ML20148P048
| ML20148P048 | |
| Person / Time | |
|---|---|
| Issue date: | 03/31/1988 |
| From: | Siegel B Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-1231, NUREG-1231-S01, NUREG-1231-S1, NUDOCS 8804080214 | |
| Download: ML20148P048 (126) | |
Text
_ _ _
NUREG-1231 Supplement No.1 1
Safety Evaluation Report related to Babcock & Wilcox Owners Group Plant Reassessment Program @
4: g L
(
.sg t
L q.g.
@M k
O U'
ei 6
C M h;
h D [' $ 93b
~
U.S. Nuclear Regulatory L
i w_
Commission -
e D
Office of Nuclear Reactor Regulation v
.M' March 1988
'v'
,, -l 'Umf,j Q.Lg/
80331 ge0A00ghG PDR Sn
/
l i
NOTICE i
Availability of Reference Materials Cited in NRC Publications Most documents cited in NRC publications will be available from one of the following sources:
- 1. The NRC Public Document Room,1717 H Street, N.W.1 i
Washington, DC 20555
- 2. The Superintendent of Documents, U.S. Government Printing Office, Post Office Box 37082, Washington, DC 20013 7082
- 3. The National Technical Information Service, Springfield, VA 22161 Although the listing that follows represents the majority of documents cited in NRC publications, it is not intended to be ;xhaustive.
j Referenced documents available for inspection and copying for a we from the NRC Public Docu-
)
l ment Room include NRC correspondence and internal NRC memoranda; NRC Office of Inspection and Enforcement bulietins, circulars, information notices, inspection and investigation notices; i
Licensee Event Reports; vendor reports and correspondence; Commission papers; and applicant and i
licensee documents and correspondence.
The following documents in the NUREG series are available for purchase from the GPO Sales 1
Program: formal NRC staff and contractor reports, NRC-sponsored conference proceedings, and NRC booklets and brochures. Also available are Regulatory Guides, NRC regulations in the Code of Federal Regulations, and Nuclear Regulatory Commission Issuances.
Documents available from the National Technical Information Service include NUREG series reports and technical reports prepared by other federal agencies and reports prepared by the Atomic Energy Commission, forerunner agency to the Nuclear Regulatory Commission.
Documents available from public and special technical libraries include all open literature items, such as books, journal and seriodical articles, and transactions. Federal Register notices, federal and state legislation, and congressional reports can usually be obtained from these libraries.
Documents such as theses, dissertations, foreign reports and translations, and non NRC conference proceedings are available for purchase from the organization sponsoring the publication cited, Single copies of NRC draft reports are available free, to the extent of supply, upon written request to thea Division of Information Support Services, Distribution Section, U.S. Nuclear Regulatory Commission, Washington, DC 20555.
Copies of industry codes and standards used in a substantive manner in the NRC regulatory process are maintained at the NRC Library,7920 Norfolk Avenue, Bethesda, Maryland, and are suitable i
there for reference use by the public. Codes and standards are usually copyrighted and may be purchased from the originating organization or, if they are American National Standards, from the American National Standards Institute,1430 Broadway, New York, NY 10018.
a 7*-
ev--
T r>==-m-9v gwyvq+------
---^7--t*
7--'"r--q'y------z w 7-r"97
- ----'--W
'm-'m-
- -#'rT
- w-'rvf=
T--Nwe-*ww-Tyw i t'**'tr4
--t
+-
W+-P*
d'T
P ABSTRACT Supplement 1 to the "Safety Evaluation Report (SER) Related to the Babcock &
Wilcox Owners Group (BW0G's) Plant Reassessment Program'~ has been prepared by the Office of Nuclear Reactor Regulation of the U.S. Nuclear Regulatory Commis-i sion (NRC). This supplement contains the NRC staff's evaluation of the BWOG reassessment of the integrated control system /non-nuclear instrumentation sys-tem, the emergency feedwater initiation and control system, reactor trip ini-tiating events, several additional open items identified in the SER, and the BWOG comments on the SER.
r l
I l
4 l
i 4
NUREG-1231, Supp. 1 iii 4
I T
m.
CONTENTS
.P. ag_e ABSTRACT................................................................
iii 1
INTRODUCTION AND BACKGROUND........................................
1-1 2
OVERVIEW 0F B&W PLANT REASSESSMENT PROGRAM.........................
2-1 2.2 Description of the NRC Staff Activities.......................
2-1 3
SUMMARY
AND CONCLUSIONS...........................................
3-1 4
I N F O RMAT I O N G AT H E R I NG..............................................
4-1 4.5 Staff Activities..............................................
4-1 j
4.5.4 Status of Utility Compliance With NRC Actions..........
4-1 5
SENSITIVITY EVALUATION.............................................
5-1
- 5. 5 Other B&W Plant Sensitivity Issues............................
5-1
- 5. 5. 3 Pressurized Thermal Shock..............................
5-1 I
6 SYSTEMS REVIEW.....................................................
6-1 1
j 6.1 Integrated Control System /Non-Nuclear Instrumentation System Review........................................................
6-1 6.1.1 Introduction...........................................
6-1 6.1.2 Requirements for the ICS/NNI Systems...................
6-5 6.1. 3 ICS Input Signal Failures..............................
6-6 i
6.1.4 ICS Design Features....................................
6-8 6.1. 5 Failure Modes and Effects Analyses.....................
6-11 6.1.6 Loss and Restoration of ICS/NNI Power..................
6-15 6.1.7 ICS/NNI Power Reliability..............................
6-19 4
6.1.8 ICS/NNI Maintenance and Surveillance Testing...........
6-23 6.1. 9 Operator Burden, Procedures and Training...............
6-26 6.1.10 Other ICS/NNI Considerations..........................
6-30 6.1.11 Previously Identified ICS/NNI Concerns arid Recommendations.........................................
6-33 6.1.12 Instrumentation and Control System Issues Not Di rectly Related to ICS/NNI............................
6-35 6.1.13 Conclusions............................................
6-38
- 6. 3 Auxiliary Feedwater/ Emergency Feedwater (AFW/EFW) System Review........................................................
6-41 6.3.3 Acceptability of Emergency Feedwater Initiation and Control System for Each B&W Plant......................
6-41 NUREG-1231, Supp. 1 v
CONTENTS (Continued)
Pate 7
EVALUATION OF HUMAN FACTORS ISSUES.................................
7-1 7.7 Response to BWOG Comments on SER..............................
7-1 9
RISK ASSESSMENT....................................................
9-1 9.6 Risk Recommendations / Relationship to SPIP......................
9-1 9.6.1 Areas Covered bv the BWOG Recommendations..............
9-1 9.6.2 Category C Core Damage Reduction Potential of the BWOG Recommendations........................................
9-1 10 REACTOR TRIP INITIATING EVENTS REVIEW..............................
10-1 10.1 Description of Reactor Trip Initiating Events Review.........
10-1 10.2 Conclusions From BWOG Review.................................
10-2 10.3 BWOG Recommendations and NRC Staff Findings..................
10-3 10.3.1 BWOG Recommendations Related to Reactor Trips Initiated by the Turbine System......................
10-3 10.3.2 BWOG Recommendations Related to Reactor Trips Initiated by the Main Feedwater System...............
10-4 10.3.3 BWOG Recommendations Related to Reactor Trips Initiated by the ICS.................................
10-5 10.3.4 BWOG Recommendations Related to Reactor Trips Initiated by the Control Rod Drive (CRD) System......
10-5 1
10.3.5 BWOG Recommendations Related to Reducing High-and Low-Pressure RCS Trips by Improving Transient Response Capability..................................
10-6 10.4 Summary of Staff Findings....................................
10-8 11 BWOG PROGRAMMATIC AND MANAGEMENT ACTIONS...........................
11-1 11.4 Safety and Performance Recommendation Integration Group......
11-1 11.5 BWOG Evaluation Program To Assess Quality of Implementation of the SPIP Recommendations..................................
11-2 12 IMPLEMENTATION.....................................................
12-1 12.2 NRC Audit of the BWOG Recommendation Approval Process........
12-1 12.3 Matrix of Approved Recommendations...........................
12-1 i
l 12.4 Implementation of Recommendations............................
12-1 1
12.5 Impact Assessment of Proposed BWOG Recommendations...........
12-3 12.5.1 Reduction i n Reactor Trips..........................,
12-3 12.5.2 Challenges to Safety Systems.........................
12-4 12.5.3 Operator Burden......................................
12-5 NUREG-1231, Supp. 1 vi
e CONTENTS (Continued)
P_aSe 12.5.4 Decay Heat Removal...................................
12-5 12.5.5 P re s s u ri zed The rmal Shock............................
12-6 APPENDICES A BIBLIOGRAPHY B ABBREVIATIONS C NRC STAFF CONTRIBUTORS AND CONSULTANTS E PREVIOUSLY IDENTIFIED ICS/NNI CONCERNS F NRC LETTERS TO THE BWOG CONCERNING THE ICS/NNI REASSESSMENT G ERRATA FIGURE 6.1 Major Elements of the BWOG I&C Committee's Reassessment of the ICS/NNI............................................................
6-3 4
TABLES 1.1 Status of BWOG SPIP and NRC Staff Reviews.........................
1-2 9.24 Recommendations That Reduce the Frequency of Core Damage From Category C Events.................................................
9-5 j
12.1 SPIP Recommendations...............................................
12-7 12.2 SPIP Recommendations Rejected by the BWOG Steering Committee.......
12-27 l
l i
i a
l f
-i 4
NUREG-1231, Supp. 1 vii 4
i
1 INTRODUCTION AND BACKGROUND The Nuclear Regulatory Commission staff (referred to as NRC staff or staff) issued a Safety Evaluation Report (SER) related to the Babcock & Wilcox Owners Group (BWOG) Plant Reassessment Program (NUREG-1231, November 1987).
Some of the BWOG submittals to the NRC staff were not received in time for the staff to complete its review of the entire BWOG program before the SER was issued.
This supplement to the SER provides the staff's evaluation of the BWOG reassessment of the integrated control system /non-nuclear instrumentation (ICS/NNI) system, the emergency feedwater initiation and control system, reactor trip initiating events, and several cdditional open items identified in the SER.
The status of the staff's evaluation of the BWOG plant reassessment program is provided in Table 1.1.
As indicated by the table, all the staff's reviews of the BWOG Safety and Performance Improvement Program (SPIP) projects have been completed.
Each section and appendix of this supplement is numbered and titled so that it corresponds to the section or appendix of the SER that is relevant to the NRC staff's additional evaluation.
Except where specifically noted, the material in this supplement does not replace the material in the corresponding SER sec-tion or appendix.
Appendix A is a continuation of the bibliography, Appendix B is a continuation of the list of abbreviations, and Appendix C is a continuation of the list of staff contributors and consultants, as they relate to this supple-ment. Appendix E is a list of some of the more significant previously identi-fied concerns involving the ICS/NNI systems, and Appendix F references NRC staff letters of interest to the ICS/NNI reassessment.
Appendix G contains errata for the SER.
a l
4 i
NUREG-1231, Supp. I 1-1 i
i
Table 1.1 Status of BWOG SPIP and NRC staff reviews SPIP project SER, SSER section Status Sensitivity study SER 5 The BWOG study was submitted in April 1987.
The NRC contractor and NRC staff review is complete.
Operating experience SER 4.2, 4.3, 4.5, The BWOG evaluation of the Tran-review 7.3 sient Assessment Program data is complete.
The NRC contractor and NRC staff review is complete.
Reactor trip initiating SSER 6.1, 10 The BWOG report was submitted events review October 1987.
The NRC staff review is complete.
Risk assessment SER 9 The BWOG evaluation is complete.
SSER 9.6 The NRC staff utilized a contrac-tor to perform an independent risk evaluation.
The NRC contractor and NRC staff review is complete.
Integrated control system /
SSER 6.1 The BWOG evaluation is complete non-nuclear instrumenta-and was submitted in Revision 5 to tion system review BAW-1919 dated July 1987.
The NRC contractor and NRC staff review is complete.
Main feedwater system SER 6.2 The BWOG evaluation is complete.
review The NRC staff review is complete.
Emergency feedwater/
SER 6.3 The NRC staff review was com-auxiliary feedwater SSER 6.3.3 pleted in the SER, except for system review the emergency feedwater initi-ation and control *ystem i
review, Section 6.3 3.
The NRC contractor and HRC staff review of Section 6 3.3 is complete.
l Instrument air system SER 6.5 The BWOG evaluation is complete.
l review The NRC staff review is complete.
Main steam pressure SER 6.4 The BWOG evaluation is complete.
[
control review The NRC staff review is complete.
Operations / maintenance SER 4.4, 7.2 The BWOG evaluation is complete personnel interviews and was submitted in March 1987.
The NRC contractor and the NRC staff review is complete.
L i
NUREG-1231, Supp. 1 1-2
\\
, 'l
-1 Table 1.1 (continued)
SPIPproject SER, SSER section Status Emergency procedures SER 7.4 The BV0G evaluation has not been j
review submitted and will not be com-pleted as part of the SPIP.
The BWOG estimates this effort will i
take until the summer of 1988.
The NRC staff will review these procedures separate from the SPIP.
Operator burden evaluation SER 7.5 The BWOG evaluation is complete and was submitted in May 1987.
The NRC contractor and the NRC staff review is complete.
I i
i i
1 1
I I
l
]
NUREG-1231, Supp. 1 1-3 1
2 OVERVIEW 0F B&W PLANT REASSESSMENT PROGRAM 2.2 Description of the NRC Staff Activities As noted in the SER, an NRC staff member identified nine concerns related to the BWOG reasses; ment program.
Four of these concerns were addressed in the SER and the remaining five were deferred to this supplement.
These concerns are listed below as they appeared in the SER.
(2) There is a potential lack of independence in Science Applications International Corporation performing similar work as a subcentractor for the NRC and for B&W plant owners.
(3) A premature finding by the BWOG of the adequacy of the ICS/NNI.
(4) The BWOG or utilities have not analyzed effectively the proposed SPIP recommendations to determine the effects on the other parts of the plant.
(5) There is a potential for catastrophic reactor pressure vessel failure re-sulting from an overcooling transient assuming one control rod stuck out and a return to criticality.
(7) The B&W plants violate offsite dose limits under the conditions of a steam generator tube rupture design-basis accident.
Concern 2 is related to a potential lack of independence of the work performed by Science Applications International Corporation (SAIC) because they were ne-gotiating a contract with a utility on a similar program while working as a subcontractor for Oak Ridge National Laboratory (ORNL) on an NRC contract with the Office of Nuclear Regulatory Research.
The NRC Office of General Counsel has reviewed this matter and is of the opinion that, assuming the work being performed for the NRC is essentially completed, there is no logical basis for attempting to restrict contractors from performing similar work for others.
With regard to any technical concerns related to the specific ICS/NNI system's j
failure modes and effects analyses, which are contained in Appendix R of BAW-1919, Revision 5, the NRC staff has performed a detailed independent review of this material and its evaluation is contained in Section 6.1.5 of this sup-plement.
The staff, therefore, concludes that this concern does not affect the results of the BWOG SPIP or the staff's review.
3 The staff has addressed concerns 3, 4, and 5 in Sections 6.1.1, 11.5, and 5.5.3 of this suoplement, respectively.
The NRC staff is evaluating the risk associated with steam generator tube events as part of unresolved safety issues A-3, A-4, and A-5 (NUREG-0844).
As a result of these evaluations, the staff has identified several issues for further study that have been designated as generic issue 67 (GI-67), "Steam Generator Staff Actions." The staff member concern 7 is being addressed by GI 67.5.1, "Reassessrent of Radiological Consequences," and GI 67.5.2, "Reevaluation of SGTR Design Basis," and thus will not be addressed in this report.
NUREG-1231, Supp. 1 2-1
3
SUMMARY
AND CONCLUSIONS The summary and conclusions in the SER were primarily based on the results of the NRC staff's review of these areas within the BWOG SPIP that were addressed in the SER.
As noted in the SER, there were three major areas remaining to be reviewed, the integrated control system /non-nuclear instrumentation system (ICS/NNI), the reactor trip initiating events, and the emergency feedweter initiation and control system for each B&W plant.
In addition to these areas, the staff has reviewed the results of the BWOG Safety and Performance Recommendation Integration Group (SPRIG), the BWOG plans for reviewing the implementation of the SPIP recommendations, and the BWOG comments (letter, December 21, 1987) on the initial SER.
The staff also has evaluated the effect of the SPIP recommendations on specific safety issues and overall B&W plant risk.
As noted in the SER, the staff concluded that SPIP was broad-based and compre-hensive.
This conclusion was based, in part, on the continuous input provided by the staff during the development of the SPIP program.
However, since the staff completed its review, it has become apparent that in several areas (such as human factors issues and the ICS/NNI review) the BWOG did not incorporate the staff's suggestions in the SPIP.
The staff continues to believe that SPIP was a comprehensive program and will result in improved safety at B&W plants by reducing the number and complexity of plant trips and transients.
Nevertheless, it would have been more comprehensive if the BWOG had addressed all of the staff's issues.
The staff recognizes that even with implementation of all the I
recommendations proposed by the BWOG, future trant:?nts cannot be totally eliminated and will occur at B&W plants.
Because the program has not fully resolved the staff's concerns, tne staff believes it more probable that these future transients will result from, or be complicated by, the same systems studied in SPIP.
The staff encourages the BWOG to continue to take actions tv a
reduce reactor trips and complex transients at B&W plants and additionally suggests that the NRC staff concerns be considered as a part of these continuing efforts.
As noted in the SER, the risk studies performed by the BWOG and the staff were examined to identify areas that contribute significant fractions of core-damage a
frequency.
From an examination of the SPIP recommendations, it also was con-l cluded that the BWOG studies were responsive to these areas.
The staff has further evaluated the SPIP recommendations versus the results of the risk as-sessmen. studies and concludes that, with proper implementation of the recommen-i dations, the Category C events will be reduced to insignificant contributors to core-damage risk for the B&W plants.
In the SER the staff found that the BWOG efforts in the area of human factors
{
concerns were deficient primarily because human factors engineering expertise were not applied in these efforts.
As a result of this finding, the staff i
suggested that the BWOG re-examine the information gathered from its efforts to provide further assurance that all significant human factors issues have been properly addressed.
The BWOG has concluded that another review of their NUREG-1231, Supp. 1 3-1
efforts is unnecessary.
The staff has reviewed the BWOG position; although no specific regulatory action to require the BWOG to re-review its previous efforts is required, the staff continues to believe that additional recommendations to improve operator performance could be developed from another review of the previous BWOG efforts.
As a result of the specific reviews described in this supplement, the staff finds that the additional recommendations developed as a result of the BWOG review of reactor trip initiating events will further enhance the BWOG efforts to reduce the frequency of reactor trips.
The staff also concludes that, while differences exist in the designs of the emergency feedwater initiating and control systems at the B&W pla.nts, all these systems meet safety grade requirements and satisfy appropriate regulatory re-quirements.
In addition, the staff concludes that all the emergency feedwater systems do not rely on the non-safety grade ICS to fulfill their safety function.
With respect to the BWOG review of the ICS/NNI systems, the staff concludes that the overall approach was appropriate for achieving resolution of ICS/NNI concerns.
The staff considers the development of upgraded / revised ICS/NNI systems requirements documents to be a significant step in the overall effort.
If existing ICS/NNI designs are modified to achieve conformance with the systems requirements, the staff believes that many of the ICS/NNI problems previously identified by the staff from operating experience at B&W plants will be resolved.
i As a result of its efforts, the BWOG has identified numerous recommendations to reduce the contribution of the ICS/NNI systems to reactor trips and complex j
post-trip response.
In general, the staff agrees with the BWOG recommendations.
l However, the staff has identified some recommendations for which it either dis-agrees with the BWOG, has reservations concerning the implementation, or offers additional guidance to the BWOG.
These recommendations are discussed in detail in Section 6.1 of this supplement.
Of special note from the BWOG ICS/NNI review is the BWOG recommendation that B&W plants should go to a known safe state on a loss of ICS/NNI power.
Proper implementation of this rec emendation is expected to eliminate transients similar to those that occurred at Rancho Seco (December 26, 1985) and Crystal River (February 26, 1980).
The staff has provided further guidance on this recommendation, most notably aimed at limiting operator response to loss of ICS/NNI events.
In summary, the staff concludes that proper implementation M the BWOG ICS/NNI recommendations will result in enhanced safety margins at B&W plants by reduc-ing the contribution of the ICS/NNI systems to reactor trips and complex post-trip response.
The staff also concludes that the systems requirements do not fully resolve all previously identified concerns (PICS) with the ICS/NNI.
How-ever, the staff believes that substantial progress has been made toward the t
I resolution of many of these PICS.
As discussed in Scction 6.1, the staff be-lieves that with proper implementation of the recommendations, along with veri-fication of proper implementation of plant modifications made in response to concerns identified in IE Bulletin 79-27, the B&W plant response to ICS/NNI l
NUREG-1231 Supp. 1 3-2
failures will be significantly enhanced and will ensure that adequate safety margins exist for B!.W plants for the long term.
The staff further notes that the BWOG has an ongoing project to evaluate replace-ment of the ICS/NNI with an advanced centrol system based on current technology.
The staff encourages the BW3G to give high priority to this study.
The staff ferther encourages the BWOG to ensure that all the staff's PICS with the ICS/NNI will be resolved by the advanced control system.
In the SER, the staff recommended that the BWOG continue to submit the yearly Operating Experience Report developed under the Transient Assessment Program (TAP) in order to allow the staff to maintain an overview of the effectiveness of the recommendations in improving the performance of the B&W plants.
The BWOG has committed to provide this information.
Additionally, the BWOG has committed to provide the staff with periodic updates to the recommendation tracking system (RTS) to keep the staff informed of progress made in implementing the recommen-dations.
As noted in this supplement, the staff believes that enhancementr, to the RTS may be needed to make this information more useful and will initiate discussions with the BWGG to pursue this further.
The staff has reviewed the efforts of the Safety and Performance Recommendation Integration Group (SPRIG) to identify key recommendations and finds these efforts and the results to be acceptable.
The staff has identified additional recommen-dations that it believes to be "high priority."
Finally, the staff reiterates its position in the SER that to ensure improved performance of the B&W plants and acceptable safety margins in the long term, aggressive implementation of the SPIP recommendations is needed by the utilities.
As discussed in Section 12.4 of this supplement, in the past 6 months little overall program progress by the utilities has been made in implementing and closing recommendations currently in the RTS.
To ensure that the recommenda-tions are aggressively pursued, the staff recommer.ded in the SER that each B&W plant owner submit its plan and schedule for implementing the SPIP recommenda-tions.
The BWOG has noted that most of the recommendations call for further evaluation and thus it may not be possible to develop plans by that time.
In recognition of the BWOG comments, the staff now recommends that each utility submits by June 1, 1988, its plans for those recommendations that have been l
evaluated, its schedule for evaluating the remaining recommendations, and a tentative implementation schedule for implementing the remaining recommenda-tions.
