NRC Generic Letter 1984-005: Change to NUREG-1201 Operator Licensing Examiner StandardsML031150645 |
Person / Time |
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Site: |
Beaver Valley, Millstone, Hatch, Monticello, Calvert Cliffs, Dresden, Davis Besse, Peach Bottom, Browns Ferry, Salem, Oconee, Mcguire, Nine Mile Point, Palisades, Palo Verde, Perry, Indian Point, Fermi, Kewaunee, Catawba, Harris, Wolf Creek, Saint Lucie, Point Beach, Oyster Creek, Watts Bar, Hope Creek, Grand Gulf, Cooper, Sequoyah, Byron, Pilgrim, Arkansas Nuclear, Three Mile Island, Braidwood, Susquehanna, Summer, Prairie Island, Columbia, Seabrook, Brunswick, Surry, Limerick, North Anna, Turkey Point, River Bend, Vermont Yankee, Crystal River, Haddam Neck, Ginna, Diablo Canyon, Callaway, Vogtle, Waterford, Duane Arnold, Farley, Robinson, Clinton, South Texas, San Onofre, Cook, Comanche Peak, Yankee Rowe, Maine Yankee, Quad Cities, Humboldt Bay, La Crosse, Big Rock Point, Rancho Seco, Zion, Midland, Bellefonte, Fort Calhoun, FitzPatrick, McGuire, LaSalle, 05000000, Zimmer, Fort Saint Vrain, Shoreham, Satsop, Trojan, Atlantic Nuclear Power Plant, Skagit, Marble Hill |
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Issue date: |
04/02/1984 |
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From: |
Eisenhut D G Office of Nuclear Reactor Regulation |
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To: |
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References |
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NUREG-1201 GL-84-005, NUDOCS 8404020019 |
Download: ML031150645 (10) |
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Similar Documents at Beaver Valley, Millstone, Hatch, Monticello, Calvert Cliffs, Dresden, Davis Besse, Peach Bottom, Browns Ferry, Salem, Oconee, Mcguire, Nine Mile Point, Palisades, Palo Verde, Perry, Indian Point, Fermi, Kewaunee, Catawba, Harris, Wolf Creek, Saint Lucie, Point Beach, Oyster Creek, Watts Bar, Hope Creek, Grand Gulf, Cooper, Sequoyah, Byron, Pilgrim, Arkansas Nuclear, Three Mile Island, Braidwood, Susquehanna, Summer, Prairie Island, Columbia, Seabrook, Brunswick, Surry, Limerick, North Anna, Turkey Point, River Bend, Vermont Yankee, Crystal River, Haddam Neck, Ginna, Diablo Canyon, Callaway, Vogtle, Waterford, Duane Arnold, Farley, Robinson, Clinton, South Texas, San Onofre, Cook, Comanche Peak, Yankee Rowe, Maine Yankee, Quad Cities, Humboldt Bay, La Crosse, Big Rock Point, Rancho Seco, Zion, Midland, Bellefonte, Fort Calhoun, FitzPatrick, McGuire, LaSalle, 05000000, Zimmer, Fort Saint Vrain, Shoreham, Satsop, Trojan, Atlantic Nuclear Power Plant, Skagit, Marble Hill |
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Category:NRC Generic Letter
