La all available information. ITEM 5 Review the action directed by the operating procedures and training instructions to ensure that:
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- a. Operators do not override automatic actions of engineered
. safety features, unless continued operation operation of engineered safety features will result in unsafe plant conditions (e.g. vessel integrity).
- b. Operators are provided additional information and instructions to not rely upon vessel level indication alone for manual actions, but to also examine other plant parameter indi-cations in evaluating plant conditions.
. . .s Y e RESPONSE TO ITEM Sa 1
. Approved station -operating-and training procedures review accomplished by September 15, 1979 to ensure that appropriate instructions clearly specify that' operating personnel au.not override automatic actions of engineered safety features u dess continued operation of these systems will result in unsafe plant conditions. However, it has been determined that several valid reasons exist fer allowing an override of an automatic initiation signal or shutdown of a system after it has been automatically initialed (see response
, to Item 1, above) . For example, a. If an automatic initiation of the HPCS System and RCIC System occurs, the operator is permitted to shutdown the HPCS System if the RCIC System is capable of maintaining acssel level. This is allowed to prevent a trip of both systems due to high water level. As noted in response to Item 3 of this Bulletin, a trip of the RCIC System requires manual operator action to reset. s,
- b. The procedures allow the operator to_ nanually overrid s automatic actuation of the ADS if it has been determined that adequate water level is being maintained by the HPCS System. In this case, LPCI or LPCS is not required and, therefore, ADS actuation can be interrupted. This ,
override is permitted to allow a controlled cooldown and depressurization of the reactor and prevents injection of suppression pool water into the reactor when it is not required.
- c. The procedures allow transfer of part or all of the RHR System from the LPCI mode of operation to the Suppression Pool Ceoling or Shutdown Cooling modes of operation when adequate reactor water level is maintained with part of the RHR System and/or other systems. This is permitted to insure that suppression pool water temperature and containment pressure limits are maintained and provides controlled cooldown of the primary system.
RESPONSE TO ITEM Sb
-The SAD concerning station operations has been revised as stated in the re-sponse to item 1 to. assure operators consider all available information in decisions.to take manual action. Operating procedures for specific events do describe expected parameter indications. Clarification, and where appropriate amplification will be made to specific operating procedures that describe parameter indications. Changes to procedures will be included in appropriate portions of operator training sessions. However, it should .t be recognized that. events may occur such that vessel level indication might be the only immediately obvious parameter affected. We are reluctant to issue instructions which might be con.idered contrary t'o the directive for operators to believe and respond conservatively to instrument indications unless the indications are proven to be incorrect.
ITEM 6. Review 'all safety-related valve positions, positioning requirements and l
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O ITD1 6 CONT'D
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g positive controls to assure that valves remain positioned (open or closed) in a manner to ensure the proper operation of engineered safety features. Also review related procedures, such as those for maintenance, testing, plant and systen startup, and supervisory periodic (e.g. , daily / shift checks,) surveillance to ensure that such valves are returned to their correct positions following necessary manipulations and are maintained in their proper positions during all operational modes. RESPONSE TO ITEM 6 Prior to a plant startup safety c-lated system valve checklists are com-pleted to establish and verify valva positioning. On plant restarts following short outages when , major maintenance is not performed, checklists may be com-pleted only on cystems where maintennnee was performed. In accordance with CC&E Company policies and job descriptions qualified op-erating personnel are the only individuals who may position valves. These activities are performed under the direction of the licensed reactor op-erator (nuclear control operator) or the licensed senior reactor operator (shift supe rviso r) . Surveillance activitics are performed using individual procedures which have been prepared in accordance with Station Administrative Directives governing surveillance. These directives provide guidance in regard to procedural preparation to ensure that systems are properly returned to service following surveillance testing. Safety related operational checklists and all surveillance procedures are -reviewed by an independent person knowledgeable in the operation of the station and by the Station Review Board. Final approval is by the Station Superintendent. In additon, all procedures including checklists, are re-viewed at least once every two years for required changes (or prior to use as may be the case for special . procedures) and following unusual transients. In addition to specific procedural control, licensed operators check main co, trol room switches, annunciator panels, etc. for normal indications each shift; this includes valve position indications. Valves critical to system operation but without position indication in the control room are locked or seal . ' in the required position. Their positioning is verified with completion of startup checklists and in most cases with periodic com-pletion of surveillance testing. Maintenance activities on safety-related equipment are performed using a station Work Request (WR). The Work Request form identifies isolation activities required and specifically identifies when the equipment is re-turned to normal operation.
.h We believe that the procedures, controls, and reviews described above are adequate to assure proper valve positioning.
ITDI 7 Review your operating modes and procedures for all systems designed to transfer potentially radioactive gases and liquids out of the primary containment to assure that undesired pumping, venting or other release of radioactive liquids and gases will not occur inadvertently.