Additionally, the staff believes this information should be perbdic-ally updated as further progress is made in evaluating the recommendations.
The staff will use this information to monitor the utilities' progress and to develop appropriate inspection schedules to ensure that the recommendations have been properly implemented.
NUREG-1231, Supp. 1 3-3 I
4 INFORMATION GATHERING 4.5 Staff Activities 1
4.5.4 Status of Utility Compliance With NRC Actions In Section 4.6 of the SER, the NRC staff stated that it would address the generic l
status of the B&W utilities' co;npliance with NRC actions developed as a result of previous B&W operating events.
To obtain this information, the staff per-6 formed computer searches using the B&W docket numbers and key.words from spec-ific subjects addressed in the BWOG "Safety and Performance Improvement Program" (SPIP) report, BAW-1919, to identify the documents the staff believes should have been reviewed during the SPIP review.
Large, stand-alone program efforts, such as the TMI action plan items, anticipated transients without scram (ATWS),
fire protection, and equipment qualification, are not included because the staff believes they are being adequately covered as separate actions.
However the status of NUREG-0737 action items, including supplemental items, for B&W plants were addressed in Section 4.5.4 of the SER.
The integrated control system /
non-nuclear instrumentation (ICS/NNI) system review included an independent search that rasulted in a table specifically related to that aspect.
This search was used as a basis for the staff's review presented in Section 6.1 of this supplement.
This search resulted in the identification of three additional generic documents that were developed as a result of previous B&W operating events.
Two documents, Inspection and Enforcement Bulletin (IEB) 80-12 and an associated generic letter dated Junc 11, 1980, were issued after an event at Davis-Besse on April 19, 1980, and were related to decay heat removal system operability.
All utilities have completed their actions with respect to this issue.
The third document, Generic Letter 85-13, transmitted NUREG-1154, which is a report on the Davis-Besse loss of main and auxiliary feedwater incident of June 9, 1985.
The issuance of NUREG-1154 prompted the BWOG to form the 1154 Task Force to review the Davis-Besse event of June 9, 1985.
The results of the BWOG review, reported in Appen-i dix H to BAW-1919, Revision 5, were evaluated by the staff and reported in the SER.
On the basis of its review, the staff concludes that the B&W utilities have satisfactorily complied with the previous staff actions developed as a result of previous B&W operating events.
1 l
1 i
l i
i l
NUREG 1231, Supp. 1 4-1 1
v-
~- -
.,m
i 5 SENSITIVITY EVALUATION 5.5 Other B&W Plant Sensitivity Issues 5.5.3 Pressuri;ed Thermal Shock An NRC staff member raised concern, identified as Concern 5 in Section 2.2 of the SER, that for some overcooling transients, assuming a control rod cannot be inserted (stuck out), there is a potential for the reactor to return critical, which could possibly lead to a catastrophic failure of the reactor vessel.
Specifically, this is a pressurized thermal shock scenario wherein the reactor coolant system (RCS) has overcooled, the pressurizer has completely filled with water as a result of the emergency core cooling systems (ECCS) activating, the RCS has pressurized, and the reactor has returned critical because insufficient shutdown margin existed as a result of the assumption of a stuck control rod.
The subsequent pressurization resulting from the reactor returning critical raises the potential for catastrophic failure of the reactor pressure vessel.
To address this concern, the staff requested the BWOG to estimate the tempera-ture at which the reactor will return critical, assuming the most reactive con-trol rod could not be inserted in the core.
The BWOG responded by letter dated i
October 16, 1987.
It stated that by assuming end-of-life core conditions for a generic B&W plant witn all rods inserted in the core, RCS temperature will need to be reduced to less than 35*F for criticality to occur.
Assuming the most reactive control rod stuck out of the core, RCS temperature will need to be reduced to approximately 310 F before the reactor will return critical.
The BWOG also stated that ECCS actuation, which results in boron addition to the RCS, will occur to compensate for the liquid contraction caused by the over-cooling.
Accounting for the negative reactivity of the boron, it was estimated that the RCS temperature again will have to be reduced to less than 35 F before the reactor will return critical.
The NRC staff finds the BWOG estimates of RCS temperatures required for the reactor to return critical to be reasonable.
In light of these low RCS temper-atures, the staff concludes that an overcooling event in a B&W plant will not result in the reactor returning critical, even if the most reactive control could not be inserted in the core.
t l
i NUREG 1231, Supp. 1 5-1 l
2 l
J 6 SYSTEMS REVIEW 6.1 Integrated Control System /Non-Nuclear Instrumentation System Review 6.1.1 Introduction H
The NRC staff stated ir the SER that it had not yet completed its review of the BWOG integrated control system /non-nuclear instrumentation system (ICS/NNI) evaluation report (Appendix R, BAW-1919).
This supplement provides a summary report of the staff's review and supersedes the information presented in the SER.
On December 26, 1985, an event occurred at the B&W-designed Rancho Seco nuclear plant that involved a loss of integrated control system (ICS) 124-volt de power.
The ICS is not a safety-related system.
The loss of ICS de power caused the valves in the main feedwater (MFW) and main steam systems to automatically reposition and caused the loss of remote control of the affected valves from the control room.
In addition, the MFW pump turbines slowed to minimum speed and the auxiliary feedwater (AFW) system pumps automatically started.
The initial effect was an undercooling of the reactor coolant system (RCS) that i
led to a reactor trip on RCS high pressure.
The reactor trip was followed by a safety features actuation system (SFAS) automatic initiation of the high-l pressure injection (HPI) system and excessive overcooling of the RCS.
Investigation of this event at Rancho Seco revealed the same concerns related j
to equipment response to loss of ICS or non-nuclear instrumentation (NNI) power that had been previously identified from earlier events at B&W reactors.
A number of the concerns were documented in IE Bulletin 79-27, "Loss of Non-Class 1E Instrumentation and Control Power S stem Bus During Operation," dated November 30, 1979, and in NUREG-0667, y' Transient Response of Babcock & Wilcox-Designed Reactors," published in May 1980.
The concerns include undesirable failure modes of plant equipment, mid-scale failures of control room indica-tions that can potentially mislead operators, and the absence of a procedure for operator actions upon the loss of ICS power, The staff had intended that l
the concerns associated with loss of ICS/NNI power be resolved through licensee i
4 reviews and responses to IE Bulletin 79-27.
However, as demonstrated by this
)
event, the implementation of modifications by the B&W utilities to resolve the i
concerns identified in IE Bulletin 79-27 were not adequate, i
The objective of the BWOG Safety Performance Improvement Program (SPIP) re-assessment of the ICS/NNI design was to perform a comprehensive review to develop recommended improvements to limit the consequences of failures, and thus, reduce the ICS/NNI contribution to the frequency of reactor trips and the complexity of transients.
The ICS/NNI reassessment and related activities of i
the SPIP were performed by the BWOG Instrumentation and Control (I&C)
Committee.
The BWOG reassessment of ICS/NNI designs is provided in Appendix R, i
"ICS/NNI Evaluation Final Report," to BAW-1919.
l This section of the supplement provides the results of the staff's technical review and evaluation of the BWOG ICS/NNI reassessment.
The staff's review NUREG-1231, Supp. 1 6-1
.- - ~ _ _
i focused on whether the recommendations developed by the BWOG from the reassess-ment would be sufficient to resolve previously identified staff concerns regarding the ICS/NNI, such as those identified in IE Bulletin 79-27 and NUREG-0667.
A list of some of the more significant previously identified con-cerns is provided in Appendix E to this supplement.
The staff believes that resolution of the concerns listed in Appendix E would significantly reduce the complexity of transients at B&W plants.
The staff met periodically with the BWOG I&C Committee during development and execution of the ICS/NNI reassessment program.
On the basis of the information exchanged at the BWOG/NRC meetings and the review of information provided in earlier versions of BAW-1919, the staff, in a number of letters to the BWOG, documented its concerns and provided comments on program direction.
In particu-q lar, these letters (1) documented specific ICS/NNI-related issues that the BWOG J
was requested to address and resolve as part of the reassessment, (2) provided
(
requests for additional information (to be documented in BAW-1919) necessary for the staff to evaluate whether previously identified ICS/NNI concerns were being adequately addressed, and (3) provided general feedback on the scope and direction of the reassessment program.
The letters of interest to the ICS/NNI j
reassessment are referenced in Appendix F to this supplement.
j Information provided in BAW-1919 Appendix I, "Review of Category 'B' and Cate-gory 'C' Events at the BWOG Plants 1980-1985," shows that Category C and significant Category B transients at B&W-designed reactors have resulted from the failure to balance heat removal with heat production under post-trip con-ditions.
Eight of the ten Category C events and twenty-six of the thirty-six i
significant Category B events have resulted from excessive primary-to-secondary heat transfer.
One Category C event and three significant Category B events involved inadequate heat transfer.
Therefore, the BWOG concluded that improved manual and automatic capability for controlling steam flow (steam generator 3
pressure) and feedwater flow (steam generator inventory), and thus steam gener-l 1
ator heat transfer, from the control room under post-trip conditions is the key to reducing the frequency and severity of Category C and Category B transients.
The ICS/NNI systems were selected for review because of their role in controll-j ing the MFW system and secondary side steam pressure relief (turbine bypass valves and atmospheric dump / vent valves) and because of their contribution to j
reactor trips and their involvement in the complexity of pre-trip and post-trip transient behavior.
i i
i Figure 6.1 is a block diagram that identifies the major elements of the BWOG j
I&C Committee's reassessment of the ICS/NNI.
The BWOG I&C Committee first reviewed the operating experience (including information provided in NUREGs, IE bulletins, B&W reports, ongoing BWOG studies, licensee event reports (LERs),
j and other sources), published reports concerning the ICS/NNI (e.g., BAW-1564, i
"Integrated Control System Reliability Analysis"), and the original design requirements and plant operating philosophy for the ICS/NNI systems to develop j
a set of current, updated system design and functional requirements for the
]
ICS/NNI.
The existing as-built, plant-specific ICS/NNI designs were then com-pared to the current requirements for the ICS/NNI systems.
From this compari-
}
son, a list of ICS/NNI prob' ems and potential recommended solutions was j
developed.
The I&C Committee then evaluated the recommended solutions against
}
l 8
1 i
)
NUREG-1231, Supp. 1 6-2
/
j i
,mn
-y.-.
_,,.,.,m..___.,
_,_.,r.-......___._._,,.m.m
,-r,--m---,
,,w,-,.-.,vm,-
c.--
t z
C
- n mca e
DJ f*J W
.DJ sn C
V v.
ca Ir."
"8
--~
gggg.,
/
\\,
~~
n.?.+iE.".~
n.!z..-
iE:
=s=-
=
... =. ~
. 3..
f....
-=
r
- = ;,,.
_ _ N.
g;gg.
\\
1
- =:..
L':l'3:i"*
==ll.
w lll * !.:':.*.
&.*37 C.,::._'f."'
.=
L'3; s
1 x.
08 0 8080 1
f g
/
~ ~ '
s P
/
l
.3.,.,.
N
' g
..,.1j
[
N x
t':V'. -
Figure 6.1 Major elements of the BWOG I&C Committee's reassessment of the ICS/NNI
I 8
the results of the Sensitivity Study performed by MPR Associates (MPR) (dis-cussed in Section 5.1 of the SER), the failure modes and effects analysis (FMEA) performed on the ICS/NNI, the review of the BWOG Transient Assessment Program (TAP) data, and interviews with plant personnel concerning operating procedures and maintenance programs.
On the basis of this evaluation the I&C Committee developed a final list of potential recommended solutions to forward to the SPIP Steering Committee for review.
The recommendations approved by the Steering Committee were issued as final Level I recommendations (i.e., rec-ommendations that will provide immediate improvements to the operation, availability, and reliability of the iCS/NNI) and were included in the SPIP recommendation tracking system (RTS) 'or implementation at BWOG plants.
In Section 2.2 of the SER, an NRC staff member raised concerns related to the reassessment program.
Concern 3 stated that the BWOG had reached a premature finding with regard to the adequacy of the ICS/NNI.
Specifically, the BWOG assumed the ICS/NNI systems to be adequate; thus the ICS/NNI reassessment did not consider replacement of the existing ICS/NNI with an advanced design.
The staff believes that the existing ICS/NNI design can be modified to resolve the concerns listed in Appendix E to this supplement and, therefore, finds the BWOG i
approach acceptable.
In addition, the staff notes that the BWOG Steering Com-mittee directed the I&C Committee to evaluate replacement of the ICS/NNI with an advanced system.
The BWOG took the approach to make immediate improvements i
in the existing ICS/NNI design while replacement of the ICS/NNI design was being studied under the equipment obsolescence program.
By letter dated October 7, 1987, the BWOG provided an updated RTS report that lists the most recent approved recommendations.
There are 49 approved recom-mendations related to the ICS/NNI.
Approved recommendations are identified by a "TR" prefix to the recommendation number.
Revision 5 of BAW-1919 (dated July 1987) lists 31 pending Level I recommendations that will be entered in the RTS if approved by the Steering Committee.
The pending Level I recommendations are i
listed in Section VII.B.2 of Appendix R to bah-1919 and have been designated by the staff as recommendations B.2.1 through B.2.31 for reference purposes.
Section VII.C of Appendix R lists 10 Level 11 recommendations; that is, rec-l ommendations potentially involving major modifications to existing equipment l
and requiring further in-depth evaluation before BWOG approval.
These rec-ommendations have been referenced by the staff as C.1 through C.10.
Section VII.D of Appendix R lists six level III recommendations; that is, recommenda-tions involving replacement of the existing ICS/NNI system with a new system based on modern digital control technology.
These recommendations have been referenced by the staff as 0.1 through D.6.
Section VII.E of Appendix R 4
identified 26 additional potential recommendations undergoing additional review by the BWOG I&C Committee; these have been referenced by the staff as E.1 through E.26.
The staff also has reviewed the ICS/NNI-related recommendations that were rejected by the BWOG Steering Committee. These rejected recommenda-I tions are identified by an "RR" prefix to the recommendation number.
In addi-tion, the staff reviewed ICS/NNI-related recommendations made by the BWOG Transient Assessment Committee in its report, "Review of Reactor Trip Initiat-ing Events at the BWOG Plants 1980-1986"; these recommendations are identified by a "TAC" prefix to the recommendation number.
i NUREG-1231, Supp. 1 6-4
}
~. - _ -. -
The sections below provide the staff's evaluation of all ICS/NNI-related rec-ommendations.
The staff discusses those areas of concern and additional considerations or actions that it believes are desirable before implementation of specific recommendations and those recommendations that it believes should not be implemented.
The recommendations are evaluated on the basis of their potential contribution toward (1) reducing ICS/NNI involvement in reactor trip initiation and in transient complexity and (2) in achieving resolution of previously identified concerns that have been common to the more severe B&W plant transients involving ICS/NNI failures.
Because of the large number of ICS/NNI-related concerns and recommendations, the complexity of the ICS/NNI systems, and the large scope of the BWOG ICS/NNI reassessment (which involved failure modes and effects analyses, an independent sensitivity study, and numerous other subprograms), the evaluation provided below is a summary / overview of the staff's conclurions related to the BWOG ICS/NNI reassessment.
A staff report discussing the more detailed technical aspects of the ICS/NNI reassessment program and recommendations is being pre-pared and will be issued separately.
6.1.2 Requirements for the ICS/NNI Systems Requirements of the ICS/NNI systems (collectively referred to as system require-ments) were developed as part of the ICS/NNI reassessment. The system require-ments include functional requirements, hardware design requirements, and pro-grammatic requirements (e.g., preventive maintenance and qualification and training of personnel).
The BWOG I&C Committee developed recommendations for ICS/NNI improvement by comparing actual installed plant-specific designs with the ICS/NNI systems requirements.
The staff reviewed the ICS/NNI systems requirements to determine whether they were sufficient to produce recommenda-tions that, if implemented, would prevent or reduce the consequences of B&W plant transients such as occurred at Rancho Seco on December 26, 1985.
j In general, the system requirements established the bases that the designers would have used in the original ICS/NNI systems had they had the benefit of the actual operating experience with regard to ICS/NNI failure modes and their effect on overall plant response.
As mentioned earlier, the current require-ments for the ICS/NNI systems were developed after a review of operating experience, the original design requirements for the ICS/NNI systems, and current plant operating philosophy.
The contents of the system requirements documents fer both the ICS and NNI are similar, each with sections addressing the following topics:
envirnnmental requirements, power supply design, loss and restoration of power, signal input reliability, pneumatic design, instru-I mentation accuracies, interfaces with other systems, general / functional re-quirements, instrumentation calibration, training / qualification, system performance, and requirements for maintaining up-to-date system documentation (e.g., schematic diagrams and instruction books).
From the staff's review of the ICS/NNI system requirements, it appears that the BWOG ICS/NNI design philosophy is (1) to improve system reliability by minimiz-ing the effects of single instrument channel component failures (e.g., trans-mitter failures) and (2) to limit the consequences of ICS/NNI power losses by ensuring that equipment failure positions and plant response are such that the NUREG-1231, Supp. 1 6-5
1 2
plant will assume a "known safe state." The BWOG has defined "known safe state" as maintaining the reactor coolant system pressure / temperature relation-ship within the bounds of the abnormal transient operational guidelines (AT0G) i normal post-trip pressure-temperature window and allowing credit for normal operator actions from the control room upon loss of ICS/NNI power.
Other sys-tem requirements that affect ICS/NNI hardware design are intended to improve the operator's capability to respond to loss of ICS/NNI power events, and to prevent plant transients when power is restored.
The staff's review of the ICS/NNI system requirements indicates that many of the problem areas identified from the investigations of B&W reactor transients involving the ICS/NNI would be resolved if existing plant ICS/NNI designs were modified to achieve conformance with the system requirements.
However, the staff also believes that a number of areas need to be further addressed in the documents related to ICS/NNI system requirements.
The staff views such docu-ments as "living documents" that will be retained by the BWOG for reference with regard to the design and operation of the ICS/NNI systems.
The staff recommends that the BWOG consider developing ICS/NNI design-basis documents that explain the basis for the individual requirements for the ICS/NNI systems and provide justification for values established by the system requirements.
These documents should be retained as living documents to support after documents related to ICS/NNI system requirements.
In the staff's opinion, the design-basis documents will significantly enhance any explanation regarding the basis for the existing ICS/NNI designs and will ensure proper development and implementation of any future ICS/NNI system modifications.
6.1.3 ICS Input Signal Failures The results of the staff's technical review and evaluation of the BWOG rec-ommendations related to improving ICS/NNI response to input failures are given below.
Specifically included is the staff's review of BAW-1919, Appendix D, "Improvement of ICS Response to Input Failures."
The following recommenda-tions were reviewed:
Number Recommendation TR-001-ICS*
Replace RC flow signal input to ICS with RC pump status.
l TR-003-ICS Remove startup FW flow correction to main FW flow function from the ICS.
TR-005-ICS*
Remove neutron flux signal auctioneering circuitry from RPS and relocate in the ICS.
TR-006-ICS Delete FW temperature correction to FW demand from ICS.
TR-104-ICS*
Incorporate automatic selection of valid inputs for ICS/NNI.
(Note:
This recommendation supersedes TR-002-ICS and 1
TR-004-ICS.)
- The BWOG has identified these recommendations as key recommendations.
h NUREG-1231, Supp. 1 6-6 4
l L
Number Recommendation TR-154-ICS*
Provide operator with unambiguous status of indicators and recorders in main control room on loss of ICS/NNI power or signal.
TR-189-ICS Set selector switches to select maximum NNI-X dependence.
l TR-196-ICS Set pressurizer level signal select relays to automatic powered transmitters.
TR-198-ICS Automatic powered reactor inlet and outlet temperature sensors should be selected, or logic changed to automatically select auto powered sensor on loss of hand power.
B.2.8 Spurious alarms should be suppressed.
B.2.10.4 All other signal-select hand stations should normally be set to select auto powered sensors.
B.2.12 "Pure" NNI-X and NNI-Y indicator and recorder channels should be developed.
B.2.13 Auctioneering circuit reliability and failure modes should be considered for improved input signal selection.
B.2.14 Maximize Jependence of hand selectable ICS/NNI input signals on only one NNI ac power source.
0.8 Eliminate mid-scale failures that can affect indication er plant control.
E.15 Develop NNI-Y Teold and T,y, signal.
TAC-3.d*
Expand TR-104-ICS to include all ICS input signals.
i The staff agrees with and encourages implementation of the above recommenda-j tions with the exception of recommendations B.2.8, TR-189-ICS, and B.2.14; the staff has reservations about recommendations TR-001-ICS and TR-006-ICS.
The staff believes that recommendation B.2.8 (which involves changing alarm logic from deenergize-to-actuate to energize-to-actuate) is not app,'opriate in all cases; for example, failures of signal monitors and associated relays would no longer result in an alarm alertir.3 the plant operators to the failure.
Reducing nuisance alarms is a good objective; however, the advari. ages and dis-advantages of this recommendation should be further evaluated for each alarm.
Recommendation B.2.8 is discussed further in Section 6.1.6.
The staff does not believe sufficient justification has been provided to support recommendations TR-189-ICS or B.2.14, which maximize dependence on NNI-X versus NNI-Y, thus potentially increasing the complexity of a loss of NNI-X power transient in order to make a loss of NNI-Y power transient more tolerable.
- The BWOG has identified these recommendations as key recommendations.
NUREG-1231, Supp. 1 6-7 i
The staff has reservations concerning recommendations TR-001-ICS and TR-006-ICS.
The staff agrees with the intent of TR-C01-ICS, which is to reduce the number of reactor trips resulting from RC flow signal failures.
However, the staff is
{
concerned about the replacement of the flow signal with an "equivalent" signal
)
based on RC pump breaker status.
It is not clear that ICS performance with the l
equivalent signal will be acceptable for losses of RC flow that are not caused l
by RC pump trips.
Another concern is that one of the proposed designs to pro-vide an equivalent RC flow signal involves replacement of the equivalent signal with the "less reliable" actual flow signal during transient conditions.
It also appears to the staff that the ICS unit load demand (ULD) runbacks on RC pump count and RC flow are not totally redundant; in which case the RC flow runback should not be deleted.
The staff believes that hardening of the RC flow signal may be preferable to replacing it with an equivalent.
BAW-1919 does not provide adequate justification for deleting the FW temperature correc-tion circuit as recommended in TR-006-ICS.
The staff believes that when the startup feedwater flow signal is removed from I
the ICS (recommendation TR-003-ICS), the startup feedwater flow indication in
]
the main control room should be retained for use by the operators.
When the neutron flux signal auctioneering circuitry is relocated from the RPS to the ICS (recommendation TR-005-ICS), the BWOG should verify that the modified designs conform to the requirements--concerning control and protection system interaction--of Section 4.7 of Institute of Electrical and Electronics Engi-neers (IEEE) Standard 279, "Criteria for Protection Systems for Nuclear Power Generating Stations." The staff agrees with the concept of the smart analog signal selector (SASS) units (recommendation TR-104-ICS) and concludes that alarms should be provided to alert the operators to SASS operation.