MONTHYEARML23200A1832023-08-0303 August 2023 Closeout of Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors ML17067A2782017-04-18018 April 2017 Non-Power Reactor Closeout of Generic Letter 2016-01, Monitoring of Neutron Absorbing Materials in Spent Fuel Pools for the Armed Forces Radiobiology Research Institute, Docket No. 50-170 (CAC No. A11010) ML17067A4042017-04-17017 April 2017 Washington State University GL 2016-01 Closeout Form Letter for Rtrs with No Credited NAM NRC Generic Letter 2007-012007-02-0707 February 2007 NRC Generic Letter 2007-01: Inaccessible or Underground Power Cable Failures That Disable Accident Mitigation Systems or Cause Plant Transients NRC Generic Letter 2006-012006-01-20020 January 2006 NRC Generic Letter 2006-01: Steam Generator Tube Integrity and Associated Technical Specifications NRC Generic Letter 1999-021999-08-23023 August 1999 NRC Generic Letter 1999-002: (Errata): Laboratory Testing of Nuclear-Grade Activated Charcoal NRC Generic Letter 1983-111999-06-24024 June 1999 NRC Generic Letter 1983-011, Supplement 1: Licensee Qualification for Performing Safety Analysis ML0823509351999-06-0303 June 1999 Generic Ltr 99-02 to All Holders of OLs for Nuclear Power Reactors,Except Those Who Have Permanenetly Ceased Operations & Certified That Fuel Permanently Removed from Rv Re Laboratory Testing of nuclear-grade Activated Charcoal ML0311101371999-06-0303 June 1999 Withdrawn NRC Administrative Letter 1999-002: Operating Reactor Licensing Action Estimates NRC Generic Letter 1999-011999-05-0303 May 1999 NRC Generic Letter 1999-001: Recent Nuclear Material Safety and Safeguards Decision on Bundling Exempt Quantities NRC Generic Letter 1998-011999-01-14014 January 1999 NRC Generic Letter 1998-001, Supplement 1: Year 2000 Readiness of Computer Systems at Nuclear Power Plants ML0311101601998-08-0303 August 1998 Withdrawn NRC Administrative Letter 1998-005: Availability of Summaries in Electronic Format of Technical Reports by Office for Analysis & Evaluation of Operational Data NRC Generic Letter 1998-041998-07-14014 July 1998 NRC Generic Letter 1998-004: Potential for Degradation of the Emergency Core Cooling System and the Containment Spray System After a Loss-of-coolant Accident Because of Construction and Protective Coating Deficiencies and Foreign Material i NRC Generic Letter 1998-031998-06-22022 June 1998 NRC Generic Letter 1998-003; NMSS Licensees and Certificate Holders Year 2000 Readiness Programs NRC Generic Letter 1998-021998-05-28028 May 1998 NRC Generic Letter 1998-002: Loss of Reactor Coolant Inventory and Associated Potential for Loss of Emergency Mitigation Functions While in a Shutdown Condition NRC Generic Letter 1997-061997-12-30030 December 1997 NRC Generic Letter 1997-006: Degradation of Steam Generator Internals NRC Generic Letter 1997-051997-12-17017 December 1997 NRC Generic Letter 1997-005: Steam Generator Tube Inspection Techniques NRC Generic Letter 1996-061997-11-13013 November 1997 NRC Generic Letter 1996-006, Supplement 1: Assurance of Equipment Operability and Containment Integrity During Design Basis Accident Conditions ML0312007011997-10-0808 October 1997 Withdrawn NRC Generic Letter 1991-018, Revision 1: Information to Licensees Regarding NRC Inspection Manual Section on Resolution of Degraded and Nonconforming Conditions NRC Generic Letter 1997-041997-09-30030 September 1997 NRC Generic Letter 1997-004: NRC Staff Approval for Changes to 10 CFR Part 50, Appendix H, Reactor Vessel Surveillance Specimen Withdrawal Schedules NRC Generic Letter 1997-031997-07-0909 July 1997 NRC Generic Letter 1997-003: Annual Financial Surety Update Requirements for Uranium