?! h a' ITEM 7 CONT'D:
In particular, ense that such an occurence would not be caused'by the resetting of engine. ed safety features instrumentation. List all such systems and indicate:
- a. Whether interlocks exist to prevent transfer when high radiation indication exists, and
- b. . Whether such' systems are isolated by the containment isolation signal.
- c. The basis on which continued operability of the above features is assured.
RESPONSE TO ITEM 7-All systems designed to transfer potentially radioactive gases and liquids f rom the primary containment are provided with automatic isolation valves. Isolation ' signals are initiated by a variety of reactor, containment or system conditions. The trip setpoints for these automatic isolation sig-nals are listed in Technical Specification Table 3.3.2-2. Valve groups that are operated by these trips are listed in Technical Specification Table 3.3.2-1. These valve groups are listed in Technical Specification Table-3.6.3-1. The containment radiation monitoring system is not part of the containment isolation system with the exception of the main steam line radiation moni-tors. The system provides information to the operator for the manual con-trol of the primary containment systems. ITEM 8 Review and modify as necessary your maintenance and test procedures to en-sure that the require:
- a. Verification, by test or inspection, of the operability of redundant safety-related systems prior to the removal of any safety-related system from service.
- b. Verification of the operability of all safety-related systems when.they are returned to service following maintenance or testing.
- c. Explicit notification of invol"ed reactor operational personnel whenever a safety-related system is removed a from and returned to service.
RESPONSE TO ITEM Ba, b and c .h The removal of equipment from service is controlled administratively and/or with the use of the WR, The procedures which deal with equipment isolation specifically reference the responsible individuals to the applicable technical specification. m .- ,w -.~ ,- --r -- --m ~ .,n . ... e-,e , - - ,m -,
c _ ~ 0 M, 3 RESPONSE TO ITE:t 8a, b and c CONT'D Maintenance performed on safety-related equipment is controlled by the WR. The (Work Raquest) SAD assigns the Shif t Supervisor the responsibility to identify technical specification requirements that pertain to any main-tenance activity. Any post work surveillance testing that is required is also identified on the WR. Written authorization to begin safety related corrective maintenance and any surveillance testing niust be obtained f rom the Shif t Supervisor. Re-moving any equipment from service must be reviewed and authorized by the Shift Supervisor. In addition, station operators are the only personnel who remove equipment from service and return it to service. Records of equipment tagged out and jumper and lifted leads are maintained by the operations group. ITDI 9 Review your prompt reporting procedures for NRC notification to assure that NRC is notified within one hour of the time the reactor is not in a con-trolled or expected condition of operation. Further, at that time an open continuous communication channel shall be established and maintained with NRC. RESPONSE TO ITEt 9 Prompt reporting procedures for URC notification will be revised to es-tablish a continuous open line of communication with the NRC as rapidly as possible in the event the reactor is not in a controlled or expected condition of operation. It is our intent to work with the Region III Office of Inspection and Enforcement in the development of continuous communication channels. ITEM 10 Review operating modes and procedures to deal with significant arounts of hydrogen gas that may be generated during a transient or other accident that would either remain inside the primary system or be released to the containment. RESPONSE TO ITEM 10 In the eventhydrogen gas is generated by metal water reaction or radiolysis, the following methods are available to relieve the gas from the reactor vessel (primary system).
- a. SAFETY RELIEF VALVES (SRV's)
There are thirteen SRV's located on the main steam lines that relieve to the quenchers located below the suppression pool water level. Since there is about 20 feet between the top of the core and the main steam line nozzles, a large vo?.ume of noncondensable gas can be relieved to the suppression peol via this pathway.
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RESPONSE TO ITEM 10 CONT'D
- b. LOSS OF COOLANT ACCIDENT A direct leakage path to the primary containment is 'reated for release of noncondensables for certain postulatei line ruptures.
- c. REACTOR ',EAD VENT The reactor vessel head vent relieves directly from the top of the vessel t:c'd via remote manual control f rom the main control room. The vent is directly piped to the reactor building equipment drain tank, and under supervision of a licensed operator or senior operator, the M0V's can be op-ersted to relieve noncondensable gas to the primary contain-ment.
After venting the hydrogen gas from the reactor vessel to the primary containment, the condensation of hydrogen and oxygen is continuously monitored by redundant trains of containment monitors. The Primary Containment Combustable Gas Control System is activated remotely from the main control room, and this system, through the use of hydrogen recombiners, can adequately handle the postulated volume of hydrogen gas gen-erated from radiolysis and/or metal water reaction. Operating procedures addressing the generation of hydrogen gas in the reactor vessel and release of hydrogen gas to the primary containment will be reviewed by December 15, 1979. G 9 6
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