The information provided in BAW-1919 does not make it clear as to whether such l
alarms are provided.
l The staff considers recommendations TR-154-ICS and C.8 to be key recommenda-tions because of their importance with regard to resolving previously identi-fied ICS/NNI concerns and strongly encourages their implementation.
6.1.4 ICS Design Features The results of the staff's technical review and evaluation of the BWOG rec-ommendations related to specific ICS design features are given below.
Included is the staff's review of BAW-1919, Appendix C, "Final Report on Re-evaluation of ICS Design Featuru " The following recommendations were reviewed:
Mumber Recommendation TR-007-ICS*
Remove British thermal unit (Btu) limits from ICS.
TR-008-ICS*
Improvements to reactor runback capability.
TR-009-ICS*
Improvements in ICS tune control circuits.
l TR-010-ICS ICS control circuit modification.
- The BWOG has identified these recommendations as key recommendations.
NUREG-1231, Supp. 1 6-8
l l
j 4
i Number Recommendation i
d TR-011-ICS Determine if the grid frequency error circuit has been detuned, f
j TR-106-ICS Remove unused hardware from ICS/NNI cabinets.
t i
TR-190-ICS Develop backup controls for pressurizer level and pressure control.
i TR-193-ICS Review / test pressurizer heater low-low level interlock logic.
TR-194-ICS Buffer hand powered indicators and recorder inputs from auto-(
matic power signals.
f Failure of inputs to RC pump interlocks must not prevent pump TR-199-ICS restart.
TR-204-ICS Eliminate or reduce automatic ICS runback rate on asymmetric rod conditions.
TR-209-ICS Add signal limiters to prevent control integrals from going into saturation.
C.1 Provide independent controls for reactor power, steam demand, t
and feedwater such that no single failure would lead to simul-taneous and spurious manipulation of control rods, turbine steam admission valves, and feedwater flow.
C.2 Separate control of feedwater pump speed from feedwater regulat-ing valve (except CR & ANO-1).
C.3 Separate control of each feedwater pump speed control from each J
other.
4 j
C.4 Separate control of each feedwater regulating valve from each l
other.
/'
i j
C.5 Elimination of "Delta Tc" controller function in the ICS.
l C.6 Control of feed pump speed based on pumps discharge pressure rather than feedwater valve delta P.
I C7 Provide separate subsystems for reactor coolant temperature and steam pressure control.
4 l
E.14 Remove automatic control functions for the atmospheric dump valves from the ICS/NNI system.
E.22 Provide redundant, 2 out of 2 logic, PORV control, separate
{
from NNI.
Evtluate leaving the PORV block valve normally
- closed, i
1 l
NUREG-1231, Supp. 1 6-9 1
The staff agrees with and encourages implementation of the above recommenda-tions with the exception of r(commendations TR-007-ICS and C.S.
The staff has reservations about recommendations TR-204-ICS, E.22, TR-010-ICS, and C.4.
Recommendation TR-007-ICS involves removal of the Btu limits function from the ICS and the addition of a means to run back main feedwater flow following a reactor trip during situations where the ICS cross-limits function is rendered inactive.
The Btt limits alarm function would remain active to alert the operators of the n0ed for manual action in the event of an overfeed / overfill condition.
Originally it was thought that removal of the Btu limits would preclude reactor trips caused by the failure of input signals to the Btu limits circuits and improve the overall reliability of the ICS by reducing the number of ICS modules, thus simplifying the ICS.
However, because of the hardening of the T input signal to the Btu limits that would occur with recommendation hot TR-104-ICS (discussed in Section 6.1.3), the staff believes that removal of the Btu limits would not significantly reduce reactor trip frequency.
In addition, while B&W plants rely on the cross-limits function to prevent overcooling of the RCS resulting from excessive MFW flow following a reactor trip, the Btu limits serve as a backup to the cross-limits function when the cross-limits function is not active.
The staff is not convinced that the addition of a rapid feedwater reduction circuit will adequately compensate for removal of the Stu limits function and believes that modifications to automatically switch the rod controls from manual back to automatic to restore the cross-limits function on a reactor trip could increase transient complexity and operator burden as well as possibly introduce potential undesirable failure modes Thus, the staff believes that the BWOG Steering Committee should reconsider its accept-ance of this recommendation.
Recommendation C.5 involves the elimination of the delta T function from cold the ICS; this function is used to control temperature differences between the two reactor coolant piping system cold legs.
Since insufficient justification was provided for removal of this function, the staff does not concur with re-moval of the delta T controls from the ICS as proposed in recommendation C.5.
cold The staff has reservations concerning recommendations TR-204-ICS and E.22.
The staff agrees with reducing the automatic ICS runback rate for asymmetric rod conditions to prevent unnecessary reactor trips; however, BAW-1919 does not provide sufficient justification to eliminate the asymmetric rod runback func-tion from the ICS.
E.22 recommends that plant operation with the PORV block valve normally closed be evaluated.
It is not clear to the staff that plant operation with the block valve closed would be an enhancement to plant safety.
This recommendation shuuld be carefully evaluated from the standpoint of transients that could challenge pressurizer safety valves and availability of the PORV as an alternate means for depressurization and heat removal.
)
Recommendation TR-010-ICS proposes ICS modifications to eliminate problems experienced when controlling T,y, with the feedwater control system, including oscillations around the T,y, set point and upsets in feecwater flow when return-
)
ing control of T,y, from the feedwater control portion of the ICS back to the reactor (control rod) control portion.
The BWOG recommends a pilot program to i
install a new conceptual design to prevent these problems on an operating reac-tor.
However, in BAW-1919 the BWOG indicates that because of the limitations NUREG-1231, Supp. 1 6-10
[
of the simulators used, the conceptual control scheme was not evaluated to confirm its effectiveness. While the staff believes that improved control is desirable, the new controls should be carefully analyzed before installation at an operating plant.
Recommendation C.4 is to separate control of each feed-water regulation valve from each other.
It is not clear to the staff what is specifically intended by this recommendation, i
The staff considers recommendation TR-190-ICS to be a key recommendation be-cause of its importance with regard to resolving previously identified ICS/NNI i
concerns; the staff strongly encourages its implementation. - The staff also strongly encourages implementation of recommendation TR-194-ICS and believes that recommendations C.1 and C 7 have significant potential benefit in develop-ing a control system design that is more tolerant to failures; therefore, this too should be actively pursued by the BWOG.
6.1. 5 Failure Mcdes and Effects Analyses During the BWOG ICS/NNI reassessment effort, updated failure modes and effects analyses (FMEA) of the ICS/NNI systems were performed and the results of these studies were used for identifying and evaluating potential ICS/NNI improve-ments.
The BWOG used the FMEA as a tool in the development and validation of recommendations for ICS/NNI improvements that, when implemented, will result in a reduction in transients either initiated or complicated by selected classes of failures in the ICS/NNI.
The scope of these FMEA considered power supply failures, input signal failures, and output signal failures, since these classes of failures represent most of the failures that have occurred in ICS/NNI systems.
The recommendations developed from these FMEA can be correlated to recommenda-tions in the RTS or recommendations pending approval by one or more BWOG com-i mittee.
Since the corresponding BWOG recomuendations are addressed in other sections of this supplement, the validity of the recommendations resulting from FMEA is not addressed in this section.
Consequently, this section deals specifically with the scope of these FMEA and the models used to perform them, t
Because the adequacy of the model significantly influences the adequacy of the analysis of simulated system performance, the staff evaluated the FMEA models and scope to ensure that the results of the FMEA were adequate for validating existing BWOG recommendations.
Additionally, it needed to determine whether i
other classes of failures had been overlooked, thereby resulting in the omission of additional recommendations that could further reduce the occurrence of plant transients.
If the models and scope of the FMEA proved to be adequate, the staff could conclude that the FMEA results were applicable to the analyses goals.
If there are deficiencies in the models and scope, then the impact of those deficiencies must be evaluated with regard to the goals of the analyses.
Since some r,taff concerns regarding FMEA scope are based on the results of earlier analyses performed on the ICS/NNI systems, these FMEA will be discussed in the order of their development.
Following the TMI-2 accident in 1979, the NRC determined that certain aspects of the B&W plant design needed additional assessment.
Consequently, in May 1979, the NRC issued a shutdown order to all B&W-designed plants.
The order 1
NUREG-1231, Supp. 1 6-11
i directed that the licensees analyze the role of control systems failures and their significance to safety, f
B&W, in response to the NRC order and on behalf of the BWOG, performed a FMEA of the ICS (BAW-1564) to identify sources of transients initiated by the ICS and to define potential areas for improvement to reduce the frequency of tran-sients.
The FMEA also was used to determine whether an ICS failure could cause a failure mode whereby plant safety systems would not protect the reactor core.
The emphasis was on analyzing ICS failures that could affect the feed-water system, emergency feedwater system, and pressurizer pressure and level controls and challenge the pilot-operated relief valves (PORVs), safety valves, and RPS/ESFAS.
The staff's review of the B&W FMEA led to the following conclusions:
(1) The significance of the ICS with regard to plant safety systems was not adequately addressed.
(2) A fault tree for loss of feedwater should have been developed on the basis of equipment diagrams rather than functional blocks.
(3) The FMEA should have included other systems with which the ICS interacts, such as NNI and power sources.
I (4) Dower supply failures should have been evaluated in detail, tnd specific l
recommendations concerning power supply reliability should have been developed.
(5) The simulation tools used in the FMEA were deficient in their dynamic range and component details; however, the deficiencies did not greatly affect the overall results, since a reactor trip was the terminating point i
for the simulations.
If more detailed evaluations of ICS/NNI failure l
modes are performed, more sophisticated system simulation tools should be i
used.
l (6) Improvement is needed in areas of ICS/NNI system arrangement, channeling,
{
and selection of input signals.
i (7) Since there is a tight coupling between the secondary system--which is controlled by the ICS--and the primary system, dynamic performance should be studied, including the effects of control limitations on plant stability.
l These comments were forwarded to the BWOG, but further analyses were not per-formed to specifically address these comments.
1 Following a series of abnormal events involving non-Class-1E power supply failures at several B&W-designed plants between 1978 and 1980, the NRC issued IE Bulletin 79-27, which included requirements for all operating nuclear i
power facilities to review Class 1E and non-Class 1E buses supplying power to safety-and non-safety-related instrumentation that could affect the ability to achieve a cold shutdown using existing procedures.
A FMEA of the power supplies to the Oconee Unit 1 ICS and NNI instrumentation systems was performed for the NRC by Science Applications International Corporation (SAIC).
A NUREG-1231, Supp. 1 6-12
detailed FMEA model was developed on the basis of equipment diagrams rather than functional modules.
While the model included the NNI and power supplies, it did not include the plant safety systems.
Deficiencies in the scope of the Oconee-1 FMEA led the staff to conclude that additional recommendations regarding the Oconee-1 ICS/NNI would be forthcoming if the FMEA model were updated to reflect the staff concerns.
These staff concerns are (1) lack of dynamic range in the simulations (2) missing power supply branches in the system models (3) no consideration of the effects of individual module failures (4) omission of safety system interactions (5) no consideration of failures external to the ICS/NNI that could affect ICS/NNI responses The staff further recognized that many of its comments described in its review of BAW-1564 were not applied to this FMEA.
Therefore, the staff concludes that the scope of this FMEA was marginally adequate and the results must be care-fully evaluated for applicability before being extended to other Model 721 ICS/NNI systems.
As part of the ICS/NNI reassessment, the BWOG performed FMEA of the ICS/NNI systems in use at Davis-Besse and Three Mile Island Unit 1 (THI-1).
The Davis-Besse FMEA was then used as the base FMEA for the analyses performed at Rancho Seco Unit 1, Crystal River Unit 3, and Arkansas Nuclear One Unit 1 (ANO-1).
Additionally, the results of the Oconee-1 FMEA were incorporated into the TMI-1 FMEA.
Significant failure modes discovered by these FMEA were then used by B&W as initial conditions to simulate plant responses.
The BWOG performed the FMEA assuming the plant was operating at full power.
The staff believes the FMEA scope should have included other plant operating states.
The 100 percent full power operating state was selected because the BWOG concluded that transients caused by single ICS/NNI failures at this power level would bound all other classes of ICS/NNI failure modes.
The staff, on the other hand, concludes that there are other operating stat h, such as single feedwater pump operation and operating with less than four reactor coolant pumps, that may lead to other significant plant transient responses that would not be revealed by analyzing only the full power cases.
Operator responses were considered only from the perspective of initial indica-tions in the control room.
Incorrect operator responses were not simulated because the BWOG considered the problem of simulating the wide range of ptten-tial failures outside the scope of the FMEA program.
Operator errors of omission or commission are difficult to analyze because of the wide variation in potential actions.
Nevertheless, the FMEA scope should have considered operator errors of conservative magnitude, such as errors of failing to perform a prescribed function or errors caused by performing the most likely incorrect function for a given scenario.
The fault tree methodology could have been employed to assist in the analysis of this class of transients.
Within the scope of the BW0G FMEA program and model limitations, the. MEA addressed loss of power to individual loads and individual buses (at all NUREG-1231, Supp. 1 6-13
_- = _ - _.
3 I
i voltage levels) regardless of whether multiple power sources were provided through auctioneering or automatic bus transfer devices.
The staff believes 1
this is a significant improvement over the original FMEA (BAW-1564).
The scope of the FMEA was adequate for addressing s^,aff concerns regarding overfill failure modes and effects, within the power level constraint imposed upon the FMEA simulation tools (100 percent power with all controls in their optimum positions).
For certain failure scenarios, the Davis-Besse FMEA showed i
activation of the Stu limiting circuits to help mitigate the overfill condition.
Both Crystal River 3 and ANO-1, which used the Davis-Besse FMEA as the base FMEA, do not have active Btu limiting circuits.
While the BWOG noted this deficiency in the FMEA for these plants, no additional analyses were performed.
The staff believes the BWOG should consider performing additional analyses for l
i these plants to ensure that adequate overfill protection is provided.
These FMEA did not include safety system interactions (such as ECCS or ESFAS interactions).
However, the staff believes that the safety systems are ade-quate to protect the plant from ICS/NNI failures.
Thus, the fact that these FMEA did not model safety system interactions is not considered to be a sig-nificant deficiency.
The FMEA scope did not include ICS/NNI system responses to failures external to the ICS/NNI, such as in the instrument air system.
Rather, the FMEA scope focused on how ICS/NNI failures affect plant response.
Since there is a tight coupling between the secondary system and the primary system, dynamic system performance should have been included in the scope of these FMEA, including the effects of control limitations on plant stability.
The BWOG appears to be j
assuming that the ICS/NNI will respond favorably in all cases where the ini-I tiating failure is not in the ICS/NNI hardware.
This has not been verified by the FMEA.
1 A limited number of the FMEA recommendations could be used to validate some of the system requirements; however, the system requircments were general, while the recommendations were more succinct; consecuently, this comparison effort could not be performed adequately.
These FV.A identified problems other 3
than loss of ICS/NNI power and indicated that mo e detailed systems require-l ments may be necessary.
/
The staff finds that these FMEA were specifically constructed to examine fail-ures, including power supply failures, which were representative of the types of ICS/NNI failures that have been experienced.
The staff also concludes that the results of these FMEA validate many of the SPIP recommendations.
- However, because simulation was limited to only 100 percent power operation, the com-pleteness of the BWOG efforts cannot be determined.
In light of the staff's findings on the FMEA methodology, it is likely that there are additional ICS/NNI failure modes that have not been discovered or t
addressed by the SPIP recommendations.
These failure modes may result in future B&W plant reactor trips and/or Category B and C transients.
However, it is the staff's opinion that the B&W plant safety systems are capable of miti-gating such events.
Thus, the limitations of the FMEA are not believed to be safety significant.
Nevertheless, to ensure that the frequency of complex 1
i NUREG-1231, Supp. 1 6-14 I
i
_=
4 7
plant transients is reduced at B&W plants, the staff encourages the B&WOG to
[
j consider the staff's comments on the FMEA program and to continue its investi-j j
gation of the plant response to ICS/NNI failures.
6.1.6 Loss and Restoration of ICS/NNI Power i
i i
The results of the staff's technical review and evaluation of the BWOG recom-mendations related to the effects of partial and total loss of ICS/NNI ac and/
or de power on plant equipment and control room instrumentation and controls and the subsequent restoration of power are given below.
The following rec-ommendations were reviewed:
i Number Recommendation I
TR-012-ICS*
Determine if operator has necessary information from procedures, indicators, etc. to detect loss of NNI power.
4 TR-032-ICS*
Evaluate restoration of ICS/NNI power.
TR-033-ICS*
Ensure that plant will go to a safe state on loss of ICS/NNI power.
l TR-036-ICS Evaluate turbine bypass valve position on loss of ICS.
I i
TR-037-ICS*
Evaluate MFW pump speed control on loss of ICS power.
1 l!
TR-062-0PS Maintain a high SPOS availability by corrective and preventive maintenance.
1 TR-096-MSS
- Evaluate design of turbine bypass and atmospheric dump systems.
t 4
TR-097-EFW*
Evaluate design of EFW flow control valves.
}
[
TR-154-ICS*
Provide operator with unambiguous status of indicators and 2
I recorders in main control room on loss of ICS/NNI power or signal.
)
TR-158-OPS Re-evaluate annunciator designs to ensure key alarms do not go unnoticed.
1 TR-159-0PS*
Evaluate secondary system controls to achieve remote manual i
control in the main control room of all post-trip steam flow paths, MFW and EFW.
i TR-167-PES Include in operating procedures guidance on restoration of power i
to electrical buses, especially if the ICS or ICS controlled j
equipment is affected.
4
)
TR-172-PRV Evaluate PORV circuitry to determine if momentary lo',s of power or restoration of power can cause PORV to open.
4
- The BWOG has identified these recommendations as key recommendations.
4 l
NUREG-1231, Supp. 1 6-15 i
a
Number Recommendation TR-178-ICS*
Ensure plant goes to a known safe state on loss of power to the ICS/NNI.
(Note:
this recommendation supersedes TR-033-ICS.)
TR-181-0PS Verify adequacy of instrumentation and displays used to assess and control the ATOG stability parameters.
TR-185-ICS Power feedwater flow recorders directly from NNI.
TR-190-ICS Develop backup controls for pressurizer level and pressure control.
TR-191-ICS Separate condensate flow from NNI power (for Crystal River only).
TR-195-ICS Supply hand and automatic power circuits from separate panels.
TR-197-ICS Provide automatic power transfer for the modulating pressurizer heater E/I converters.
TR-211-ICS Develop modification to remove automatic ICS trip on NNI single power failure.
B.2.2 Power automatic (branch H) and hand (branch HY) power from dif-ferent panels.
B.2.4 Indicate loss of ac power on load side of ICS automatic bus transfer (ABT) to control room operators.
B.2.7 Automatic trip of feedwater pump,.
B.2.8 Suppression of spurious alarms.
B.2.9 Power supply failure alarms.
B. 2.10 Recommendations to "purify" automatic control response following loss of hand power (applicable to plants with Bailey Controls l
Model 721 equipment).
B.2.30 Review indicator, recorder, and signal select power supply tags j
to ensure complete information for signals that involve more than one hand station.
B.2.31 On input power momentary disruptions, the ICS and NNI shall not cause the PORVs, turbine bypass valves, and atmospheric dump l
valves (where applicable) to inadvertently operate, affecting the ability to attain a known safe state.
C.8 Eliminate mid-scale failures that can affect indication or plant control.
- The BWOG has identified these recommendations as key recommendations.
NUREG-1231, Supp. 1 6-16
Number Recommendation E.9 Close main FW block valves on loss of ICS de power.
E.10 Evaluate ICS response to NNI power failure for other input signal select switch positions.
E.11 Provide identification of automatic power bus transfer, which is discernable outside the system cabinets.
E.16 Develop independently powered hand or automatic controls for turbine bypass valves, main and startup feedwater valves, and main feed pump speed control.
RR-TIR-24 The BWOG I&C Comittee should develop a post-trip or post-transient troubleshooting procedure to aid in diagnosing ICS module failures.
In general, the staff agrees with and encourages implementation of the above recomendations with the exceptions of B.2.8 and TR-211-ICS, Recomendation B.2.P. involves changing alarm logic from deenergize-to-actuate to energize-to-actuate to reduce the number of spurious signal monitoring alarms that occur upon ICS/NNI power supply failures.
The staff agrees with the objective of eliminating irrelevant / nuisance alarms on power supply failures, but does not concur with recommendation B.2.8 in the absence of further direction to ensure that (1) the revised alarm configuration will not suppress useful information pertinent to timely identification of other ICS/NNI system failures and (2) the revised alarm configuration will not have the potential for misleading the operator in the absence of a legitimate alarmed condition.
The alarm system should allow for timely identification of total or partial ICS/NNI power losses.
The staff is puzzled by the recent appearance (letter dated March 2, 1988) of recommendation TR-211-ICS.
This recommendation was previously rejected by the BWOG because it was in conflict with recomendation TR-178-ICS regarding known safe state (KSS).
As discussed later in this section, the staff considers the l
KSS concept to be the most significant recommendation of the ICS/NNI reassess-ment.
Thus, the staff cannot agree with a recomendation that conflicts with j
the KSS recommendation.
The staff strongly encourages the BWOG to reconsider its acceptance of TR-211-ICS.
Implementation of recomendations TR-012-ICS, TR-158-ICS, B.2.4, and B.2.9 should help to improve ICS/NNI-related alarm / annunciator circuit designs.
The staff i
also believes that implementation of recomendation TR-154-ICS to provide the operator with unambiguous status of control room indicators and recorders upon j
loss of ICS/NNI power or signals will result in substantial benefit to the operator in responding to such events, provided that the indication of failed status is obvious and that timely identification of the failed status does not rely on operator analysis.
A number of the proposed recommendations are directed toward reducing the contribution of ICS/NNI-controlled equipment failures to transient severity /
NUREG-1231, Supp. 1 6-17
_. _ =.
I l
complexity by identifying and recommending preferred / desired equipment failure f
modes for ICS/NNI power losses and by retaining manual control capability for the equipment following power losses.
These recommendations include TR-036-ICS, TR-037-ICS, TR-096-MSS, TR-097-ICS, TR-159-0PS, and TR-172-PRV.
If properly 1
j implemented, these recommendations would help to resolve previously identified l
ICS/NNI concerns, The staff is concerned that the scope of some recommendations is too limited, and that a number of the recommendations do not contain sufficient details to l
ensure that implementation will resolve the related concerns.
It is not clear i
i from the information provided in BAW-1919 why certain chosen equipment failure
)
modes are considered optimum.
It should not be assumed that a given failure mode is proper for all plant conditions.
BAW-1919 does not contain sufficient l
information for the staff to conclude that implementation of the recommendations will resolve concerns regarding the loss of both automatic and remote manual control capability for plant equipment upon loss of ICS/NNI power, which can i
complicate plant recovery / stabilization.