Recovery Licensees NRC Generic Letter 1997-021997-05-15015 May 1997 NRC Generic Letter 1997-002: Revised Contents of Monthly Operating Report NRC Generic Letter 1995-061997-01-31031 January 1997 NRC Generic Letter 1995-006: Changes in Operator Licensing Program NRC Generic Letter 1996-081996-12-15015 December 1996 NRC Generic Letter 1996-008: Interim Guidance on Transportation of Steam Generators NRC Generic Letter 1996-051996-09-18018 September 1996 NRC Generic Letter 1996-005: Periodic Verification of Design-Basis Capability of Safety-Related Motor-Operated Valves NRC Generic Letter 1996-041996-06-26026 June 1996 NRC Generic Letter 1996-004: Boraflex Degradation in Spent Fuel Pool Storage Racks NRC Generic Letter 1995-091996-04-0505 April 1996 NRC Generic Letter 1995-009: Supplement 1: Monitoring and Training of Shippers and Carriers of Radioactive Materials NRC Generic Letter 1996-021996-02-13013 February 1996 NRC Generic Letter 1996-002: Reconsideration of Nuclear Power Plant Security Requirements Associated with an Internal Threat NRC Generic Letter 1996-031996-01-31031 January 1996 NRC Generic Letter 1996-003: Relocation of the Pressure Temperature Limit Curves & Low Temperature Overpressure Protection System Limits NRC Generic Letter 1989-101996-01-24024 January 1996 NRC Generic Letter 1989-010, Supplement 7: Consideration of Valve Mispositioning in Pressurized-Water Reactors NRC Generic Letter 1996-011996-01-10010 January 1996 NRC Generic Letter 1996-001: Testing of Safety-Related Logic Circuits NRC Generic Letter 1995-101995-12-15015 December 1995 NRC Generic Letter 1995-010: Relocation of Selected Technical Specifications Requirements Related to Instrumentation ML0310701501995-10-31031 October 1995 Withdrawn - NRC Generic Letter 1995-008: 10 CFR 50.54(p) Process for Changes to Security Plans Without Prior NRC Approval NRC Generic Letter 1993-031995-10-20020 October 1995 NRC Generic Letter 1993-003: Verification of Plant Records NRC Generic Letter 1995-071995-08-17017 August 1995 NRC Generic Letter 1995-007: Pressure Locking and Thermal Binding of Safety-Related Power-Operated Gate Valves NRC Generic Letter 1995-031995-04-28028 April 1995 NRC Generic Letter 1995-003: Circumferential Cracking of Steam Generator Tubes NRC Generic Letter 1995-041995-04-28028 April 1995 NRC Generic Letter 1995-004: Final Disposition of the Systematic Evaluation Program Lesson-Learned Issues NRC Generic Letter 1995-021995-04-26026 April 1995 NRC Generic Letter 1995-002: Use of Numarc/Epri Report TR-102348, Guideline on Licensing Digital Upgrades, in Determining the Acceptability of Performing Analog-To-Digital Replacements Under 10CFR 50.59 NRC Generic Letter 1995-011995-01-26026 January 1995 NRC Generic Letter 1995-001: NRC Staff Technical Position on Fire Protection for Fuel Cycle Facilities ML0312004431994-09-0202 September 1994 Withdrawn - NRC Generic Letter 1994-004: Voluntary Reporting of Additional Occupational Radiation Exposure Data NRC Generic Letter 1994-031994-07-25025 July 1994 NRC Generic Letter 1994-003: Intergranular Stress Corrosion Cracking of Core Shrouds in Boiling Water Reactors NRC Generic Letter 1994-021994-07-11011 July 1994 NRC Generic Letter 