Furthermore, the staff cannot conclude that sufficient instrumentation remains available in the control room for the i
operator to assess plant status and to achieve a safe shutdown condition following a loss of ICS/NNI power, or that sufficient periodic surveillance is performed to ensure operability of the instrumentation.
The staff believes that recommendation TR-190-ICS to develop backup controls for pressurizer level and pressure control should be a high priority recommendation because it will result in a significant reduction in ICS/NNI transient complexity by providing redundant primary system level and pressure control cap 6bility.
The i
staff encourages the BWOG (1) to provide further guidance for implementation of TR-190-ICS to ensure independence of the backup controls from the normal controls and (2) to address the problems of ICS/NNI failures concerning reactor makeup flow control, letdown flow control, and seal injection flow control.
The most significant recommendation resulting from the BWOG ICS/NNI reassess-ment effort is TR-178-ICS, which requires that the plant goes to a KSS in response to loss of ICS/NNI power transients and only relies on manual actions for which thi >perator is normally trained and that can be accomplished from i
the control room.
However, based on the staff's review of Appendix R-g, "Known Safe State on the loss of ICS/NNI Power," it appears that some losses of power can still cause failure of instrumentation and control channels for which back-ups do not exist and can potentially result in relatively complex transients requiring manual actions beyond those normally required of the operator.
At present, the staff is not convinced that all B&W plants will be able to attain i
a KSS for all potential ICS/NNI power losses, in accordance with the BWOG de-finition of KSS.
The staff believes that the combined benefit of implementa-tion of approved Level I recommendations will result in substantial improvement I
in the ability of B&W plants to achieve a KSS following losses of ICS/NNI power.
)
The BWOG Steering Committee rejected recommendation RR-TIR-24 to have the BWOG I&C Committee develop a post-trip or post-transient troubleshooting procedure j
to aid in diagnosing ICS module failures.
The basis for rejection was that the recommendation had no benefit to reduce the number of reactor trips or complex transients.
The staff disagreet with the BWOG basis for rejection.
It appears to the staff that careful inspection and determination of the root cause(s) for ICS module failures could result in corrective actions that could improve NUREG-1231, Supp. 1 6-18
ICS/NNI performance and reduce the ICS contribution to reactor trip frequency and transient complexity.
The staff recommends that the BWOG reconsider this recommendation because it has the potential benefits.
The staff considers recommendations TR-154-ICS, TR-178-ICS, TR-190-!CS, and C.8 to be key recommendations because of their importance with regard to resolving previously identified ICS/NNI concerns; the staff strongly encourages their implementation.
6.1. 7 ICS/NNI Power Reliability TFe staff's technical review and evaluation of the BWOG recommendations related tc reliability of the electrical power sources and distribution systems exter-nai and Internal to the ICS/NNI are given below.
The following recommendations were reviewed:
Number Recommendation TR-013-ICS*
Prevent loss of power to the ICS or NNI.
TR-039-ICS*
Wire the power supply monitor in the ICS/NNI directly to the output bus after the auctioneering diodes.
TR-053-SFI Correct overheating problems that can lead to electronic power supply malfunctions.
TR-102-ICS*
Install redundant de power supplies for NNI-Y (for AP&L only).
TR-103-ICS Fuse external power leaving ICS/NNI cabinets (for FPC only).
TR-105-ICS*
Perform field verification of ICS/NNI drawings.
TR-113-PES Review breaker control power distribution to determine effects of a loss of the battery bus.
TR-116-PES Review de charging system and ensure the charging voltage does not exceed plant equipment voltage rating.
TR-117-PES
- Modify inverter overcurrent protection to ensure the breaker /
fuses open on overcurrent before inverters fail.
TR-118-PES Evaluate loadings on at and de vital buses to ensure adequate margins exist without trip of equipment.
TR-119-PES
- Implement preventive maintenance for electrical buses.
TR-182-ICS Evaluate installing automatic bus transfer switches of MFW pump controllers (for Davis-Besse only).
TR-183-ICS Preventive maintenance and testing for ABT switches.
d
- The BWOG has identified these recommendations as key recommendations.
NUREG-1231, Supp. 1 6-19
j Number Recommendation j
TR-184-ICS Provide separate fuses for hand stations that use ac power, f
TR-186 ICS Minimize access to ICS/NNI cabinets during operation and train maintenance personnel on location of power diJtribution components.
t 1
TR-187 ICS Install current and voltage meters for NNI power supplies (for Davis-Besse only).
l j
TR-188-ICS Maintain de power supply current balance and perform a periodic f
full load test for each power supply.
TR-192-ICS Remove / modify NNI-Z power supply and signal select logic (for i
Rancho Seco only).
4 TR-203-PE5*
Establish preventive maintenance to increase reliability of j
inverters.
E.1 Increase de current limit.
In general, the staff agrees with and encourages implementation of t*e above i
recomendations with the exception of TR-039-ICS, TR-192-ICS, and E.1.
It is not clear to the staff that the advantages and disadvantages of all possible
[
power supply monitor (PSM) locations have been thoroughly investigated.
The
(
staff believes that there are disadvantages associated with connecting the PSM sense / operate lines directly to the 124-volt de ICS/NNI buses as recommended in TR-039-ICS.
The staff disagrees with removal of the NNI-Z power supplies at l
1 Rancho Seco as recommended in TR-192-ICS because it appears that their value 1
(providing the capability to switch from failed to valid signals for input to l
the ICS and control room indications) outweighs any disadvantages concerning inadvertent PSM actuations.
Raising the current limit of the ICS/NNI 224-volt i
de power supplies as recommended in E.1 would reduce design margins for the i
i power supplies and may actually reduce system reliability under certain condi-tions; therefore, this is discouraged, t
i The staff believes that recommendation TR-116-PES, which requires verification that de system charging voltage does not exceed the voltage rating of the asso-i ciated plant equipment, should be expanded to also address potential problems that may occur because the equipment voltage ratings will be exceeded during l
a l
the time the plant batteries are subject to an equalizing charge.
)
The staff believes that a number of the recommendations addressed in this section, if implemented, should significantly enhance ICS/NNI power distribution j
1 system reliability.
The staff concludes that the likelihood of reoccurrence of
~}
specific past events will be reduced by implementation of the recommendations l
that focus on problem areas that have contributed to transient complexity.
l However, the staff believes that certain recommendations (e.g., TR-013-ICS, j
TR-053-SFI, and TR-113-PES) may be too general to ensure that implementation of 1
the recommendation will satisfy the program objectives and systems requirements established by the BWOG.
1 j
l i
)
- The BWOG has identified these recommendations as key recommendations.
1 i
1 1
]
NUREG-1231, Supp. 1 6-20 l
1 I
_- ~._
l l
i TwomajorareasofconcernregardingICS/NNIpowerreliability,whichthestaff l
does not consider to be resolved by the ICS/NNI reassessment, are (1) resolution of the concerns addressed in IE Bulletin 79-27 and (2) the design and operation of the ICS/NNI PSM modules.
These issues are discussed below, i
i On November 10, 1979 an event occurred at the B&W-designed Oconee nuclear plant that involved a loss of ICS and NNI power.
As a result of this event, 3
i the NRC issued on November 30, 1979, IE Bulletin 79-27, "Loss of Non-Class 1E i
Instrumentation and Control Power System Bus During Operation." This bulletin l
l required licensees (1) to review the effects of loss of power to each Class j
IE and non-Class IE bus supplying power to plant instrumentation and controls l
and (2) to determine the resulting effect on the capability to achieve a safe l
(cold) shutdown condition via procedures following the power los,s. A further 4
requirement of the bulletin was to develop procedures necessary to achieve cold shutdown following bus power failures.
The intent of IE Bulletin 79-27 was to ensure that the loss of power to any bus could not result in control system actions that cause a plant upset / transient condition requiring operator action
}
j concurrent with the loss of control room information (e.g., indications and j
alarms) upon which the actions would be based, i
j On February 26, 1980, an event occurred at the B&W-designed Crystal River nuclear plant that involved a loss of NNI power.
The event at Crystal River involved i
i the types of concerns that were identified by IE Bulletin 79-27.
By letter i
l M d March 6, 1980, the staff required all licensees of B&W-designed reactors t
i te. expand the review of IE Bulletin 79-27 to include the implications of the i
Crystal River event.
On March 7, 1980, the staff issued IE Information Notice 80-07, which described the Crystal River event and stated that IE Bulletin 1
79-27 was intended to cause licensees to investigate the loss of individual power supplies as well as the loss of inverters and vital buses.
In addition, l
NUREG-0667, "Transient Response of Babcock & Wilcox - Designed Reactors,"
i published in May 1980 recommended that prompt followup actions be taken on IE Bulletin 79-27 to improve ICS/NNI reliability.
j On December 26, 1985, an event occurred at the B&W-designed Rancho Seco nuclear
]
plant that involved a loss of ICS 124-volt de power.
Upon loss of ICS de power, I'
equipment control modules lost power, thus providing zero de voltage, and switching relays lost power, going to the de-energized state.
This causeri ICS-controlled plant equipment to change positions, initiating a plant tran-l sient, and caused the loss of remote manual control of key ICS-controlled plant l
j equipment from the control room.
Furthermore, the operators were misled by an
(
MFW flow recorder that had failed to the mid-scale position upon the loss of 4
power.
l The incident at Rancho Seco on December 26, 1985, was significant because it again demonstrated that a single failure in the non-safety-related ICS/NNI could subject the plant to an undesirable transient and challenge the operator's capability to mitigate the transient without resulting in primary system under-cooling or overcooling.
The event demonstrated that the implementation of modifications to meet the concerns of IE Bulletin 79-27 were inadequate at B&W-i designed reactors.
The staff's final review of utility responses to IE Bulletin 79-27 only focused on whether there was reasonable assurance that the concerns of the bulletin had l
NUREG-1231, Supp. 1 6-21 l
~
been properly addressed.
Additional background information regarding licensees' responses to IE Bulletin 79-27 and the staff's evaluation of these responses is i
provided in Section 7, "Precursors to the December 26, 1985 Incident at Rancho Seco and Related NRC and SMUD Actions," of N'JREG-1195. "Loss of Integrated Con-trol System Power and Overcooling Transient at Rancho Seco on December 26, 1985."
The staff believes that if a more thorough and in-depth review of plant designs l
i by the utilities in accordance with the bulletin had been performed the poten-J tial for this type of event would have been detected and resulted in hardware i
an3 fr procedural modifications to ensure that a safe shutdown condition could
)
be achieved following an ICS power loss.
~
Because of the deficiencies identified in the IE Bulletin 79-27 rs iew performed v
at one B&W-designed plant and the importance of proper resolution of the bulle-i 1
tin concerns, the staff requested the BWOG to address these concerns as part of i
the ICS/NNI reassessment effort.
The BWOG has indicated that the IE Bulletin 79-27 issue is plant specific and should be addressed to the individual utili-ties and is not within the scope of BAW-1919.
Therefore, the staff is unable to conclude from its review of the BWOG ICS/NNI evaluation that the concerns identified in IE Bulletin 79-27 have been satisfactorily implemented for B&W reactors.
The staff considers resolution of the IE Bulletin 79-27 concerns at B&W reactors to be necessary to resolve concerns regarding the effects of j
non-safety-related (control) system power supply failures at B&W plants.
j Therefore, the staff recommends that, as a part of the plant-specific imple-mentation audits discussed in Section 12.2, the plant specific actions taken 4
i in response to IE Bulletin 79-27 also be audited.
l The BWOG has proposed (recommendation TR-039-ICS) to change the way in which the PSM is connected within the ICS and NNI 124 volt de distribution systems.
The change involves connecting the PSM positive and negative 24-volt de bus i
i monitoring circuits to termination points on the buses themselves, thus sensing l
bus voltage directly.
The PSM module has typically been included in a "daisy chain" circuit that provides power to additional ICS/NNI electronic modules.
4 It is not clear to the staff that connecting the ICS/NNI PSMs directly to the i
buses is the preferred / optimum arrangement.
One apparent advantage of locating the power supply module at the end of a daisy chain would be that the PSM would detect and provide protection for degraded voltage conditions within the dis-tribution system (caused by individual module failures within the daisy chain) as well as for degraded voltage at the 124-volt de buses.
The staff acknowl-3 edges that in the existing ICS/NNI power distribution configurations, which l
consist of multiple daisy chain end points, it is not possible to monitor the
]
voltage to all system modules with a single PSM.
However, it appears that with i
relatively minor changes it may be possible to modify the existing power dis-t tribution configuration such that a single PSM can be used to monitor the q
1 voltage supplied to the majority of the system modules.
Connecting the PSM
]
inputs directly to the buses does not provide a means to prevent control system I
operation with degraded voltage to individual cabinets or groups of modules because of the series resistances or open circuits in the distribution wiring; although it may reduce the probability of PSM actuation, it does not reduce the
{
probability of this type of problem.
l In addition, the staff has identified conceras regarding questionable design characteristics of the PSM module itself, including l
l l
1 1
i NUREG-1231, Supp. 1 6-22 i
1
use of monitored voltage to operate the undervoltage detection circuits the lack of a seal-in feature upon sensing that the monitored voltage no exceeded alarm and trip set point values the small hysteresis value between the trip set point and reset point The staff recommends that the BWOG consider performing additional investigations concerning the adequacy / desirability of the PSM design and operation.
This review should include an independent design analysis of the PSM to determine the appropriateness of the design concept to serve its intended purpose and a detailed circuit analysis to determine the adequacy of the design to imple-ment the design concept.
The design analysis should include an assessment af the appropriateness of using the same voltage to operate the PSM as is being monitored by the PSM and an assessment of the need for "seal in" features.
The analysis also should include a determination of the optimum PSM location within the ICS/NNI distribution systems and the benefit that could be gained by using redundant PSMs in conjunction with any PSM design changes considered to be appropriate.
Furthermore, if it is determined that a cesign function of the PSM is not to provide protection against ICS/NNI module operation with voltages outside of the vendor-specified 124-volt de 110 percent tolerance band, then methods should be developed for ensuring that module operating volt-age remains within the specified tolerances.
6.1,8 ICS/NNI Maintenance and Surveillance Testing The staff's technical review and evaluation of the BWOG recommendations related to ICS/NNI periodic maintenance and surveillance testing (including system tuning are given below).
The following recommendations were reviewed:
Number Recommendation TR-009-ICS*
Improvements in ICS tune control circuits.
TR-010-ICS ICS control circuit modification.
TR-011-ICS Determine if the grid frequency error circuit has been detuned.
TR-038-ICS*
Develop and implement a preventive maintenance program for the ICS/NNI.
TR-068-MFW Develop a post
- maintenance testing program for the MFW pump turbines and governor controls.
TR-079-MFW Put MFW regulating valves, main block valves, and startup con-trol valves on a refueling frequency for an operational check.
TR-105-ICS*
Perform field verification of ICS/NNI drawings.
TR-106-ICS Remove unused hardware from ICS/NNI cabinets.
TR-107-ICS*
Improved maintenance and tuning of ICS.
- The BWOG has identified these recommendations as key recommendations.
NUREG-1231, Supp. 1 6-23
l
[
i l
i 2
Number Recommendation i
TR-111-RPS Review safety system surveillance procedures for checking which j
channel is available for testing prior to initiation of test.
i 1
TR-119-PES
- Implement preventive maintenance for electrical buses.
{
TR-163-EFW*
Review EFW surveillance and test procedures to ensure that l
components used in the E0Ps are included in the test program.
{
1 TR-164-EFW*
Review EFW preventive maintenance program, including minimizing potential from common-cause failures arising from maintenance i
i and testing procedures.
t TR-165-EFW*
Review EFW maintenance ar.d test procedures to eliminate con-flicting and confusing instructions, i
1 TR-166-EFW Implement a program to improve and maintain the availability I
and performance of the EFW systems.
l TR-168-MTS Provide guidance in procedures when troubleshooting the EHC.
j
)
TR-183-ICS Preventive maintenance and testing for ABT switches.
TR-186-ICS Minimize access to ICS/NN! cabinets during operation and train f
maintenance personnel on location of power distribution 1
components.
TR-188-ICS Maintain de power supply current balance and periodically per-l 3
form full load test for each power supply.
[
4 1
TR-203-PES
- Establish preventive maintenance to increase reliability of
{
inverters.
}
t i
TR-208-ICS Label ICS/NNI switches 51 and 52 to detect energized vs. tripped l
positions.
3 i
l TR-210-ICS Establish program to monitor control system.
\\
j B.2.15 Perform periodic surveillance of ICS and NNI cabinet DC voltage.
{
l B.2.16 Periodically record ICS module input and output voltages.
j B.2.17 Periodically test functions of ICS not normally demanded.
l B.2.25 Review / revise ICS and NNI drawings as needed for legibility, j
j B.2.26 Improve documentation for ICS tuning.
f E.26 Build a data bank for analyzing plant transients and assistir.g in predicting areas (hardware) that degrade overall system performance.
- The BWOG has identified these recommendations as key recommendations.
i l
NUREG-1231, Supp. 1 6-24 1
\\u
Number Recommendation TAC-2.e The post-maintenance program recomended in TR-068-MFW should include proper checkout and tuning of newly installed systems or equipment prior to unit startup, and planned tuning at power should be incorporated into plant procedures.
TAC-3.a Include the lessons learned from the Rancho Seco experience regarding high resistance relay contacts as a part of the rec-omended ICS/NNI preventive maintenance program (TR-038-ICS).
TAC-3.c Modify the existing RTS recomendation TR-038-ICS to include internal and external power supplies to the ICS/NNI.
RR-TIR-26 Review and upgrade maintenance procedures as necessary to ensure that proper checkout of replacement modules and assemblies in the CRDS is made before their use.
This recomendation should be extended to ICS and EHC components, circuit boards, etc.
RR-TIR-8 Review ICS tuning to ensure control settings are compatible with a turbine trip runback from less than 45 percent power.
The staff generally agrees with and encourages implementation of all of the above recommendations.
As an important element necessary for improved ICS/NNI performance, the BWOG has identified the need to perform routine maintenance and surveillance on ICS and NNI components and actuated equipment to ensure and periodically verify their proper operation.
Poor maintenance and surveillance practice for ICS/NNI components and actuated equipment at B&W plants have directly contributed to the frequency and severity of transients involving the ICS/NNI.
Many reported reactor trips and operational transients have been attributed, in whole or in part, to inadequate maintenance and surveillance.
On the basis of its review of the combination of approved and proposed rec-omendations developed by the BWOG concerning maintenance and surveillance of the ICS/NNI and other systems / equipment, the staff concludes that the BWOG is placing proper emphasis in this area.
For example, the BWOG has emphasized the importance of correct system tuning (i.e., ensuring that system components /
modules are adjusted such that controlled systems /equipthent perform their design functions without unnecessary oscillations, perturbations, or delays that could challenge plant operators and plant equipment).
Although the BWOG reassessment effort has been very productive with regard to establishing ICS/NNI maintenance and surveillance programs, the staff is con-cerned that the guidance provided ir recomendations TR 038-ICS (develop and implement a preventive maintenance program for ICS/NNI) and TR-107-ICS (improve maintenance and tuning of ICS) is too general to ensure resolution of certain specific concerns regarding instabilities and perturbations resulting from improper cor. trol system tuning and surveillance.
The recomendations do not include or reference the guidance provided in BAW-1919, Appendix C or Appendix R-n, with regard to ICS tuning.
These recomendations and the associated ref-erenced source documents also do not include information concerning several additional issues, including the scope and intent of certain tests, recomended PSM module alarm and trip set point values, or specific ICS signals to be monitored during power escalation, j
NUREG-1231, Supp. 1 6-25
i i
i l
The surveillance frequency specified in the ICS/NNI systems requirements docu-l ments for instrument channel calibrations, actuated equipment tests, and system j
tuning is "at least every other refueling outago."
It is not clear M the staff that this surveillance frequency is adequate to fulfill the intent of BWDG recomendation TR-038-ICS, which requires periodic ICS/NNI surveillance /
1 preventive maintenance to improve ICS/NNI reliability.
Loss of ICS/NNI c w er i
events at B&W plants have been attributed to the lack of preventive mainter.ance j
and/or surveillance.
An interval of at least every other rv/uol Mg outage j
could be 3 years or more.
The staff considers such an interval to be excess"e for calibration of electronic modules.
Electrical / electronic equipmeat is l
j typically checked annually for operability /t 'ft, if not more frequenUy i
depending on its application and the consequences of failure.
The BWOG ha'. net j
provided information in BAW-1919 that supports.the frequencies selected for periodic ICS/NNI surveillance and preventive maintenance.
The BWOG rejected recomendation RR-TIR-8 (TAC-1.h), which concerned tuning to 4
ensure control settings are compatible with a turbine trip runback from less i
than 45 percent power, on the basis that ICS tuning is not a function of the anticipatory reactor trip system (ARTS) set point and that the intent of the l
recomendation is accomplished by TR-107-ICS.
The staff agrees with this.
l The BWOG also has rejected recomendation RR-TIR-26 (TAC-5.b), which called for the review and upgrade of maintenance procedures as necessary to ensure that proper checkout of replacement modules and assemblies in the CROS is made before their use.
It also was recommended that this proposal be extended to such j
i items as ICS and EHC components and circuit boards.
This recommendation was l
rejected on the basis that it is not a feasible or p'.ctical solution to the l
problem.
The staff does not agree that checkout of replacement components is not feasible.
On the contrary, the staff believes checkout of replacement components is not only feasible, but is a necessary and proper practice.
The 1
BWOG should reconsider this recommendation.
The staff concludes that implementation of these recommendations will contrib-ute to improved maintenance and surveillance of the ICS/NNI.
In addition to
[
recommendations TR-038-ICS and TR-107-!CS, the staf f considers recomendations TR-105-ICS, TR-111-RPS, and TR-163 EFW through TR-165-EFW to be key recomen-i I
dations because of their importance to resolving previously identified ICS/NNI j
concerns; the staff strongly encourages their implementation.
Recommendations TR-009-ICS through TR-Oll-ICS and recommendation TR-119-PES are further dis-2
}
cussed in Sections 6.1.4 and 6.1.7, respectively.
6.1.9 Operator Burden, Procedures and Training The results of the staff's technical review and evaluation of the BWOG recom-mendations intended to reduce the burden on control room operators nd to i
improve procedures and operator training for events involving. J/NNI failures i
are given below.
The staff also reviewed non-ICS/NNI recommen1ations concern-t itg operator burden, procedures, and training.
The following recomendat ons l
were reviewed-Number Recommendation i
I l
TR-012 ICS Determine if the operator has necessary information fw proce-j dures, indicators, etc. to detect loss of NNI power.
{
NUREG-1231, Supp. I 6-26 1
1
Number Recommendation TR-018-MFW Provide training on MFW system components.
TR-034-ICS Training for loss of ICS power.
TR-035-ICS Familiarize operators with Rancho Seco event.
TR-059-0PS Training for personnel who make emergency notifications.