1994-002: Long-Item Solutions and Upgrade of Interim Operating Recommendations for Thermal Hydraulic Instabilities in Boiling Water Reactors NRC Generic Letter 1994-011994-05-31031 May 1994 NRC Generic Letter 1994-001: Removal of Accelerated Testing and Special Reporting Requirements for Emergency Diesel Generators NRC Generic Letter 1993-081993-12-29029 December 1993 NRC Generic Letter 1993-008: Relocation of Technical Specification Tables of Instrument Response Time Limits NRC Generic Letter 1993-071993-12-28028 December 1993 NRC Generic Letter 1993-007: Modification of Technical Specification Administrative Control Requirements for Emergency & Security Plans NRC Generic Letter 1993-061993-10-25025 October 1993 NRC Generic Letter 1993-006: Research Results on Generic Safety Issue 106, Piping and the Use of Highly Combustible Gases in Vital Areas. NRC Generic Letter 1993-051993-09-27027 September 1993 NRC Generic Letter 1993-005: Line-Item Technical Specifications Improvements to Reduce Surveillance Requirements for Testing During Power Operation NRC Generic Letter 1993-041993-06-21021 June 1993 NRC Generic Letter 1993-004: Rod Control System Failure and Withdrawal of Rod Control Cluster Assemblies, 10 CFR 50.54(f) NRC Generic Letter 1993-021993-03-23023 March 1993 NRC Generic Letter 1993-002: Public Workshop on Commercial Grade Procurement and Dedication NRC Generic Letter 1993-011993-03-0303 March 1993 NRC Generic Letter 1993-001: Emergency Response Data System Test Program 2023-08-03 The following query condition could not be considered due to this wiki's restrictions on query size or depth: <code> [[:Beaver Valley]] OR [[:Millstone]] OR [[:Hatch]] OR [[:Monticello]] OR [[:Calvert Cliffs]] OR [[:Dresden]] OR [[:Davis Besse]] OR [[:Peach Bottom]] OR [[:Browns Ferry]] OR [[:Salem]] OR [[:Oconee]] OR [[:Mcguire]] OR [[:Nine Mile Point]] OR [[:Palisades]] OR [[:Palo Verde]] OR [[:Perry]] OR [[:Indian Point]] OR [[:Fermi]] OR [[:Kewaunee]] OR [[:Catawba]] OR [[:Harris]] OR [[:Wolf Creek]] OR [[:Saint Lucie]] OR [[:Point Beach]] OR [[:Oyster Creek]] OR [[:Watts Bar]] OR [[:Hope Creek]] OR [[:Grand Gulf]] OR [[:Cooper]] OR [[:Sequoyah]] OR [[:Byron]] OR [[:Pilgrim]] OR [[:Arkansas Nuclear]] OR [[:Three Mile Island]] OR [[:Braidwood]] OR [[:Susquehanna]] OR [[:Summer]] OR [[:Prairie Island]] OR [[:Columbia]] OR [[:Seabrook]] OR [[:Brunswick]] OR [[:Surry]] OR [[:Limerick]] OR [[:North Anna]] OR [[:Turkey Point]] OR [[:River Bend]] OR [[:Vermont Yankee]] OR [[:Crystal River]] OR [[:Haddam Neck]] OR [[:Ginna]] OR [[:Diablo Canyon]] OR [[:Callaway]] OR [[:Vogtle]] OR [[:Waterford]] OR [[:Duane Arnold]] OR [[:Farley]] OR [[:Robinson]] OR [[:Clinton]] OR [[:South Texas]] OR [[:San Onofre]] OR [[:Cook]] OR [[:Comanche Peak]] OR [[:Yankee Rowe]] OR [[:Maine Yankee]] OR [[:Quad Cities]] OR [[:Humboldt Bay]] OR [[:La Crosse]] OR [[:Big Rock Point]] OR [[:Rancho Seco]] OR [[:Zion]] OR [[:Midland]] OR [[:Bellefonte]] OR [[:Fort Calhoun]] OR [[:FitzPatrick]] OR [[:McGuire]] OR [[:LaSalle]] OR [[:05000000]] OR [[:Zimmer]] OR [[:Fort Saint Vrain]] OR [[:Shoreham]] OR [[:Satsop]] OR [[:Trojan]] OR [[:Atlantic Nuclear Power Plant]] OR [[:Skagit]] OR [[:Marble Hill]] </code>.
[Table view]The following query condition could not be considered due to this wiki's restrictions on query size or depth: <code> [[:Beaver Valley]] OR [[:Millstone]] OR [[:Hatch]] OR [[:Monticello]] OR [[:Calvert Cliffs]] OR [[:Dresden]] OR [[:Davis Besse]] OR [[:Peach Bottom]] OR [[:Browns Ferry]] OR [[:Salem]] OR [[:Oconee]] OR [[:Mcguire]] OR [[:Nine Mile Point]] OR [[:Palisades]] OR [[:Palo Verde]] OR [[:Perry]] OR [[:Indian Point]] OR [[:Fermi]] OR [[:Kewaunee]] OR [[:Catawba]] OR [[:Harris]] OR [[:Wolf Creek]] OR [[:Saint Lucie]] OR [[:Point Beach]] OR [[:Oyster Creek]] OR [[:Watts Bar]] OR [[:Hope Creek]] OR [[:Grand Gulf]] OR [[:Cooper]] OR [[:Sequoyah]] OR [[:Byron]] OR [[:Pilgrim]] OR [[:Arkansas Nuclear]] OR [[:Three Mile Island]] OR [[:Braidwood]] OR [[:Susquehanna]] OR [[:Summer]] OR [[:Prairie Island]] OR [[:Columbia]] OR [[:Seabrook]] OR [[:Brunswick]] OR [[:Surry]] OR [[:Limerick]] OR [[:North Anna]] OR [[:Turkey Point]] OR [[:River Bend]] OR [[:Vermont Yankee]] OR [[:Crystal River]] OR [[:Haddam Neck]] OR [[:Ginna]] OR [[:Diablo Canyon]] OR [[:Callaway]] OR [[:Vogtle]] OR [[:Waterford]] OR [[:Duane Arnold]] OR [[:Farley]] OR [[:Robinson]] OR [[:Clinton]] OR [[:South Texas]] OR [[:San Onofre]] OR [[:Cook]] OR [[:Comanche Peak]] OR [[:Yankee Rowe]] OR [[:Maine Yankee]] OR [[:Quad Cities]] OR [[:Humboldt Bay]] OR [[:La Crosse]] OR [[:Big Rock Point]] OR [[:Rancho Seco]] OR [[:Zion]] OR [[:Midland]] OR [[:Bellefonte]] OR [[:Fort Calhoun]] OR [[:FitzPatrick]] OR [[:McGuire]] OR [[:LaSalle]] OR [[:05000000]] OR [[:Zimmer]] OR [[:Fort Saint Vrain]] OR [[:Shoreham]] OR [[:Satsop]] OR [[:Trojan]] OR [[:Atlantic Nuclear Power Plant]] OR [[:Skagit]] OR [[:Marble Hill]] </code>. |
yv?,-kAl 1 ,v REG(,4., -al-, 0 (0 UNITED STATES NUCLEAR REGULATORY
COMMISSION
WASHINGTON, 0. C. 20555 April 2, 1984 I j TO ALL POWER REACTOR LICENSEES
AND APPLICANTS
FOR OPERATTNG
LICENSES SUBJECT: CHANGE TO NUREG-1021, "OPERATOR
LICENSING
EXAMINER STANDARDS" (Generic Letter 84-05 )Generic Letter 83-44 notified licensees of the availability of NUREG-1021.
Recently the Nuclear Regulatory Commission has revised NUREG-1021, ES-201, Section H to improve the security of the written operator and senior operator licensing examination administration procedure while naintaining a meaningful review by facility representatives.
A copy of' this change is enclosed for your information and for your use in keeping your copies of NUREG-1021 current.Cormments on NUREG-1021 are welcome and will be considered in future revisions.
Comments should be directed to Nir. Don Beckham, Chief, Operator Licensing Branch, Division of Human Factors Safety.Sincerely, t dt Darrell G. dised h i rector Division of-Licensing Office of Nuclear Reactor Regulation Encl osure:z"'-
Revision 1, ES-Section H 8404020019
840402 PDR ADOCK 05000003 P PDR j /- 1\ ) y/-- f- I & T IZ--~ ) -
FACILITY EXAM REVIEW PROCEDURE AMENDMENT
TO EXAMINER STANDARD ES-201 ES-201, Section H, "Facility Staff Review of Examination" A review of the written examination by facility personnel may be appropriate to ensure that plant specific questions in the examination are correct and up-to-date.
When the Examination Question Bank is operational and the questions have been culled and identified by content area, the examination review may be eliminated.
Until that time, an examination review as described below will be conducted.
The facility review of the examination shall be conducted as follows: No type of facility review of the written examination shall be allowed prior to or while the written examination is in progress.