TR-067-MFW*
Wherever possible, eliminate automatic MFW pump trip functions.
TR-069-MFW*
Eliminate automatic control of the MFW block valve except during a reactor trip.
TR-070-MFW*
Provide capability to override a close signal to the MFW block valve.
TR-091-MFW Eliminate need for en auxiliary operator to open a deaerator feedwater tank drain iine after reactor trips (for Davis-Besse only).
TR-098-MFW*
Overfill protection for MFW system.
TR-099-0PS*
Include guidance on excessive MFW, throttling AFW, and throttling HPI in plant procedures.
TR-121-IAS Make appropriate personnel aware of importance of instrument air system prohibition of use for tools and need to report air system damage.
TR-128-IAS Review training and loss of air response procedures for instru-ment air system.
TR-130-IAS Expand procedure for the loss of instrument air (for ANO-1 only).
TR-144-IAS Develop or upgrade a loss-of-instrument air procedure (for FPC, GPUN, SMUD and TED only).
TR-154-ICS*
Provide operator with unambiguous status of indicators and re-corders in main control room on loss of ICS/NNI power or signal.
TR-155-EFW" Limit maximum flowrate delivered by the EFW system.
TR-156-0PS Provide a designated communicator to relay emergency plan messages.
TR-167-PES Include in the operating procedures guidance on restoration of power to electrical buses, especially if the ICS or ICS con-trolled equipment is affected.
j TR-171-0PS Evaluate alarm set points to determine if adequate tine is provided for operator response.
- The BWOG has identified these recommendations as key recommendations.
NUREG-1231, Supp. 1 6-27
i Number Recommendation TR-178-ICS*
Ensure plant goes to known safe I&C state on loss of power to the ICS/NNI.
TR-181-0PS Verify adequacy of instrumentation and displays used to assess and control the AT0G stability parameters.
TR-207-0PS Review operator training with regard to the manual control of WW post-trip.
TR-212-ICS Power source switches or breaker positions should be labeled to prevent their energized-vs.-tripped positions from being misinterpreted.
B.2.22 Train ICS/NNI cabinet monitoring personnel in location of power distribution components.
B.2.23 Train ICS/NNI maintenance personnel with respect tc power supply distribution to indicators, recorders, and hand stations.
B.2.24 Periodically train operator and maintenance staff in the identification and definition of power supply failure modes and the location and use of alternate controls.
B.2.27 Improve / add NNI output functions and signal input ranges for enhanced readability.
E.6 Evaluate operator's ability to rapidly respond to a power fail-ure to the ICS/NNI.
E.18 Review operating procedures with respect to present ICS trip circuits.
TAC-2.e The post-maintenance program recommended in TR-068-MFW should include proper checkout and tuning of newly installed systems or equipment before unit starttp, and planned tuning at power should be incorporated into plant procedures.
RR-TIR-24 The BWOG I&C Committee should develop a post-trip or post-transient troubleshooting procedure to and in diagnosing ICS module failures.
RR-TIR-26 Review and upgrade maintenance procedures as necessary to ensure that proper checkout of replacement modules and assemblies in the CRDS is made prior to their use.
This recommendation also should be extended to ICS and EHC components, circuit boards, etc.
Several recommendations that relate to operator bJrden, procedures, and training are addressed in other sections of this report (not all of the recommendations
- The BWOG has identified these recommendations as key recommendations.
NUREG-1231, Supp. 1 6-28
addressed in other sections e W ted above).
These include recommendations TR-104-ICS, TR-154-ICS and B.2.14 aoNiressed in Section 6.1.3; recommendations TR-007-ICS, TR-009-ICS and TR-010-ICS addressed in Section 6.1.4; recommenda-tions TR-012-ICS, TR-032-ICS, TR-096-MSS, TR-097-EFW, TR-154-ICS, TR-167-PES, TR-178-ICS, TR-181-0PS, and RR-TIR-24 addressed in Section 6.1.6; and recommen-dations TR-009-ICS, TR-010-ICS, TR-038-ICS, TR-068-MFW, TAC-2e, and RP-TIR-26 addressed in Section 6.1.8.
The BWOG Operator Support Committee defined operator burden as those factors that unnecessarily hinder the plant operator's ability to prevent a reactor trip or mitigate plant transients post-trip.
Staff concerns related to operator burden include dependence on operator actions from outside the main control to mitigate the consequences of events involving the loss of ICS/NNI power and proper control system tuning for reactor operation at low power levels to prevent undue operator burden in this region (e.g., having to manually control system parameters without tripping the reactor during star'up).
Operator burden following loss of ICS/NNI pwer is closely associated with the concept of KSS.
An early effort by the BW0G identified in the RTS (TR-033-ICS) was to make system changes to ensure that on any loss of ICS/NNI power the plant would go to a KSS without any operator action required.
This action has since been superseded by recommendation TR-178-ICS, which states that the plant should go to a KSS on loss of power to the ICS/NNI; but with heat balance main-tained by either automatic control and/or operator action for which the oper-ator is normally trained and can be taken from the control room.
The BWOG has stated that the operator action required for a loss of ICS/NNI power that results in a plant trip should be the same as for any other reactor trip.
The staff agrees that it is reasonable to expect such action from the operator.
However, the operator action required to achieve a KSS for some loss of ICS/NNI power transients has not been sufficiently addressed in BAW-1919, as discussed in Section 6.1.6 of this report.
Appendix P, "A Comparative Study of the Sensitivity of B&W Reactor Plants," concludes that the operator burden follow-ing reactor trip (but with no malfunction of the automatic control system) in B&W plants is no greater than for other pressurized-water reactor (PWR) plants, but that the burden imposed on operators to oversee the automatic control system and to take corrective action when they fail is greater in B&W plants than in other PWRs.
This is because the control of key variables are vested in a single system (the ICS).
The staff concludes that implementation of the recommendations listed above generally should help to reduce operator burden, hence the probability of operator error, when responding to losses of ICS/NNI power.
A number of the recommendations proposed by the BWOG are designed to limit the consequences of losses of ICS/NNI power and to prevent the need for manual operator actions outside the control room following reactor trips and ICS/NNI failures.
Im-provements in procedures and training are also expected to reduce operator burden.
As part of the ICS/NNI reassessment, the BWOG was requested to evaluate proce-dures for loss and restoration of ICS/NNI power, maintenance and surveillance procedures for ICS/NNI components, and the adequacy of procedures used to achieve safe plant shutdown following the loss of power to any bus (as required by IE Bulletin 79-27, discussed in Section 6.1.7 of this report).
The staff's NUREG-1231, Supp. 1 6-29
review of BAW-1919, Appendix R g, "Known Safe State on the loss of ICS/NNI Po.<er," indicates that not all plants can achieve a KSS without operator u tion.
Appendix R g does not provide a discussion of the procedures that would cortrol the operater action, but does identify required operator action to achieve KSS and the timing of and the procedures governing the action as areas for which the BWOG expects to continue to study as a followup action to the SPIP.
BAW-1919 states that the E0Ps are backed up by procedures that address abnormal plant configurations created by such conditions as station blackout, loss of ICS/NNI power or loss of instrument air.
It also states that the connections and interactions between the E0Ps and the specific procedures for abnormal plant configurations should be clear and concise with proper priorities main-tained.
As part of the ongoing emergency procedures review discussed in Section 7.4 of the SER, the BWOG is reviewing the procedural hierarchy for each plant to determine if the proper links and priorities have been maintained and to make recommendations on where improvements should be made.
BAW-1919 also states that training of personnel is a plant-specific item and that it shall be the owner's responsibility to designate and train system or performance personnel who will be responsible for proper operation of the ICS/NNI systems.
The staff concludes that the SPIP reassessment has proposed recommendations that if adequately implerrented will reduce operator burden and improve proce-dures and training with regard to the ICS/NNI.
However, the staff is concerned that insufficient references to source documents and insufficient detail con-cerning the basis for the recommendation exists for some recommendations.
To ensure the resolution of concerns, as intended by the recommendations, the BWOG should ensure that all appropriate source documents are referenced.
6.1.10 Other ICS/NNI Considerations The results of the staff's technical review and evaluation of the ICS/NNI reassessment with regard to resolution of concerns not covered in other sec-tions of this report and the associated recommendations listed in BAW-1919.
This section discusses the following topics:
steam generator overfill protection FSAR assumptions NRC Standard Review Plan Section 7.7, "Control Systems" integrated vs. discrete controls advanced control systems The following recommendations were reviewed:
Number Recommendation TR-098-MFW*
Overfill protection for MFW system.
TR-155-EFW*
Limit maximum flowrate delivered by the EFW system.
- The BWOG has identified these recommendations as key recommendations.
NUREG-1231, supp. 1 6-30
=-.
Number Recommendation C.1 Provide independent controls for reactor power, steam demand, and feedwater such that no single failure would lead to simul-taneous and spurious manipulation of control rods, turbine steam admission valves, and feedwater flow.
C.7 Provide separate subsystems for reactor coolant temperature and steam pressure control.
D.1 Identify, document, and resolve problem areas in the ICS/NNI system that, based on past operating experiences and system evaluations, can be corrected by an advanced control system.
D.2 Develop an optimal system configuration for the advanced control systems to determine desirable interfacing to other plant sys-tems and to minimize plant changes required to accommodate the new system.
D.3 Explore technological improvements / capabilities for incorpora-tion into the advanced system to improve operability reliability and maintainability.
D.4 Review control loop architecture for possible upgrades to pro-vide improved unit control and man-machine interfaces.
D.5 Ensure capabilities of the plant safety related systems to achieve a safe shutdown condition tre not dependent upon or compromised by action of the plant control system.
D.6 Provide advanced control system features to regain plant sta-bility during abnormal plant conditions prior to challenging the plant protection systems.
E.14 Remove automatic control functions for the atmospheric dump valves from the ICS/NNI system.
E.23 Post-trip EFIC actuation and isolation of MFW on high steam generator (SG) level.
RR-6 Emphasize the need to establish communication with the control room before taking action on safety system equipment if an unplanned actuation occurs during maintenance, testing, etc.
RR-TIR-19 The plants that trip both MFW pumps on high SG level should 1
evaluate sequential vs. simultaneous MFW pump trip.
The staff generally agrees with and encourages implementation of the above recommendations.
The staff has identified concerns regarding the potential for steam generator overfill (and potential overcooling) events from excessive MFW flow and/or excessive AFW flow.
Overfill could occur as a result of the MFW startup valve NUREG-1231, Supp. 1 6-31
failing mechanically or the loss of its ICS control signal upon loss of ICS power (this would cause the valve to go to the 50 percent open position, which could allow excessive post-trip MFW flow).
In addition, although AFW system control has been made independent of the ICS, the AFW system designs typically include preferred failure modes for valves that ensure adequate AFW flow for decay heat removal; these same failure modes (e.g., flow control valves failing open on loss of power or instrument air) could lead to SG overfill.
The staff requested the BW0G to address the potential for SG overfill from both the MFW and AFW systems.
The BW0G stated that the plant-specific FMEA program (addressed in Section 6.1.5 of this report) would address potential ICS failures that could cause SG overfill.
Several of the recommendations proposed by the BWOG should help to reduce the probability of SG overfill events.
The staff encourages implementation of recommendations TR-098-MFW and E.23 for mitigation of potential overfill events.
The staff agrees with the BWOG basis concerning rejection of recommendation RR-TIR-19 since the simultaneous tripping of both MFW pumps is consistent with the goal of preventing SG overfill and is consistent with recommendation E.23.
However, the staff requests that the BWOG reconsider the basis for rejection of recommendation RR-6, which provides for emphasizing the need for communica-tions between the control room and plant staff prior to working on safety-related equipment if an unplanned actuation occurs during maintenance or testing.
It appears that the potential benefit of this recommendation would warrant its approval.
Following the loss of ICS power event at Rancho Seco on December 26, 1985, the staff identified concerns regarding the FSAR Chapter 15 transient and accident analyses, which appear to assume proper operation of the non-safety-related ICS/NNI.
The ICS/NNI systems requirements state that:
The ICS/NNI system is classified as non-safety.
This classification derives from the fact that ICS/NNI is in no way used to protect the integrity of the primary reactor coolant pressure boundary, to assure the capability to shutdown (SIC) the facility and maintain it in a safe shutdown condition, or to assure the capability to prevent or mitigate the consequences of accidents which could result in unplanned offsite releases in excess of the criteria established in 10 CFR 100.
BAW-1919, Appendix R-t, "FSAR Assumptions Relating to ICS/NNI," provides a table that identifies ICS/NNI actions assumed in the FSAR analyses for 15 different accident scenarios at the six operating B&W reactor facilities (Crystal River, Davis-Besse, Oconee, THI-1, Arkansas-1, and Rancho Seco).
It appears that the ICS may be assumed to function properly to help mitigate the consequences of postulated accidents for some of the scenarios.
The BWOG has stated that the information provided in Appendix R-t is preliminary and that the BWOG I&C Committee will continue its evaluation in this area as a post-SPIP follow-on activity.
The staff encourages the BWOG to verify that the mitigation of postulated accidents is not dependent on proper operation of the ICS/NNI.
Section 7.7, "Control Systems," of the NRC Standard Review Plan (SRP), NUREG-0800, provides review guidance concerning non-safety-related control systems to be used to determine compliance with applicable regulatory requirements.
Control NUREG-1231, Supp. 1 6-32
system designs are considered to conform to the applicable general design criteria (GDC) if all of the following conditions listed in SRP Section 7.7 are satisfied:
(1) The review should confirm that the control systems satisfy the require-ments of the acceptable criteria and the system design bases.
(2) The review should confirm that the plant accident analysis in Chapter 15 of the SAR does not rely on the operability control systems to ensure safet..
(3) The review should confirm that the safety analysis includes consideration of the effects of both control systems action and inaction in assessing the transient response of the plant for accidents and anticipated operational occurrences.
(4) The review should confirm that the consequential effects of anticipated operational occurrences and accidents do not lead to control systems failures that would result in co, sequences more severe than those bounded by the analysis in Chapter 15 of the SAR.
(5) The review should confirm that the failure of any control system compo-nent or any auxiliary supporting system for control systems do not cause plant conditions more severe than those bounded by the analysis of antici-pated operational occurrences in Chapter 15 of the SAR.
(The evaluation of multiple independent failures is not intended.)
The staff encourages the BWOG to pursue che implementation of recommendations that ensure B&W ICS/NNI designs conform to the guidance listed in SRP Sec-tion 7.7.
The BWOG has proposed recommendations (C.1 and C.7) on the basis of the MPR sensitivity study discussed in Section 5 of the SER.
If implemented, these recommendations would tend to separate ICS control functions to reduce the ICS contribution to transient complexity upon single failures.
The staff encour-ages implementation of these recommendations because they will cause the ICS to respond more like separate / discrete control systems to single failures.
An advantage of discrete control systems is that when a single control system fails, the other control systems, being electrically separate and independent, are not affected and, therefore, tend to stabilize overall plant conditions.
In contrast, when the integrated control system fai's, the effects may be fed throughout the plant, causing the overall plant conditions to degrade rapidly.
l As part of the BWOG ICS/NNI reassessment, consideration was given to the use of advanced control system designs to perform functions currently performed by the ICS.
The design changes being considered in recommendations TAC-6.d and TAC-6.e and the e,hanges being considered with regard to new advanced control system designs (0.1 through D.6) should result in substantial improvement in re-ducing the complexity of transients caused or aggravated by the ICS/NNI system.
6.1.11 Previously Identified ICS/NNI Concerns and Recommendations The staff's conclusions regarding the effectiveness of the BWOG SPIP ICS/NNI reassessment in resolving concerns identified during the investigation of NUREG-1231, Supp. 1 6-33
l events at B&W reactors involving ICS/NNI system failures are given below.
The staff reviewed a number of documents providing the results of investigations of B&W reactor transients involving the ICS/NNI to compile a list of previously identified concerns (PICS).
The documents included BAW-1564, "Integrated Control System Reliability Analysis," dated Au)ust 1979.
IE Bulletin 79-27, "Loss of Non-Class IE Instrumentation and Control Power System Bus During Operation," dated November 30, 1979.
NUREG-0667, "Transient Response of Babcock & Wilcox-Designed Reactors,"
dated May 1980.
NUREG-1195, "Loss of Integrated Control System Power and Overcooling, Transient at Rancho Seco on December 26, 1985," dated February 1986.
The staff's review of these and other documents resulted in a list of PICS.
A list of the more significant PICS, many of which appeared to be common to the more severe B&W plant transients involving the ICS/NNI, is provided in Appendix E to this report. While Appendix E includes guidance / recommendations for resolving the PICS, it should be noted that many of these recommendations have not been imposed as regulatory requirements.
A complete list of PICS related to the ICS/NNI, including concerns of lesser significance and/or that may be applicable to only a small number of the B&W plants, will be provided in the staff's detailed technical report on the ICS/NNI reassessment.
Following the loss of ICS power at Rancho Seco on December 26, 1985, the staff questioned why complex plant transients involving the ICS/NNI were continuing to occur given that a number of improvements to ICS/NNI systems were thought to have been made.
Investigation of this event and the March 19, 1984 loss of NNI power event (also at Rancho Seco) indicated that the PICS raised from earlier loss of ICS/NNI power events (e.g., loss of ICS/NNI power at Oconee Unit 3 on November 10, 1979, and loss of NNI power at Crystal River on February 26,1980) continued to exist at some B&W plants.
From the onset of the ICS/NNI reassess-ment effort, the staff specifically asked the BWOG (via the letters referenced in Appendix F) to address the PICS listed in Appendix E.
In general, based on the review of information provided in BAW-1919, the staff cannot conclude that the PICS listed in Appendix E will be fully resolved through implementation of the approved recommendations in the RTS.
It does appear, given implementation of the SPIP-approved recommendations, that PIC number 6, regarding the provision of backup power supplies and automatic trans-fer capability for power supplies to increase ICS/NNI power reliability, will be resolved.
This seems to be an area of significant improvement to ICS/NNI systems over the years.
For a number of the PICS, the BWOG recommendations (approved and proposed), if implemented, would provide significant progress towards resolution.
This ap-pears to be especially true for PICS 2, 8,12, and 19 as a result of the rec-ommendations to dictate the failure modes of ICS/NNI-controlled equipment, thus reducing the consequences / complexity of events involving ICS/NNI power fail-ures.
The staff also believes that significant improvements concerning reso-lution of PICS 10, 11, 14, 17, 18, 22, and 24 will be achieved if the BWOG I&C NUREG-1231, Supp. 1 6-34
i Committee recommendations (approved and proposed) are properly implemented.
Conver:ely, based on the information provided in BAW-1919, it is not apparent that PICS 1, 4, 5, 13, 16, 20, and 23 have been/will be resolved upon imple-mentation of the 8WOG recommendations.
The staff's review concludes that insufficient information is provided in BAW-1919 to make a determination con-cerning resolution of PICS 3, 7, 9, 15, 21, and 25.
The overall staff conclusion regarding resolution of the PICS listed in Appendix E through the BWOG SPIP is that the combination of approved and proposed recommendations, if implemented, would help to resolve a number of these concerns.
Therefore, the BWOG SPIP has been successful to a significant degree in achiev-ing its goal of a reduction in the frequency of reactor trips and in the com-plexity of transients involving ICS/NNI-related failures.
However, the staff remains uncertain regarding the resolution of a number of PICS.
In addition, some of the PICS appear not to have been resolved by the BWOG SPIP.
6.1.12 Instrumentation and Control System Issues Not Directly Related to ICS/NNI The staff's review and evaluation of instrumentation and control (I&C) systems addressed by the BWOG that are not directly related to the ICS/NNI are given below.
BWOG recommendations in three areas were reviewed:
(1) systems designed to isolate MFW and/or AFW flow to a depressurized steam generator (i.e., rupture control systems), (2) main feedwater system related I&C issues, and (3) the main turbine system electrohydraulic control (EHC) system.
The following recommendations were reviewed:
Number Recommendation TR-014-MFW*
Install monitoring system on MFW pumps to document causes of pump trips.
TR-016-MFW*
Investigate oil system pressure in MFW pump.
TR-017-MFW*
Evaluate MFW pump control systems.
TR-019-MFW Ensure there are sufficient annunciator and trip signals for MFW supply system.
TR-021-ICS*
Identify causes for MFW pump control problems.
TR-025-MTS*
Review EHC system for loss of input power.
TR-052-SFI*
Filter steam generator level signals in steam feedwater rupture control system.
TR-066-MFW*
Ensure that a single electrical failure will not cause a loss of both feedwater trains.
TR-067-MFW*
Wherever possible eliminate automatic MFW pump trip functions.
- The BWOG has identified these recommendations as key recommendations.
NUREG-1231, Supp. 1 6-35
Number Recommendation TR-069-MFW*
Eliminate automatic control of the MFW block valve except during a reactor trip.
TR-070-MFW*
Provide capability of override a "close" signal to the MFW block valve.
TR-071-MFW Install valve position indication for the startup and MFW regu-lating valves.
TR-072-MFW Eliminate the transfer from the startup to the MFW flowmeter when the MFW block valve opens.
TR-085-MFW Modify main FW pump recirculation valve for automatic control during startup and shutdown.
TR-088-MFW Eliminate automatic plant runback on low MFW pump discharge pressure or establish set point to achieve a successful runback.
TR-091-MFW Eliminate need for an auxiliary operator to open a deaerator feedwater tank drain line af ter reactor trips (for Davis-Besse only).
TR-100-MTS*
2eview MSR drain tank level control and drain line configuration.
TR-169-MTS Evaluate possibility for defeating the high vibration trip during main turbine valve testing (for GE turbines only).
TR-170-MFW Evaluate placing orifice snubbers in the MFW pump control oil system.
TR-173-MFW Ensure in procedures that MFW pump status to ARTS /RPS is reset after each MFW pump is operational.
TR-180-MTS Provide a monitoring capability for the EHC system for purpose of root cause determination.
TR-200-MTS Install a time delay relay or an orifice between the EHC oil system and the ARTS sensing line to prevent oil pressure perturbations.
{
TR-201-MTS Review EHC overspeed and fast control and intercept valve circuits.
TAC-1.f The BWOG I&C Committee should evaluate a design modification to automatically reset the ARTS bistables on turbine reset.
TAC-2.d Expand RTS recommendation TR-067-HFW to include the condensate pumps and condensate (or feedwater) booster pump.
Also, modify the wording of TR-067-MFW to harden those trip signals that are required for pump / plant protection, j
i
- The BWOG has identified these recommendations as key recommendations.
1 i
NUREG-1231, Supp. 1 6-36
Number Recommendation TAC-2.i The BW0G I&C Committee should evaluate implementing a design modification to automatically reset the ARTS bistables on MFW pump reset.
RR-2.1.7 Although each utility has taken action to reduce the propensity for turbine overspeed trips, further actions are recommended.
RR-11 Provide indication of the status of the MFW pump and turbine trip bistables into ARTS /RTS in the main control room.
RR-TIR-19 The plants that trip both MFW pumps on high SG level should evaluate sequential vs. simultaneous MFW pump trip.
The BWOG has identified the MFW system as one that has contributed signifi-cantly to the frequency of reactor trips and that has been frequently involved in post-trip complications, especially with regard to primary system overcool-ing.
Reactor trips and post-trip complications have been caused by problems /
failures involving the MFW pumps and MFW system valves and flow control.
Problem areas identified include instrumentation and control circuit malfunc-tions, spurious pump trips, and maintenance-and procedure-related concerns.
EHC system failures also have contributed to the frequency of reacter trips and post-trip transient complexity.
EHC system problems have included pump /cil system failures, control circuit malfunctions, and power supply reliability.
The staff generally agrees with and encourages implementation of the above recommendations, which have the potential to significantly reduce the frequency of reactor trips and the complexity of post-trip plant behavior.
It is noted that recommendations TR-067-MFW and RR-TIR-19 are discussed in greater detail in Sections 6.1.9 and 6.1.10, respectively.
The BWOG investigation of MFW pump trips found that the root cause was listed as "not specified" or."unknown" in 40 percent of the cases.
This was the basis for recommendation TR-014-MFW, which calls for installation of a system to monitor the causes of MFW pump trips.
It is expected that such a system could provide the exact causes for MFW pump trips and eventually lead to plant modifications designed to reduce MFW pump trips and hence lower their contri-bution to the frequency of reactor trips.
Therefore, the staff agrees with the BWOG designation of TR-014-MFW as a key recommendation and encourages imple-mentation of recommendation TR-180-MTS concerning similar monitoring for the EHC system.
Steam generator rupture control systems and AFW/EFW initiation systems at B&W plants have been found to be susceptible to spurious actuations caused by perturbations / noise (e.g., main steam line pressure oscillations from turbine stop valve fast closure) that affects the associated steam generator level sensing instrumentation.
The BWOG has recommended the use of time delay cir-cuits or mechanical damping to reduce the impact of short duration pressure oscillations internal to the instrument sensing lines.
Recommendation TR-052-SFI addresses steam generator level sensing instrumentation, and rec-ommendations TR-170-MFW and TR-200-MTS similarly address MFW and EHC system oil pressure sensing instrumentation.
The staff has previously approved NUREG-1231, Supp. 1 6-37
similar techniques and considers them acceptable and beneficial to preventing spurious trips /actuations.
Item II.K.2.10, "Safety-Grade Anticipatory Reactor Trip," of NUREG-0737, "Clari-fication of TMI Action Plan Requirements," requires installation of a safety grade anticipatory reactor trip system (ARTS) to initiate reactor trip upon a loss of MFW and upon trip of the main turbine.
Recommendations TR-173-MFW, TAC-1.f, TAC-2.i, and RR-11 are related to the ARTS.
Recommendations TAC-1. f and TAC-2.i are for the BWOG I&C Committee to evaluate design modifications that would automatically reset the ARTS bistables following a reactor trip upon reset of the MFW pumps or main turbine.
In general, the concept of automatic reset of protective actions / functions is inconsistent with the intent of Sec-tion 4.16, "Completion of Protective Action Once It Is Initiated," of IEEE Standard 279"Criteria for Protection Systems for Nuclear Power Generating Stations." The design of a circuit modification to provide automatic reset of the ARTS bistables would have to be such that its failure would not prevent reactor trip on MFW pump trip or turbine trip when required.
The only approved recommendation in the RTS regarding the ARTS is TR-173-MFW, which calls for a procedure to be used to ensure the MFW pump trip ARTS bistables are properly reset when the MFW pumps are returned to service.
Using procedures to reset the ARTS bistables would seem to be adequate since there have been relatively few reactor trips caused by the failure to reset.
The staff discourages the pursuit of design modific'tions for automatic reset of the ARTS.
The BWOG has rejected recommendations RR-2.1.7 and RR-11 related to the reduc-tion of the frequency of reactor trips.
The staff believes that implementation of these recommendations would provide some benefit; however, not implementing these recommendations does not appear to significantly increase the potential for reactor trips.
Since it is not clear that the cost-benefit of implementa-tion would be worthwhile, the staff does not disagree with the BWOG's basis for rejection.
The staff strongly encourages implementation of recommendations TR-067-MFW and TR-071-MFW and agrees that recommendations TR-067-MFW and TAC-2.d are appropriate and should be implemented.
6.1.13 Conclusions The staff concludes that the BWOG I&C Committee established a sound methodology for review and evaluation of ICS/NNI designs at operating B&W reactors and that the methodology is appropriate for achieving resolution of ICS/NNI concerns.
The BWOG developed ICS/NNI system requirements (largely based on lessons learned from ICS/NNI operating history), compared the existing as-installed ICS/NNI designs to the system requirements, and proposed recommendations con-sidered necessary to achieve conformance of the installed designs to the sys-tems requirements.
Implementation of the recommendations is intended to achieve the desired reduction in reactor trip frequency and post-trip transient complexity that result from ICS/NNI-related failures at B&W piants.
The staff considers the development of upgraded / revised ICS/NNI systems require-ments to be a significant step of the ICS/NNI reassessment.
The staff was encouraged that some of the systems requirements take the form of acceptance criteria.
The staff concludes that, if existing ICS/NNI designs are modified to achieve conformance with the systems requirements, many of the problem areas i
NUREG-1231, Supp. 1 6-38
idei.tified from the investigation of B&W reactor transients involving ICS/NNI-related failures would be resolved.
However, the staff believes that improve-ments in the ICS/NNI systems requirements are needed in several areas and that development of ICS/NNI system design-basis documents by the BWOG would be useful to support the systems requirements documents.
Some of the recommendations resulting from the comparison effort are too general and may not refer to sufficient source documents (tnat provide the basis / background for the rec-ommendation) to ensure that implementation will satisfy the intended ICS/NNI reassessment program objectifes, or achieve conformance with the systems requirements.
Where appropriate, the BWOG should upgrade the references for the recommendations.
From the onset of the BWOG SPIP, through a series of meetings with the BW0G, and through a number of letters to the BWOG (the letters are listed in Appendix F),
the staff has emphasized that an objective of the SPIP reassessment should be to resolve the concerns identified from previous B&W reactor transients involving ICS/NNI-related failures.
The more significant concerns are listed in Appendix E.
The staff concentrated its review on resolution of the PICS because the loss of ICS power event at Rancho Seco on December 26, 1985, demonstrated that many of the PICS continued to exist despite past efforts to resolve them.
Many of the concerns have been found to be common to the more severe transients at B&W reactors involving ICS/NNI power losses.
The staff concludes that insuffi-cient information was provided in BAW-1919 for the staff to make a positive determination concerning resolution of the majority of PICS.
The BWOG provided responses to several of the staff letters concerning resolution of PICS in Appendix R-m, "Responses to NRC Questions Involving the ICS/NNI Evaluation,"
but many of the responses do not directly address the specific concerns and/or do not provide sufficient information for the staff to conclude that certain PICS have been resolved.
The staff impression was that the BWOG was focusing its efforts toward resolution of the PICS.
However, recent conversations wi th the BWOG indicate that it was never the intent of the SPIP to investigate the PICS because they were thought to have been resolved and to reinvestigate them was not deemed necessary.
The BWOG SPIP has made significant progress toward resolution of many of the PICS (see Section 6.1.11 of this report); therefore, the ICS/NNI reassessment program should prove successful in accomplishing its goal of achieving a reduction in reactor trip frequency and transient complexity.
The degree of reduction is difficult to assess.
It must be realized that the reassessment effort has resulted in an overall improvement in ICS/NNI designs and that the design and adequacy of B&W plant safety-related systems to mitigate the consequences of accidents and transients has not been questioned (with the exception of concerns regarding AFWS initiation and control, which are now i
considered to be resolved).
An important concept evaluated during the SPIP reassessment effort that has significant potential with regard to limiting the consequences of events involving ICS/NNI power losses, is that of "known safe state" (KSS).
The BWOG has identified recommendation TR-178-ICS, which recommends that each B&W plant ensure that a KSS is attained on the loss of ICS/NNI power, as a key recommen-dation approved for implementation.
Although it is not clear to the staff, based on a review of BAW-1919, that a KSS can be achieved following ICS/NNI power losses without reliance on substantial operator actions in excess of NUREG-1231, Supp. 1 6-39 i
those normally required for other reactor trips, the staff concludes that implementation of the approved and proposed ICS/NNI related recommendations will make it easier to attain a KSS following power losses.
The staff encour-ages the BWOG to continue to pursue actions to achieve the KSS as defined in BAW-1919.
Implementation of the recommendations to dictate equipment failure modes on loss of ICS/NNI power to reduce the consequences / complexity of such events and to retain remote manual control of ICS/NNI-controlled critical com-ponents from the control room following power losses are strongly encouraged by the staff.
The staff concurs with the approach to implement recommendations to reduce the consequences of ICS/NNI power losses as well as trying to prevent the losses by increasing power supply reliability.
ICS/NNI power losses have continued to occur despite past efforts to increase power supply reliability.
The Abstract to NUREG-1195 includes the following statement:
The fundamental causes for this transient were design weaknesses and vulnerabilities in the ICS and in the equipment controlled by that system. These weaknesses and vulnerabilities were not adequately compensated by other design features, plant procedures or operator training.
These weaknesses and vulnerabilities were largely known to SMUD and the NRC staff by virtue of a number of precursor events and through related analyses and studies.
Yet, adequate plant modifications were not made so that this event would be improbable, or so that its course or consequences would be altered significantly.
The information was available and known which could have prevented this overcooling transient; but in the absence of adequate plant modifications, the incident should have been expected.
Based on the review of information provided in BAW-1919, the staff concludes that in the absence of documented resolution of the PICS listed in Appendix E, some weaknesses and vulnerabilities in ICS/NNI designs probably continue to exist and losses of ICS/NNI power should still be expected to occur.
- However, in light of the apparent progress made in resolving some of the PICS, the staff concludes that the consequences of ICS/NNI power losses should not be as severe if the BWOG SPIP recommendations are implemented.
Areas of concern that the staff considers to remain unresolved upon completion of the SPIP include implementation of modifications that adequately address the concerns of IE Bulletin 79-27 ICS/NNI power supply monitor (PSM) design and application adequacy of backup instruments and controls following ICS/NNI power losses mid-scale failures of control room instrumentation operator actions required to achieve KSS following ICS/NNI power losses FSAR transient and accident analysis assumptions concerning the ICS/NNI The basis for the staff's conclusion that certain PICS remain unresolved and that a conclusion could not be reached concerning resolution of other PICS, is that the concerns were not addressed in sufficient detail in BAW-1919.
The BWOG considered several of the PICS to be plant specific (e.g., review of con-cerns identified in IE Bulletin 79-27); thus they were not addressed by the SPIP.
The staff concludes that plant-specific reviews and site audits should be performed to obtain the information necessary to address resolution of the PICS as part of the implementation audits discussed in Section 12.2.
The staff NUREG-1231, Supp. 1 6-40
I believes that with proper implementation of the BWOG recommendations, along with resolution of the concerns identified in IE Bulletin 79-27, the B&W plant re-sponse to ICS/NNI failures will be significantly enhanced and will ensure that adequate safety margins exist for B&W plants.
6.3 Auxiliary Feedwater/ Emergency Feedwater (AFW/EFW) System Review 6.3.3 Acceptability of Emergency Feedwater Initiation and Control System for Each B&W Plant In the SER, the NRC staff stated that it would address the acceptability of the AFW initiation and control system for each B&W plant in a supplement. The staff also will address details such as non-automatic response to steam and feedwater line breaks.
The NRC staff has completed its review of each B&W plant's AFW initiation and control system.
The review included the following plants:
Arkansas Nuclear One (AN0-1), Crystal River, Rancho Seco, Three Mile Island Unit 1 (TMI-1),
Davis-Besse, and Oconee Units 1, 2, and 3.
A general description of each dif-ferent type of AFW initiation and control system and its acceptability are dis-cussed below under the applicable B&W plant (s).
6.3.3.1 AN0-1, Crystal River, and Rancho Seco The emergency feedwater initiation and control (EFIC) system is the initiation and control system for AN0-1, Crystal River, and Rancho Seco.
The AFW/EFW sys-tem is designed to provide secondary coolant to the once-through steam gener-ators (OTSGs) if the main feedwater (MFW) system becomes unable to perform this function or if AFW is needed to promote natural circulation in the reactor cool-ant system.
The EFIC system is designed to automatically initiate AFW and to control the level in the OTSGs. When monitored plant parameters indicate the need for it, the system may be initiated manually at the discretion of the operator.
The EFIC system is designated as Class 1E equipment and is contained in cabinets that are physically and electrically isolated from one another.
Each channel receives analog inputs from the OTSG 1evel and steamline pressure transmitters associated with each OTSG.
The level signals are compensated, as necessary, to reflect true water level.
The EFIC system also receives initiation signals from the reactor protection system (RPS) and the safety features actuation sys-tem (SFAS).
During plant operation, the EFIC system constantly monitors these input signals.
An AFW initiation signal is generated when a process parameter exceeds its set point value, but an actual actuation will take place only if at least two channels issue initiating commands.
The EFIC system is designed to initiate AFW (1) on low water level in either OTSG, (2) on low pressure in either OTSG steam line, (3) when all four reactor coolant pumps trip (received from the RPS), (4) on loss of both MFW pumps at greater than 20 percent power (received as an anticipatory trip on loss of MFW pumps from the RPS), (5) on high reactor building pressure, or (6) on low reactor coolant system pressure.
The EFIC system receives high reactor build-ing pressure and low reactor coolant system pressure initiation signals from the safety features actuation system (SFAS).
High building pressure is used to NUREG-1231, Supp. 1 6-41
initiate AFW because it indicates a possible main feed line break, a main steam line break, or a loss-of-coolant accident inside containment.
The EFIC system actuation logic is arranged in a 1-out-of-2, taken twice logic.
All four EFIC channels provide AFW initiation commands to the AFW trip logic modules.
These trip logic modules are physically located in the A and B EFIC channel cabinets.
The EFIC system includes logic used to isolate AFW flow to a ruptured or depressurized 0TSG.
This logic is referred to as the feed only good generator (F0GG) or "vector" logic.
Upon actuation, the vector logic precludes the con-tinued addition of AFW to a depressurized 0TSG; thus minimizing the overcooling effects of a steam leak.
The vector logic may isolate AFW to one OTSG only, never to both.
The EFIC system also will isolate MFW flow to an 0TSG when either a pressure of less than 600 psig or a high water level is detected in that OTSG.
Four redundant instrument channels, A, B, C, and D, are provided to monitor each of these parameters for each OTSG.
The EFIC system MFW isolation logic is arranged in a 1-out-of-2 taken twice logic identical to the AFW system actua-tion logic.
In the event of an AFW initiation, the operator may gain control of the system pumps and isolation valves by pressing the Manual Permissive button for the respective AFW train (A or B).
This allows the operator to take manual control oi the actuated equipment to optimize response to the existing condition, that is, shut down pumps and control valves, as required to handle the situation.
If the condition clears that caused the initiation while operating in the manual mode after an AFW initiation, the system will revert back to the automatic mode of operation and will re-initiate AFW if a condition occurs that requires it.
The AFW trains can be manually actuated by the operator pressing both AFW Initi-ate buttons on the control panel.
The system will respond in exactly the same way as if it were automatically actuated.
A more detailed description of the AFW EFIC system, as well as plant-specific information (e.g., pump and valve nomenclature and set points) is contained in the respective NRC staff's plant-specific SERs.
The staff has reviewed and evaluated the AFW EFIC system for AN0-1, Crystal River, and Rancho Seco, and has concluded that each is acceptable.
Plant-specific SERs were issued to each licensee in letters dated July 13, 1982, September 17, 1982, and November 10, 1987, respectively.
6.3.3.2 TMI-1 The AFW initiation system at THI-1 is designed to (1) initiate the start of the AFW system's two motor-driven pumps and one turbine-driven pump, which provide the in-flow of water to the OTSGs; (2) initiate automatic control to maintain the required water level in each OTSG; and (3) provide a manual control capa-bility.
The system is capable of withstanding a design-basis event and a NUREG-1231, Supp. 1 6-42 J
single active failure while supplying a heat removal path to allow safe shut-down of the reactor.
The design modification also ensures that a single active failure will never inadvertently initiate the AFW system nor isolate the MFW system.
The licensee performed these modifications before startup following the Cycle 6 refueling outage, which occurred in November 1986.
The initiation and control of the AFW system is performed by the heat sink pro-tection system (HSPS).
The HSPS is designed to initiate the AFW system on low water level in either OTSG; on high containment pressure, when both main feed-water pumps trip; or when all four reactor coolant pumps trip.
Indication of high containment pressure provides an alert of a possible main feed line break, a main steam line break, or a loss-of-coolant accident inside containment.
This indication, along with an indication of low water level in the OTSGs, would indicate a possible failure of the secondary heat sink.
The initiating signals for the main steam line rupture detection circuitry portion of the AFW system originate in the HSPS.
The HSPS is designed to initiate MFW and AFW isolation on low pressure in either 0TSG main steam line supply to the turbine ge%rator while supplying water to the "good" generator.
The sensing portion of the HSPS for steam generatoc level and containment pres-sure consists of four independent channels, with a two-out-of-four actuation logic.
This arrangement is designed to withstand a single failure in any sirgle channel simultaneously with another channel bypassed for maintenance, test, or repair.
The actuation portion of the HSPS electronics consists of two indepen-dent trains.
The trains are designed such that a single failure of either train will not prevent at least one train of the AFWS from operating.
The automatic initiation signals from the sensors to the control circuitry and the control circuitry for the AFW system at THI-1 comply with the general func-tional requirements of IEEE Standard 279-1971 with regard to the requirement to automatically initiate appropriate protective action whenever a condition moni-tored by the system reaches a preset level.
The main steam line rupture detec-tion circuitry associated with the HSPS was reviewed to the extent necessary to ensure proper separation and isolation from the AFW system initiating circuitry portion of the HSPS.
The automatic initiation circuitry from the sensors to the control circuitry and the control circuitry of the AFW system was reviewed to determine compliance with the single-failure criterion of IEEE Standard 279-1971.
The criteria of paragraph 4.17 of IEEE Standard 279-1971 specify that the protection system shall include means for manual initiation of each protective action at the sys-tem level.
The only system level protective action for the AFW system is AFW injection, which requires starting the AFW system's pumps and opening the control valves.
This action can be performed by overriding the automatically selected set points at each of the AFW system's controller display stations in the con-trol room.
By switching the controllers to manual control, the operator can initiate the automatically initiated circuitry, which starts the AFW pumps and opens the control valves.
These controllers and appropriate indicators are physically located next to each other in the control room so that the operator can take control quickly and ef ficiently with the minimum number of operations.
The staff's review of the AFW injection design provided assurance that the present means of manual initiation meets the criteria of paragraph 4.17.
NUREG-1231, Supp. 1 6-43
The system design at THI-1 provides for four separate automatic AFW initiation signals.
Three of these signals meet all the requirements of IEEE Standard 279-1971.
TMI-1 requested a deviation from IEEE-279 with regard to the fourth signal, for which manual removal of a bypass of the signal was proposed.
For this fourth signal, the licensee believed it would be preferable to have the results of operating experience at low steam generator levels before estab-lishing the set point level that would trigger automatic initiation of the AFW system.
This would prevent numerous inadvertent initiations of the AFW system.
Therefore, the licensee proposed a design in which automatic initiation of the AFW system on low level was bypassed when the plant was operating below 30 per-cent power, in which case it was proposed that the system would be actuated manually.
The licensee also proposed that this bypass be removed manually when the plant exceeded 30 percent power.
In a letter dated February 18, 1987, the NRC staff told the licensee that manual removal of the bypass was not acceptable because it did not meet the guidelines of IEEE Standard 279-1971, which call for automatic removal of operating by-passes when permissive conditions are not met.
In addition, the licensee had not provided adequate justification for the system proposed.
However, as an interim resolution to this problem, the NRC staff, by issuance of License Amendment No.124, has permitted manual removal of the bypass until the end of Cycle 6 operation.
At that time, the AFW system will be modified to delete manual removal of the bypass of the fourth initiation signal and will be ir, full compliance with IEEE Standard 279-1971, the licensee has provided ade-quate compensating features to ensure that bypass removal will function properly.
The licensee agreed to implement design and procedural controls to ensure that the bypass switch is in the appropriate position.
First, when the bypass is actuated, an alarm indicates in the control room.
Second, the control room operator must acknowledge, by signature, that the bypass status has been checked and that the bypass has been removed when the reactor power is increased above
[
30 percent power.
In approving this design in Amendment No. 124, the NRC staff imposed additional restrictions on the licensee.
Use of this manually removed bypass is restricted to the control'sd conditions of a normal reactor startup or shutdown during 1
Cycle 6 operatun only.
Thus, this bypass will be used very infrequently and only during a very short period (hours) following a decision to either start up or shut down the unit.
Operation at power levels above 30 percent is the normal i
and more frequent mode of operation, and the bypass is always removed during I
this mode of operation.
For this limited period, the use of these special pro-cedural controls is adequate to satisfy the basic purpose of NUREG-0737.
The NRC safety evaluations and inspections indicate that a safety grade, highly reliable AFW system has been installed and is operational at THI-1.
A system that fully meets the guidelines of IEEE Standard 279-1971 will be provided by the end of Cycle 6.
During Cycle 6, compensating features to satisfy the intent of NUREG-0737 will be provided.
l 1
NUREG-1231, Supp. 1 6-44
On the basis of the NRC staff's review of the THI-1 EFW automatic initiation system, the licensee's response to the requirements of Item II.E.1.2 is adequate to determine compliance to the criteria of IEEE Standard 279-1971 and, therefore, is acceptable 6.3.3.3 Davis-Besse The AFW system at Davis-Besse is desi0ned to supply an independent source of feedwater to the steam generators when the normal feedwater system is not avail-able to maintain the heat sink capabilities of the the steam generators.
The primary sources of water for the AFWs are the two, non-safety-related condensate storage tanks (CSTs).
The secondary source of water for the turbine-driven AFW pumps is the safety grade service water system with an automatic switchover from the CST on low suction pressure at the pumps.
The AFW system is an engi-neered safety feature system that is relied upon to aid in preventing core dam-age in the event of transients such as a loss of normal feedwater, a steam sys-tem pipe rupture, or a small-break loss-of-coolant accident.
The system consists of two redundant safety-related essential trains, each with its own steam-turbine-driven pump, associated valves, piping, controls, and instrumentation.
A non-safety-related, motor-driven feedwater pump (MDFWP),
associated valves, piping controls, and instrumentation also ate able to provide flow equivalent to one AFW pump to either steam generator.
Each of the AFW pumps is capable of supplying water to either or both steam generators, as is the MDFWP.
Each AFW pump has a design flow of 1050 gpm (which includes 250 gpm minimum recirculation) at 1050 psig.
One turbine-driven AFW system train is completely independent of ac power.