After all of the candidates have completed the examination and all examination materials and notes have been turned in to the examiner, the Chief Examiner should have knowledgeable member(s)
of the facility staff (training coordinator, operations supervisor, etc.) review the written examination and the answer key to identify any inappropriate questions and to ensure that the questions will elicit the answers in the key. Discussions may be necessary for clarification.
Normally the examiner who prepared the examination should be present throughout the review to explain questions, sources of answers and to ensure that the facility reviewers'
questions about the examination are answered to the extent possible.
The examiner should be capable of providing clarification on examination questions.
Therefore, if the person writing the examination is not available, the other examiners must be certain that they are familiar with the intent of the questions.
A maximum of one facility staff member per section per examination may be present during the review.The review is limited to a maximum of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> (elapsed time). The Chief Examiner may extend this limit only if approved by the appropriate Regional Section Chief or his designee.
All questions and comments made by the facility (other than questions asked to facilitate the review) shall be noted by the examiner.
Although resolution of comments should be reached, if possible, the major emphasis of the review should be to identify all facility concerns rather than to reach agreement on resolution.
After the review, all copies of the examinations and answer keys will be collected by the examiner(s)
and no other comments will be accepted by the examiners(s).
Any additional comments should be provided in writing to the appropriate Regional Branch Chief, not later than five (5) working days following the end of the site visit.
-2-Guidance on conducting the debriefing session (exit interview)
with the facility staff before leaving the site is contained in Standard ES-104, Section B.Prior to grading the examinations, the examiner who conducted the review shall resolve all facility comments, shall correct the examination questions and answer key, if appropriate, and-shall document all facility comments, whether or not he considered them appropriate, and his resolution of the comments.
This documentation, the revised master examination and answer key, and examination results shall be sent to the facility.The examiner shall include on the master copy of the examination the names of the persons who reviewed the examination and answer key. The examiner shall complete appropriate sections of Table ES-201-6.Upon completion of examination grading, the Regional Office shall send an examination report to the utility. The report shall document the examination review meeting with the licensee.
Copies of this report will be sent to PDR's. Copies of examination summary sheets, which are currently provided to utilities pursuant to ES-104, could be enclosed with this letter, but shall be withheld from public disclosure for privacy reasons. A sample examination report is included as Attachment
4 to this standard.
-2-Guidance on conducting the debriefing session (exit interview)
with the facility staff before leaving the site is contained in Standard ES-104, Section B.Prior to grading the examinations, the examiner who conducted the review shall resolve all facility comments, shall correct the examination questions and answer key, if appropriate, and-shall document all facility comments, whether or not he considered them appropriate, and his resolution of the comments.
This documentation, the revised master examination and answer key, and examination results shall be sent to the facility.The examiner shall include on the master copy of the examination the names of the persons who reviewed the examination and answer key. The examiner shall complete appropriate sections of Table ES-201-6.Upon completion of examination grading, the Regional Office shall send an examination report to the utility. The report shall document the examination review meeting with the licensee.
Copies of this report will be sent to PDR's. Copies of examination summary sheets, which are currently provided to utilities pursuant to ES-104, could be enclosed with this letter, but shall be withheld from public disclosure for privacy reasons. A sample examination report is included as Attachment
4 to this standard.
Attachment
4 to ES-201 North Carolina Power Authority ATTN: Mr. H. G. Jones Manager of Power 550A Chesnut Street Anyplace, NC 37401 Gentlemen:
SUBJECT: EXAMINATION
REPORT On December 12-16, 1983, NRC administered examinations to employees of your company who had applied for licenses to operate your Edison Nuclear Power Station. At the conclusion of the examinations, the examination questions and preliminary findings were discussed with those members of your staff identified in the enclosed report.In accordance with 10 CFR 2.7.90(a), a copy of this letter and the enclosure will be placed in NRC's Public Document Room unless you notify this office by telephone within ten days of the date of this letter and submit written application to withhold information contained therein within thirty days of the date of this letter. Such application must be consistent with the requirements of 2.790(b)(1).
Should you have any questions concerning this letter, please contact us.Sincerely, David M. Smith, Chief Project Branch 1 Division of Project and Resident Programs Enclosures:
1. Examination Report 2. Examination(s)
and Answer Key(s) (SRO/RO)cc: Plant Superintendent Plant Training Manager Examiner Enclosure
1 SAMPLE EXAMINATION
REPORT Facility Licensee: North Carolina Power Authority 500A Chesnut Street Anyplace, NC 37401 Facility Docket No.: 50-123 Facility License No.: CPPR-195 Examinations administered at Edison Nuclear Power Station near Spring City, North Carolina Chief Examiner: ,Sami V. -Sm-ith Date Signed Approved by: Frank R. Adams, Section Chief Date Signed Summary Examinations on December 12-16, 1983 Written, oral, and simulator examinations were administered to four SROs, three ROs, and two instructor candidates.
A written examination was administered to one additional RO candidate.
Two SROs, two ROs and one instructor passed these examinations.
All others failed, REPORT DETAILS 1. Persons Examined SRO Candidates W. T. Bounds L. B. Spivey D. E. Huskins J. T. Heck RO Candidates S. T. Allen R. F. Kahle 0. P. Gibson A. F. Sloan Instructor Candidates I. M. Smart P. A. Mills 2. Examiners*S. Y. Smith, NRC J. M. Johnson, EG&G R. F. Radio, EG&G*Chief Examiner 3. Examination Review Meeting At the conclusion of the written examinations, the examiners met with R. P. Johnson, C. L. Boggs and M. E. Peoples of the Training Department to review the written examinations and answer key. As a result of this review, Questions
2.10 and 6.4 of the RO and SRO examinations respectively were deleted. It was determined that although these questions were obtained from facility supplied information, a recent vendor analysis negated the requirement for this system asked for in the questions.
The design change was documented in DCM-83-16.
The facility questioned the applicability of Question 3.3 of the RO examination, but provided no supporting references.
The question was considered appropriate by the staff and retained because the knowledge and skills covered by this question are important to the performance of his job as described in the Job Task analysis.4. Exit Meeting At the conclusion of the site visit the examiners met with representatives of the plant staff to discuss the results of the examinations.
Those individuals who clearly passed the oral and/or simulator examination were identified in this meeting. The examiners made the following observations concerning your training program:
-2-a. Areas of generic weaknesses were found in the use of procedures, radiation protection, and theory, both nuclear and thermodynamic.
The facility committed to place more emphasis in these areas in future training programs (Open Item 84-b. Areas in which the examiners believe that the candidates exhibited good training and knowledge were control room familiarization, instrumentation, and facility administrative procedures.
a I-2-a. Areas of generic weaknesses were found in the use of procedures, radiation protection, and theory, both nuclear and thermodynamic.
The facility committed to place more emphasis in these areas in future training programs (Open Item 84-b. Areas in which the examiners believe that the candidates exhibited good training and knowledge were control room familiarization, instrumentation, and facility administrative procedures.
.;.
- QUESTION DELETED FROM WRITTEN EXAMINATIONS
Question 2.10 a. Describe the accident which the Boron Injection Tank (BIT) is designed to mitigate.(1.0)b. Describe the design features of the BIT, i.e., how does it accomplish its function during an accident situation.
(1.0)Answer 2.10 a. The ECCS including the capability by means of most critical accident in the main steam line BIT provides shutdown boron injection.
The for shutdown capability break.b. The BIT contains a nominal 12 wt.% boric acid and is connected to the discharge of the centrifugal charging pumps. Upon receipt of-an SI signal, the charging pumps provide the pressure to inject the boric solution into the RCS when the isolation valves open.REF: I&E Training Center, Also Edison NPS, STM Systems Manual, Chapter 4.2.13-6.Reason for deletion: Westinghouse Analysis, W-001, provided justification why the BIT was no longer required.
The Tank is still in place, however, it's contents have been replaced with boron at RCS concentration.
Automatic responses to SI signals have been removed (ref: DCM-83-16).
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