Each of the AFW supply paths (including the MDFWP) to the steam generator contains two check valves and a motor-operated isolation valve.
The flowpath from the MDFWP includes two check valves and a flow control valve.
On initiation of the AFW system, the isolation valve will be automatically opened.
The turbine-driven AFW pumps are initiated on low steam generator level, low steam generator pressure, loss of the four reactor coolant pumps, high steam generator level, and high differential pressure between the main feedwater line j
and steam line.
Manual initiation is accomplished by operator action in the control room.
The MDFWP is operated from the control room with non-safety grade controls and instrumentation.
The MDFWP and its associated equipment are normally aligned to receive power from one diesel generator.
Operating this equipment with power from the other diesel generator requires some operator action from outside the control room.
The licensee has committed to making the necessary modifications to permit aligning all necessary MDFWP-associated equipment to either diesel generator from within the control room before Cycle 6.
The steam and feedwater rupture control system (SFRCS) is designed as an engi-neered safety features system to monitor plant parameters such as steam gener-ator water level and pressure, differential pressure between the steamline and main feedwater line for each steam generator, and the loss of all four reactor coolant pumps.
Under plant conditions indicative of a main steamline break,
{
main feedwater line break, or loss of heat sink, the SFRCS initiates appropriate actions to isolate a ruptured steam generator and initiate AFW system flow to NUREG-1231, Supp. 1 6-45
the intact steam generator (s).
Valves controlled by the SFRCS to isolate a ruptured steam generator include the main steam isolation valves (MSIVs), the MFW regulating startup valves, and the AFW system containment isolation valves.
The SFRCS also controls the AFW system steam admission and pump discharge valves.
The NRC staff's preliminary review of the SFRCS design, following the event on June 9, 1985, concluded that the SFRCS was unacceptable because it was not capa-ble of performing its required safety functions (providing AFW flow to the intact steam generator) following a design-basis event and a single active failure.
Furthermore, the staff raised concerns regarding the SFRC's capability to cut off all sources of feedwater to both steam generators, requiring operator inter-vention and successful operation of several active components to re establish core cooling.
The licensee performed a single-failure analysis of the SFRCS to crisure that for each analyzed event, given any credible active single failure, AFW would be available to the intact steam generator.
On the basis of the results of the licensee's analysis and the short-term modifications to the SFRCS to resolve the single-failure concerns identified in NUREG-1154 with respect to reopening an AFWS containment isolation valve to feed an intact steam generator, the staff concluded that the design of the SFRCS was acceptable to allow plant restart.
The staff also concluded that the short-term modifications to the SFRCS were sufficient to resolve staff concerns regarding SFRCS isolation of all sources of feedwater to both steam generators.
The NRC staff's SER, NUREG-1177, dated June 1986, delineates the SFRCS and the staff's conclusion in more detail.
6.3.3.4 Oconee i
The AFW system at Oconee consists of two motor-driven pumps and one turbine-driven pump.
The motor-driven pumps and the turDine-driven pump start auto-t matically on low main feedwater pump discharge pressure or low main feedwater pump control oil pressure.
The AFW system is capable of feeding to either or both steam generators under automatic or manual initiation and control.
The automatic initiation signals and circuitry for AFW at Oconee comply with the general functional requirements of IEEE Standard 279-1971 with regard to the requirement to "automatically initiate appropriate protective actions when-ever a condition monitored by the system reaches a preset level." The use of 1
low main feedwater pump disenarge pressure or low main feedwater pump control 1
oil pressure signals in the AFW system initiation logic represents an antici-patory loss of feedwater.
This design feature permits automatic initiation of emergency feedwater in a more timely manner to reduce the likelihood of steam generator dryout.
The anticipatory loss of feedwater initiation is detected by pressure switches sensing diverse variables.
Unlike the other B&W plants, Oconee has no automatic features for isolation of a depressurized steam generator in the event of a main steam or feedwater line break.
Following such events, operator action must be taken to isolate MFW or AFW to the affected generator and to ensure adequate AFW to the intact steam generator.
These actions can be accomplished by manual action within the con-trol room.
Emergency Operating Procedure EP/1/A/1800/01 directs the control room operator in the necessary steps to isolate the affected steam generator.
NUREG-1231, Supp. 1 6-46 l
.-1
---7r
---m i--r
+ - ---
--+---w--
--v v
T-'
In response to IE Bulletin 80-04, issued February 7, 1980, analyses have been performed for Oconee, which assume blowdown of both steam generators prior to operator action being taken and no ICS action to isolate the MFW.
These anal-yses confirm that core cooling will be maintained and that the main steam and feedwater line break consequences, such as peak containment pressure and reactor vessel overcooling, are maintained within acceptable l!mits.
Thus, the staff concluded that automatic isolation of the steam generator is not required.
Further details and the bases for the staff's conclusioni are provided in the staff's October 14, 1982 safety evaluation of Oconee (letter to H. B. Tucker, Duke Power, from J. Stolz, NRC) and in Chapter 15.3 of the Oconee FSAR.
The automatic initiation signals and circuitry used at Oconee comply with the single-failure criterion of IEEE Standard 279-1971.
The scope of the single-failure analysis was limited to the emergency feedwater initiation circuitry, electrical power sources, and control systens.
No single failure within the manual or automatic initiation systems prevents initiation of protective action by manual or automatic means.
As stated in the NRC letter dated June 3,1981, the NRC staff concluded that NUREG-0737 Item II.E.1.2 had been adequately resolved for Oconee.
6.3.3.5 Conclusion The staff has reviewed and evaluated each B&W plant for conformance to NUREG-0737, Item II.E.1.2, Parts 1 and 2, "Auxiliary Feedwater System Initiation and Flow Indication," and concludes that each is acceptable.
In addition, the staff concludes that the emergency feedwater system does not rely on the non-safety grade ICS to fulfill its safety function.
l 1
J l
NUREG-1231, Supp. 1 6-47
i 7 EVALUATION OF HUMAN FACTORS ISSUES 7.7 Response to BWOG Comments on SER In its letter dated December 21, 1987, the BWOG provided comments to the NRC staff on the staff's SER.
The NRC staff has evaluated these comments and de-termined that, with the exception of one comment pertaining to the further evaluation of the six identified concerns related to the human factors (con-tained in Section IV of Appendix K of BAW-1919), the BWOG has not provided any new or additional information that was not known or factored into the staff's review at the time of the SER was written.
Based on the BWOG letter, it appears that the staff and the BWOG are in mutual agreement that human factors are important.
The BWOG has acknowledged this by its commitment to consider human factors in the future.
However, whereas the staff believes that human factors are sufficiently important to warrant the participation of personnel trained and experienced in the discipline of human factors engineering in multidisciplinary teams of investigators, the BWOG takes the position that it is sufficient for engineers to consider human factors in the course of their traditional activities.
Although the BWOG has committed to the future consideration of human factors, adequate details were not provided for the staff to assess the acceptability the BWOG actions.
Therefore, the BWOG's response to the SER has not changed the staff's evaluation of the BWOG efforts in the area of human factors as described in Section 7 of the SER.
i NUREG-1231, Supp. 1 7-1
9 RISK ASSESSMENT 9.6 Risk Recommendations / Relationship to SPIP As part of the BWOG Safety and Performance Improvement Program (SPIP), the Owners Group has recommended a number of plant modifications intended to reduce the frequency and severity of transients at B&W plants.
These recommendations are documented in Appendix J to the SPIP report (BAW-1919, Revision 4, April 1988).
In the SER these recommendations were assessed to identify those that would help to reduce the frequency of core damage resulting from the Category C events.
The NRC staff also has examined the potential risk benefit achievable from selected recommendations; this is discussed below.
Since the details of the actual plant implementation of the recommendations are not known, the actual plant-specific benefit cannot be estimated.
Rather, the maximum possible risk benefit was estimated assuming that the recommendation was properly developed and implemented.
This helps provide a perspective on what may be the more effec-tive areas for action and potential benefit.
Since the time of this evaluation, Appendix J has been updated to include addi-tional recommendations approved by the BWOG and to identify recommendations that have been rejected or superseded.
These recent updates to Appendix J have not been factored into this evaluation.
9.6.1 Areas Covered by the BWOG Recommendations Each of the 154 BWOG recommendations identified in Revision 4 of the SPIP report was reviewed and separated into one of the following 12 categories identified as being meaningful to the evaluation.
(1) improvements affecting integrated control system /non-nuclear instrumentation (2) improvements affecting main feedwater (3) improvements affecting instrument air (4) impron ments in plant operations (5) improvements affecting main steam (6) improvements affecting plant electrical supply (7) improvements affecting motor-operated valves j
(8) improvements affecting plant administration l
(9) improvements affecting the main turbine systems (10) improvements affecting main steam /feedwater isolation (11) improvements affecting emergency feedwater (12) improvements affecting the reactor protection system Table 9.1 of this supplement identifies the specific recommendations discussed in this evaluation by number and title, as assigned in the SPIP report, and category.
9.6.2 Category C Core Damage Reduction Potential of the BWOG Recommendations 1
In this section, the staff identifies the BWOG recommendations that could poten-tially reduce the contribution of Category C events to core damage frequency.
NUREG-1231, Supp. 1 9-1
l Based on the results presented in Section 9 of the SER, it is possible to iden-tify the improvement areas that will most likely produce reasonable reductions.
It is very important to note that only the effect of these improvements on the Category C core damage frequency is evaluated in this section.
Other benefits that may be accrued to other parts of the risk profiles of these B&W plants are not within the scope of this study to consider, nor are they even possible to evaluate using the models or information compiled in the analysis conducted within the context of this study.
The first area to be considered is failures in the integrated control system (ICS).
In Section 9 of the SER, the staff illustrated that these failures were the dominant contributor to core damage from Category C events in the B&W plants studied.
Further, in looking at just what are the causes for these failures being so dominant, it was clear that reasonable core damage reductions could not be achieved without making improvements in the ICS area.
To summar-ize the results for the four plants where Category C events can be considered to be significant contributors to core damage frequency, the contribution from loss of ICS versus total Category C contribution is as follows:
Crystal River 3:
1.3E-5 out of 2.5E-5, or about 50%
Davis Besse:
8.4E-6 out of 1.2E-5, or about 70%
ANO-1:
1.7E-5 out of 3.0E-5, or about 55%
Rancho Seco:
1.0E-5 out of 1.5E-5, or about 65%
For the two plants where Category C events are not considered to be significant contributors to care damage frequency, the results are as follows:
Oconee:
4.6E-6 out of 6.0E-6, or about 75%
Three Mile Island:
5.7E-7 out of 1.9E-6, or about 30%
The reason for these results is twofold:
(1) ICS failure causes loss of main feedwater and also may induce other effects on the secondary systems that increase the severity of events, and (2) ICS failure causes substantial instru-ment upset that, even though backed eo by safety grade instrumentation, can result in a substantial amount of ope"ation confusion, increasing the chance of operator error.
The BWOG recommendations include a large set of improvements in the ICS area; Table 9.1 lists those that, if properly implemented, could have a significant effect on reducing core damage from ICS upset events.
In particular, if recom-mendations TR-001-ICS, TR-002-ICS, TR-004-ICS, TR-013-ICS, TR-038-ICS, TR-102-ICS, and TR-104-ICS were implemented, they will have the most significant effect on reducing the number of plant trips resulting from failures in the ICS (a signi-ficant reduction in initiating event frequency).
Recommendations TR-033-ICS and TR-036-ICS will reduce the severity of the effect of the loss on plant systems (such as main feedwater).
Recommendation TR-035-ADM, if effectively implemented, j
will largely reduce the operator confusion problem for the remaining occurrences of total system upset.
In addition, TR-017-MFW will enhance feedwater avail-ability during loss of ICS and TR-060-0PS will further enhance operator response during loss of ICS.
NUREG-1231, Supp. 1 9-2
The combination of these effects could be such that the loss of ICS core damage contribution, presented above, would be made so small as to no longer be a dominant contributor to the Category C core damage frequency.
The maximum amount of the reduction would, for all intents and purposes, be the total per-centage contribution shown above for the present plant configurations, meaning that the expected Category C frequencies would be:
Crystal River 3:
1.2E-5 Davis Besse:
3.6E-6 ANO-1:
1.3E-5 Rancho Seco:
5.0E-6 Oconee:
1.4E-6 Three Mile Island:
1.3E-6 Thus, in addition to Oconee and Three Mile Island, two more plants (Davis Besse and Rancho Seco) would have the contribution of Category C events to core damage frequency reduced to the level of not being significant by effectively imple-menting those recommendations discussed above.
Following implementation of the ICS recommendations, the next logical area to consider for possible plant improvement (based on staff and contractor results) is in the reliability of main feedwater and/or bleed-and-feed capability.
Failure of these functions also was illustrated in the SER to be a dominant contributor to core damage from Category C events.
For the two plants where Category C events would still be considered to be significant contributors to core damage frequency, the contribution from total loss of feedwater and bleed-and-feed versus total Category C contribution, assuming the ICS recommendations are imolemented in a manner to achieve minimum Category C frequency, is as follows; Crystal River 3:
4.2E-6 out of 1.2E-5, or about 35%
ANO-1:
1.0E-5 out of 1.3E-5, or about 75%
For the four plants where Category C events now would not be considered to be significant contributors to core damage frequency, the results are as follows:
Davis Besse:
3.6E-6 out of 3.6E-6, or about 100%
Rancho Seco:
3.3E-6 out of 5.0E-6, or about 65%
Oconee:
4.9E-7 out of 1.4E-6, or about 35%
Three Mile Island:
5.6E-7 out of 1.3E-6, or about 45%
The reason for this contribution to core damage is threefold:
(1) a high over-all frequency of loss of main feedwater events plus (2) a relatively high human error probability for failure to properly respond, and (3) a relatively low prob-l ability of recovering feedwater once it has failed.
Obviously, the contributions given above constitute the maximum possible reduction in core damage frequency if it were possible to eliminate these total loss of cooling sequences.
There are a number of BWOG recommendations (identified in Table 9.1) that would enhance performance in the three areas mentioned above, most of which pertain to the main feedwater system.
The particular recommendations with the greatest potential for reducing core damage are TR-014-MFW, TR-015-MFW, TR-066-MFW, NUREG-1231, Supp. 1 9-3
TR-074-MFW, and TR-052/053/054-SFI with respect to the high frequency of loss of main feedwater events.
With respect to the low probability of feedwater recovery, recommendations TR-070-MFW and TR-071-MFW (for main feedwater recovery) and TR-055/056/057-ADM (for the emergency feedwater recovery) appear to be poten-tially the most effective.
Recommendations TR-018-MFW and TR-067-MFW would be i
effective to both reduce the frequency of loss of main feedwater events and en-hance main feedwater recoverability.
In the area of operator response to total loss of feedwater, recommendations TR-060-0PS and TR-064-0PS would enhance oper-ator response in recognizing the need to initiate bleed and feed and recognizing the need to recover EFW, respectively.
In addition to these recommendations, the modification of the Davis-Besse bleed-and-feed capability, which was dis-cussed in Section 5.3.8 of the SER, is deemed to be essential to enhance bleed-and-feed reliability at that plant.
The implementation of these recommendations could potentially result in a.com-bined reduction of an order of magnitude in the frequency of core damage result-ing from these total loss of cooling event sequences.
After requantifying the model for the suggested modifications, the minimum core damage frequency of Category C events for each plant is calculated to be:
Crystal River 3:
8.6E-6 Davis Besse:
5.0E-7 ANO-1:
3.0E-6 Rancho Seco:
1.8E-6 Oconee:
9.9E-7 Three Mile Island:
8.0E-7 l
Thus, Category C events will become insignificant contributors for all B&W plants.
As noted above, these conclusions reflect the maximum risk effectiveness of the i
selected BWOG recommendations.
Details of implementation will determine how much, if any, of this benefit will be realized at a specific plant.
For example, I
the design features of Davis-Besse are such that the potential improvement cited above may not be fully achievable. At present, deficiencies in bleed and feed manifest themselves as beneficial to preventing transient-induced loss-of-coolant accidents (LOCAs).
If bleed-and-feed capability is improveu as part of the modifications, it is likely that some of this benefit will be lost to an in-creased core damage frequency resulting from a LOCA, although to what extent is not possible to quantify in the absence of more detailed information on the new Davis-Besse bleed-and-feed capabilities.
The final result will probably fall more in the 2E-6 range rather than SE-7.
These results do, however, indicate that many of the recommendations have been well directed towards responding to the operational and design features of B&W plants that have generated concern.
Thus, the staff concludes that proper implementation of the recommendations will increase safety of B&W plants by decreasing the overall plant risk.
1 4
NUREG-1231, Supp. 1 9-4
Table 9.24 Recommendations that reduce the frequency of core damage from Category C events Category /
Recommendation Number Recommendation ICS/NNI System Recommendations:
TR-001-ICS Replace reactor coolant (RC) flow signal input to integrated control system (ICS) with equivalent signals based on RC pump status.
TR-002-ICS Eliminate plant transients and trips due to a single failure of a T and T signal.
Implement a modification to hot cold automatically detect invalid RC temperature input to ICS.
TR-004-ICS Implement a modification to automatically detect an invalid I
input to ICS of turbine header pressure.
TR-013-ICS Install the necessary equipment to prevent loss of i 24-volt power to the ICS or non-nuclear instrumentation (NNI) result-ing from the loss of a tingle power source.
TR-033-ICS Make appropriate changes to ensure that plant will go to a known, safe state without any operator action required on loss of ICS/NNI power.
TR-036-ICS Evaluate turbine bypass valve position on loss of ICS.
1 TR-038-ICS Develop and implement a recommended preventive maintenance program for ICS/NNI.
j TR-102-ICS Install redundant de power supplies for NNI-Y Arkansas Power
& Light Co. (AP&L) only.
TR-104-ICS Incorporate automatic selection of valid input signals for ICS/NNI.
Main Feedwater System Recommendations:
i TR-014-MFW Install a monitoring system in main feedwater pump (MFWP) trip circuitry to document the primary causes of MFWP trips.
TR-015-MFW Determine if a low suction pressure trip is needed.
Then l
1 decide what trip or response to low suction pressure should be implemented.
1 TR-017-MFW Evaluate the MFWP control systems and their interaction with the ICS.
Implement a program to identify improvements netded in both control systems.
i NUREG-1231, Supp. 1 9-5 t
Table 9.24 (Continued)
Category /
Recommendation Number Recommendation Main Feedwater System Recommendations (Continued)
TR-018-MFW Provide training to operators and maintenance personnel and ensure procedures are adequate for line-up, operation, and maintenance of main feedwater (MFW) system components.
TR-066-MFW Check all MFW and condensate system protective circuits, interlocks, motors, etc., to ensure that a single electri-cal failure will not cause a loss of both feedwater trains.
TR-067-MFW Evaluate the set point and functions of the automatic MFWP trip features.
Wherever possible, eliminate these trip functions altogether.
TR-070-MFW Provide capability to override a close signal to the MFW block valve to enable the control room operator to stop the block valve at e.ny intermediate position during valve closure.
TR-071-MFW install valve oosition indication for the startup and MFW regulating valves (and low load control valves at applicable i
plants).
r TR-074-MFW Schedule instrumentation and control (I&C) calibration and inspection work so as to minimize the number of times the main FW pumps and turbines I&C equipment is disturbed during power operation.
Plant Operations Recommendations:
TR-060-0PS Stress in operator training that emergency operating proce-dures are to be followed explicitly even when such procedures are considered as drastic actions.
TR-064-0PS Operator training to reset turbine-driven emergency feedwater (EFW) pumps after overspeed trips should be part of formal training programs and should include hands-on training.
Plant Administration Recomendations:
TR-035-ADM Familiarize operators with Rancho Seco event.
TR-055-ADM Coordinate activities of plant operations, security and radcon (health physics) personnel to facilitate timely access to critical equipment.
TR-056-ADM Move chain link fences as necessary, to provide better access to critical components, NUREG-1231, Supp. 1 9-6
. ~. _ _
l 4
Table 9.24 (Continued)
Category /
Recommendation Number Recommendation Plant Administration Recommendations (Continued)
TR-057-ADM Consider ways to improve access to critical components where problems have been identified with gaining access to critical components because of 10 CFR Appendix R fire barriers.
Main Steam /Feedwater Isolation System Recommendations:
TR-052-SFI AP&L, General Public Utility Corp. (GPUN) and Sacramento 6
Municipal Utility District (SMUD) need to filter their steam generator level signals in the steam feedwater rupture control system.
TR-053-SFI Correct overheating problems that can lead to electric power supply malfunctions and correct problems caused by degraded voltage power supplies (AP&L, GPUN, and SMUD).
j TR-054-SFI Redesign pneumatic hardware for main steam isolation valves
]
to ensure this equipment is exercised during surveillance testing (AP&L only).
f f
i 5
l 2
i 1
NUREG-1231, Supp. 1 9-7
10 REACTOR TRIP INITIATING EVENTS REVIEW I
In the SER, the NRC staff reported that its evaluation of the recommendations i
i contained in the BWOG report, "Review of Reactor Trip Initiating Events at the BWOG Plants 1980-1986" (Report 47-1168891-00, September 1987), would be provided in a supplement to the SER.
This supplement documents the staff's review and j
supersedes the original evaluation contained in the SER.
i i
10.1 Description of Reactor Trip Initiating Events Review i
As part of the SPIP, the BWOG Transient Assessment Committee reviewed reactor trip events that occurred at B&W plants during 1980-1985.
The committee per-formed a quick sort of the BWOG Transient Assessment Program (TAP) data to identify major trip initiators and a detailed review of Category B and C events, which emphasized post-trip response.
The results of these BWOG efforts are discussed in Section 4 of the SER.
e To further the BWOG efforts to reduce the number of reactor trips, the BWOG Transient Assessment Committee reviewed the 235 reactor trip events, focusing l
on the initiating events that resulted in reactor trips at B&W plants from i
i January 1,1980, to July 31, 1986.
The BWOG transmitted the "Review of Reactor i
Trip Initiating Events at the B&WOG Plants 1980-1986" to the staff in a letter dated October 1, 1987.
f
]
The BWOG reviewed each reactor trip to determine which event initiated the j
sequence of events that ultimately resultad 5 a iea uor trip.
These initiat-
~
ing events were grouped into the sis t;ategories listed below.
(1) turbine system a
(2) feedwater systems l
(3) integrated control systems (ICS)
(4) reactor coolant pumps (RCPs)
(5) control rod drive (CRD) system (6) other In addition, each reactor trip was examined to determine what trip function (e.g., high reactor coolant system (RCS) pressure) actually tripped the reactor and what inherent or design features should have prevented, but did not prevent, t
the reactor from tripping.
This allowed each event to be reviewed from two t
perspectives:
(1) prevention of the initiating event and (2) avoidance of the l
reactor trip by preventing or limiting the RCS response to the initiating event.
)
The B&W utilities have completed many corrective actions directed toward pre-
{
venting trips.
The BWOG review evaluated the effectiveness of these completed 4
actions.
This review also assessed the potential effectiveness of implementa-l tion of the recommendations developed from the SPIP.
On the basis of this
{
assessment, additional recommendations were developed to reduce the reactor j
trip frequency at B&W plants.
i NUREG-1231, Supp. I 10-1 i
1 l
10.2 Conclusions From BWOG Review On the basis of its review of the reactor trip events, the BWOG concluded that corrective actions taken by the B&W plants have been effective in reducing the reactor trip frequency by nearly 40 percent (1984-1986 average versus 1980-1983 r.ve rage).
Additionally, 41 recommendations in the RTS were identified as app 11-cable to reducing the frequency of initiating events leading to reactor trips.
Nevertheless, the BWOG concluded that additional efforts are necessary to fur-ther reduce the frequency of reactor trips and developed 36 additional recommen-dations.
The specific recommendations developed and the NRC staff's assessment of these recommendations are provided in Section 10.3 of this supplement.
Other general conclusions from the study include:
(1) The turbine system was the leading trip initiator during 1980 through 1986, although corrective actions taken by the B&W plants have reduced the turbine trip frequency by two-thirds since 1983.
Eight new recommendations have i
been proposed to further reduce turbine trip initiators.
(2) The feedwater system was the second leading trip initiator during 1980 through 1986.
From 1984 through 1986, main feedwater (MFW) upsets, initi-ated by either the MFW system itself or in response to ICS failures, have accounted for over 50 percent of the reactor trips.
Although numerous corrective actions have been implemented at the B&W plants, the frequency of MFW upsets has not declined.
The BWOG expects that, with the implemen-tation of the recommendations in the RTS, the frequency of MFW-initiated reactor trips will decline.
However, an additional 11 recommendations were proposed for the MFW system.
(3) The ICS was the third leading trip initiator, primarily resulting from the loss of signal inputs.
Although the RTS includes numerous recommendations related to the ICS, five new recommendations have been developed for this system.
(4) The control rod drive (CRD) system was the fourth leading trip initiator and corrective actions already taken by several of the B&W plants have been effective in reducing its contribution.
Nonetheless, five new rec-ommendations have been developed to further reduce CRD system initiated trips.
1 (5) The RCPs were the fifth leading trip initiator, primarily as a result of spurious actuations of the reactor coolant pump power monitors (RCPPMs).
Corrective actions already taken at B&W plants have effectively eliminated these events and no additional recommendations were deemed necessary, (6) The high RCS pressure trip was the leading RPS trip function accounting for 59 percent of the RPS trips.
These trips were primarily the result of feed-water upsets, loss of ICS inputs, and MFW trips.
While specific recommen-l dations were developed to decrease these initiators, additional recommen-l dations were made to limit the RCS thermal-hydraulic response to these initiators and thereby reduce the frequency of high pressure RCS trips.
4 4
i 1
NUREG-1231, Supp. 1 10-2 l
4
1 10.3 BWOG Recommendations and NRC Staff Findings 10.3.1 BWOG Recommendations Related to Reactor Trips Initiated by the Turbine System As a result of its review, the BWOG developed the eight recommendations listed below relating to reducing reactor trips initiated by the turbine system.
(1) Plants with General Electric turbines should evaluate bypassing the high-vibration trip during stop-valve testing.
(2) Plants should install a time delay relay, or an orifice, between the elec-trohydraulic control (EHC) oil system and the anticipatory reactor trip system (ARTS) sensing line, (3) The BWOG Instrumentation and Control (I&C) Committee should review the EHC overspeed and fast control and intercept valve circuits to determine why l
they are so frequently actuated and how they can be corrected to prevent recurrence.
(4) Protective covers should be placed over local level / trip switches, the inadvertent actuation of which can directly result in turbine or reactor trips.
(5) Procedures governing plant startup should be verified to ensure they have sufficient instructions for properly resetting anticipatory reactor trip system (ARTS) bistables/ contact buffers.
(6) The BWOG I&C Committee should evaluate a design modification for automat-ically resetting the ARTS bistables on turbine reset.
(7) The turbine trip runback should be included in the operator training program.
(8) Integrated control system (ICS) tuning should be reviewed to ensure con-trol settings are compatible with a turbine trip runback from less than 45 percent power.
Recommendations 1, 2, and 3 above were evaluated in Section 8.3 of the original SER and are already entered into the BWOG Recommendation Tracking System (RTS) l
]
as TR-169-MTS, TR-200-MTS, and TR-201-MTS, respectively.
Recommendation 4 has been placed in the RTS as TR-213-ADM.
Recommendations S and 6 are still being
]
processed by the BWOG (see SER Section 12.1).
The NRC staff notes that these i
j recommendations will reduce unneeded reactor trips and thus are acceptable.
Recommendations 7 and 8 have been rejected by the BWOG Steering Committee.
Recommendation 7 was rejected because it is standard industry practice to re-d train operators after modifications are made, and recommendation 8 was rejected i
because ICS tuning is covered under recommendation TR-107-ICS.
The staff notes that the specific training mentioned in recommendation 7 is involved with the i
implementation of the revised arming threshold for the ARTS on turbine trip.
Thus, the standard practice to retrain operators after plant modifications will cover this recommendation.
The staff also believes that recommendation 8 is NUREG-1231, Supp. 1 10-3 l
adequately covered by the existing recommendation to improve tuning of the ICS.
Therefore, the staff agrees with the rejection of these recommendations.
Refer i
to Section 6.1 of this SSER for further evaluation regarding the electrical aspects of many of these recommendations.
10.3.2 BWOG Recommendations Related to Reactor Trips Initiated by the Main Feedwater System As a result of its review, the BWOG identified the 11 recommendations listed r
below to reduce the frequency of trips initiated by the MFW system.
(1) Each plant should review, as a minimum, the MFW system design to identify and eliminate single failures that can cause the loss of multiple /
redundant equipment necessary to support MFW pump operation.
(2) Each plant should review the feedwater startup and operations procedures to determine if switchover of steam supplies can be made at lower power levels and if the second MFW pump should be running when switchover is made.
(3) A time delay relay or an orifice between the MFW pump oil system and the ARTS sensing lines should be installed to prevent transient pressure oscil-lations from actuating the RPS circuitry.
(4) The BWOG should identify additional recommendations to improve the capa-bility for successful plant runback on loss of one MFW pump.
(5) RTS item TR-067-MFW should be expanded to include the condensate and booster pumps, and the wording of TR-067-MFW should be modified to ensure that those trip signals that are required for pump / plant protection are hardened.
(6) The post-maintenance program recommended in TR-068-MFW should include proper checkout and tuning of newly installed systems or equipment before I
unit startup and planned tuning at power should be incorporated into plant procedures.
1 (7) The Transient Assessment Committee should develop a conceptual design for a multiparameter control room display that would provide the operator with a primary versus secondary heat balance.
(8) The Crystal River / Davis-Besse problems associated with deaerator tank level control and the Crystal River /Oconee problems associated with feed-water recirculation during startup need to be reviewed and ne'.essary design changes implemented to prevent their recurrence.
(9) The BWOG I&C Committee should evaluate implementing a design change to automatically reset the ARTS bistables on MFW pump reset.
(10) Each plant should provide the status of the ARTS bistables on each MFW pump trip in the control room.
(11) The plants that trip both MFW ptsnps on high steam generator level should evaluate sequential versus simultaneous MFW pump trip.
NUREG-1231, Supp. I 10-4
=._ -
1 Recommendation 11 above has been rejected by the BWOG Steering Comittee l
because it determined that simultaneous MFW pump trip was the proper response to the steam generator high-level signal because of the severity of the conse-quences of steam generator overfill and potential overcooling concerns.
The staff agrees with the disposition of this recomendation.
Recommendations 1 and 2 were evaluated in Section 6.2.2 of the original SER and are both in the RTS.
Recommendation 3 has already been entered into the RTS as TR-170-MFW, which was incorporated from the review of Transient Assessment Program (TAP)
Repert No. ANO-85-06.
The remaining recommendations are still being processed j
by the Steering Committee and therefore are not in the RTS.
Many of the remain-ing recomendations are closely related to existing recommendations in the i
tracking system so that further coordination is necessary for their disposition.
The staff concurs with these recommendations and encourages the BWOG to approve 4
and enter them into the RTS to further aid in reducing the frequency of reactor j
trips initiated by the MFW system.
Refer also to Section 6.1 of this supple-ment for a further evaluation concerning the electrical aspects of these recommendations.
J 10.3.3 BWOG Recommendations Related to Reactor Trips Initiated by the ICS The BWOG identified five recommendations to reduce reactor trips initiated by the ICS.
(1) The lessons learned from the Rancho Seco experience regarding high-resistance relay contacts should be included as a part of the recommended l
ICS/NNI preventive maintenance program in TR-038-ICS.
i t
(2) A periodic preventive maintenance program should be established to in-crease the reliability of inverters and vital buses.
(3) Existing RTS recommendation TR-038-ICS should be modified to include internal and external power supplies to the ICS/NNI.
(4) RTS recommendation TR-104-ICS should be expanded to include all ICS input signals.
(5) The BWOG I&C Committee should develop a post-trip or post-transient troubleshooting procedure to aid in diagnosing ICS module failures.
i These recommendations are included in the NRC staff's evaluation of the BWOG ICS/NNI system review documented in Section 6.1 of this supplement.
1 J
10.3.4 BWOG Recommendations Related to Reactor Trips Initiated by the Control Rod Drive (CRD) System Five recommendations related to reducing reactor trips initiated by the CR0 system are listed below.
l
)
}
(1) Each B&W plant should evaluate eliminating or reducing the automatic ICS runback rate on assymetric rod condition from 30 percent to 3 percent full power per minute or some other reduced rate that is compatible with plant i
Technical Specifications.
l NUREG-1231, Supp. 1 10-5 I
i r
i (2) Plant operating procedures should provide instructions for de-energizing CRD controllers to stop uncommanded rod group insertions.
The method used at Arkansas Nuclear One, Unit 1 (ANO-1), should be reviewed for applica-bility at each B&W plant, i
(3) Each B&W plant should review its CR0 cabinets for proper labeling and i
j consistency with procedures.
(4) Maintenance procedures should be reviewed and upgraded as necessary to I
4 ensure that replacement modules and assemblies in the CRD system are i
properly checked out before use.
This recommendation should be extended t
to include ICS and EHC components, circuit boards, etc.
s (5) Plant procedures should include a requirement to trend the power / imbalance versus time relationship during xenon oscillations.
The first two recommendations have been entered into the RTS as TR-204-ICS and TR-214-CRD, respectively.
The first recommendation is evaluated in Section 6.1.4 of this supplement as part of the staff's overall evaluation of the BWOG ICS/
NNI review.
The third and fourth recommendations were rejected by the BWOG Steering Committee.
The third recommendation was rejected because this was i
deemed to be an industry issue being addressed on a broader plant-wide basis.
The staff believes that the ongoing control room design reviews being performed by industry are sufficient to address this concern.
Thus, the staff agrees with this rejection.
The fourth recommendation was developed as the result of several reactor trips that occurred after failed components were replaced with components that also were defective.
However, the BWOG Steering Committee re-jected the recommendation to check out replacement components before installa-2 tion because it is not a feasible or practical solution to the problem.
The i
staff believes that the recommendation is directly responsive to the operating experience and that the Steering Committee should re-evaluate its rejection.
The last recommendation is still being processed by BWOG.
I a
The NRC staff believes that the recommendations developed by the BWOG are respon-j sive to the operating experience at B&W plants and that implementation of the recommendations should reduce the reactor trips initiated by the CRD system, s
Thus, the staff finds the recommendations acceptable.
However, the BWOG should i
reconsider its rejection of the recommendation to check out replacement com-ponents before their installation.
10.3.5 BWOG Recommendations Related to Reducing High-and Low-Pressure RCS f
Trips by Improving Transient Response Capability The seven recommendations listed below were developed by the BWOG to reduce the 1
frequency of reactor trips by improving the RCS thermal-hydraulic response to initiating events.
(1) Toledo Edison should evaluate lowering the low pressure RCS trip set point from 1985 psig to 1900 psig.
J (2) The BWOG Analysis Committee should re-evaluate the variable low pressure i
RCS trip set point.
}
NUREG-1231, Supp. 1 10-6 i
i a
i (3) The BWOG Analysis Committee should evaluate lowering the initial, steady-
~c state RCS pressure set point from its present value of 2155 psig to pro-vide additional margin to the high pressure RCS trip set point.
(4) The BWOG should investigate methods for increasing the capacity of pres-
[
surizer spray flow to improve plant response to high pressure transients.
(5) The BWOG should investigate methods for lowering the temperature of pressurizer spray flow to improve the plant response to high pressure i
(6) The pressurizer spray control circuitry should be modified to automatic-l ally open the valve to the fully open position on initiation of high-pressure transients, or a fast-acting spray valve should be installed.
(7) An anticipatory spray signal should be developed to open the pressurizer i
spray valve for predefined events that could lead to a high pressure transient.
Only the first recommendation has been entered into the RTS as TR-205-RPS.
The remaining recommendations are still being processed by the BWOG.
The first two recommendations involve change in set points for the RPS.
The first recommends an evaluation of the low pressure RCS trip set point for Davis-Besse.
The Davis-Besse low pressure RCS trip set point is 1985 psig while the other B&W plants have trip set points of either 1800 or 1900 psig.
Thus, the BWOG concluded that this change could probably be implemented at Davis-Besse without a significant decrease in safety margins.
The objective of the second recommendation is to eliminate the variable low pressure RCS trip in the RPS.
The BWOG notes that only three reactor trips have occurred since 1980 i
from this trip function and that other trip functions in the RPS would have i
actuated and safely mitigated these events.
The BWOG report contains insuffi-cient information for the staff to conclude that these changes are appropriate.
However, these modifications would require changes to the plant Technical Speci-t fications and would have to be approved by the staff.
Thus, pursuit of the i
studies recommended by the BWOG is acceptable to the staff since the utilities
+
must perform and submit safety evaluations for staff review and approval before implementing RPS set point changes, b
The third recommendation is to evaluate lowering of the steady-state operating pressure on the B&W plants in order to provide additional margin to the high-pressure RCS trip set point.
Implementation of such a change in the normal plant operating conditions would affect the plant safety analyses and would probably require RPS set point changes and changes to the plant Technical Speci-fications.
Should the BWOG decide to pursue this modification, the staff will require the BWOG to submit a plant safety evaluation for the change for stafd review and approval before its implementation.
The remaining recommendations are focused on improving the effectiveness of the pressurizer spray system and thereby limiting the magnitude of overpressure transients.
Such action will provide enhanced capability for B&W plants to respond to some plant upsets, such as loss of a single MFW pump, without result-ing in a reactor trip.
The staff believes that modifications to the pressurizer j
(
NUREG-1231, Supp. 1 10-7 i
= _..
i 4
i spray system will prevent unnecessary reactor trips for some events.
- However, the BWOG must ensure that any changes that may be implemented will not affect the plant safety analyses.
6 The reco.nmendations made by the BWOG in this category will require careful study to ensure plant safety margins are not significantly reduced.
- However, the staff believes that, with proper analyses and engineerin0, some of these l
changes can probably be implemented in such a way as to reduce unnecessary j
reactor trips while not compromising plant safety.
10.4 Summary of Staf f Findings i
The NRC staff concludes that the implementation of BWOG recommendations will reduce the number of reactor trips.
However, the staff believes that the BWOG shoold i
reconsider its rejection of the recommendations to (1) place protective covers over local level / trip switches that can directly result in a turbine or reactor trip and (2) ensure that replacement components are properly checked out before their use.
In addition, the utilities should submit plant safety evaluations for staff review and approval before implementating changes to the RPS or the steady-state operating pressure of B&W plants.
l 4
1
)
i j
i l
t i
l l
i l
l I
l
)
i 1
NUREG-1231, Supp. I 10-8
l l
11 BWOG PROGRAMMATIC AND MANAGEMENT ACTIONS 11.4 Safety and Performance Recommendation Integration Group As stated in the SER, the NRC staff had not completed its review of the key re-comendations and therefore would not comment on the acceptability of those key recomendations until the staff's review of ICS/NNI, reactor trip initiating events, and the emergency feedwater initiation and control system was completed.
J After the staff completed its review of these issues, it proceeded to review the recomendations contained in the BWOG RTS report dated September 1987 to determine the adequacy of the key recomendations identified by the BWOG. As mentioned in the SER, the Safety and Performance Recomendation Integration Group (SPRIG) developed specific criteria to determine the key recommendations.
The staff has determined that the criteria used by the BWOG are acceptable and therefore the key recommendations identified by the BWOG are generally appropriate.
In addition, the staff reviewed all the recommendations in the RTS to determine j
whether any of those recommendations not designated as non-key recomendations have sufficient safety significance to be considered high priority recomenda-tions and thus should be given additional attention during the staff's review of the utility's implementation plans.
Following is the staff's list of addi-4 tional RTS recommendations it considers as high priority and its bases:
(1) TR-115-PES - Test diesel generators to ensure they will carry loads under expected sequential loading conditions.
If the diesel generators have not been tested to ensure they can carry loads, then a potential safety concern exists since diesel generators provide the required emergency L
power in the event of a loss of offsite power.
l (2) TR-122-IAS - Systematically inspect the instrument air system for leaks.
This can be done easily and may significantly improve the performance of the instrument air system since air leaks can lead to contamination and overextension of air system capacity.
1 (3) TR-128-IAS - Review training and loss of air response procedures for in-1 strument air systems.
A potentially large benefit in ensuring safe plant j
shutdown can be gained from a relatively small effort by improving the capability to recover from a loss-of-instrument-air event.
(4) TR-143-IAS - Inspect accumulators and their check valves in the instru-ment air system.
If those accumulators supplying backup air for safety-l
.i related equipment have not been inspected / tested on a regular basis, then a potential safety concern exists since accumulators are relied on to ensure valve operability follcwing a loss of normal instrument air.
(5) TR-144-IAS - Develop or upgrade loss-of-instrument-air procedures (appli-cable to only three BWOG plants).
The concern is the same as indicated under item 3 above.
j i
)
NUREG-1231, Supp. 1 11-1 r
l
(6) TR-157-0PS - Validate emergency operating procedures (EOPs) to determine if adequate staffing and prioritization exists.
The BWOG review of Cate-gory B and C events shows increased operator attention is required to establish plant control for these events.
Thus, validation of the E0Ps will provide assurance that the operator guidance is available for miti-gating these events.
(7) TR-174-MSS - Improve response of modulating turbine bypass valves.
As dis-cussed in Section 6.4 of the SER, main steam pressure control has not been ideal at B&W plants and has led to increased operator burden.
Improved response of the modulating turbine bypass valves will reduce the complexity of and increase the safety margins of post-trip steam pressure control.
(8) TR-179-MFW - Identify areas for enhancing the reliability of the MFW and condensate systems an( controls.
The BWOG studies indicate that the frequency of MFW system-initiated reactor trips has not declined despite many MFW system modifications.
Therefore, improving MFW/ condensate system performance warrants priority consideration.
(9) TR-181-0PS - Verify adequacy of instrumentation and displays used to assess and control the stability parameters in the Abnormal Transient Operator Guidelines.
Adequate instrumentation and display are required to ensure proper response by control roorr operators.
(10) TR-190-ICS - Develop backup manual and automatic controls for pressurizer level and pressure control powered from another power source.
The staff evaluation of the capability of the B&W plants to reach a known safe state (KSS), following a partial or full loss of NNI/ICS power, found that in certain conditions the operator burden in controlling the plant is excessive and beyond that allowed by the BWOG definition for KSS.
When NNI-X power is lost, the operator is required to manually control pressurizer level and pressurizer pressure when all indications in the control room of these two plant variables and all operable controls in the control room for these variables may have been lost.
(11) TR-201-MTS - Review EHC overspeed and fast control and intercept valve circuits.
Since these circuits have led to seven reactor trips, improv-
)
ing their reliability could lead to a significant decrease in the number of turbine-related trips.
As noted in Section 12.2 of the SER, the NRC staff will review each utility's implementation plans.
The staff will ensure that the key recomendations iden-tified in the SPRIG report and the high priority recommendations noted above are implemented in a timely manner.
Furthermore, the staff expects each util-ity to evaluate oil recommendations, not just the key recommendations.
The staff will perform plant-specific audits and will monitor the utility's imple-mentation programs to ensure proper disposition and timely implementation of all the recommendations in the RTS.
4
)
11.5 BWOG Evaluation Program To Assess Quality of Implementation of the SPIP Recommendations In a letter to D. Crutchfield dated Novembe-r 24, 1987, the BWOG stated that a i
program has been initiated to evaluate how each member utility dispositions the NUREG-1231, Supp. I 11-2 i
recommendations produced by SPIP.
To ensure the quality of implementation, the BWOG has taken the following three phase approach:
(1) to evaluate a member utility's program for managing the SPIP recommendations, (2) to assess the tech-nical adequacy of the implementation details for selected recommendations, and (3) to monitor the attainment of the SPIP goals.
The BWOG Executive Committee has formed an evaluation team and has established mid-February as the completion date for reviewing Phase 1 of the program for each of the seven B&W utilities.
The staff believes there is a need for a program such as that developed by the BWOG.
The majority of the recommendations developed by the BWOG do not pro-vide specific design details.
Rather, the recommendations generally define de-sign criteria with the engineering of the detailed plant-specific modifications left to the utility.
This approach is appropriate in light of the specific differences that exist among B&W plants.
The BWOG's plans to audit the imple-mentation of the recommendawiens to ensure that the interpretation of the rec-ommendations and the quality of the implementation are commensurate with the intent of the BWOG SPIP.
Therefore, the staff concludes that the BWOG program to audit the implementation of the recomendations should contribute toward achieving the goals of the SPIP.
One of the concerns identified by an NRC staff member (SER Section 2.2, item
- 4) relates to the fact that neither the BWOG nor the utilities have effec-tively analyzed the proposed SPIP recommendations to determine the effects on 2
other parts of the plant.
During a meeting with the BWOG on the reassessment i
program, the Advisory Committee un Reactor Safeguards (ACRS) asked a similar question related to whether the SPIP involved any systems interactions l
studies.
The BWOG response, provided in BAW-1919,Section XI, p. XI-77, was that systems interactions were indirectly addressed through a detailed systems review of several major systems and their dependence upon and relationship to one another.
The BWOG also stated that the sensitivity study, risk assess-ment, and failure modes and effects analysis in the ICS/NNI system evaluation, in combination with the detailed systems reviews, involved systems interaction activities.
l The staff agrees with the BWOG that the reviews. performed involve some systems interaction studies.
However, because many of the SPIP recommendations are general in nature, detailed systems interaction studies cannot be performed at this time.
As detailed engineering is performed by the individual utilities to implement the recommendations, systems interactions can be better defined and more effectively analyzed.
The staff notes that one of the guidelines for the audit program established by the BWOG is to evaluate, on a plant-specific basis, the potential for introduc-ing new failure modes or systems interactions.
It is the staff's opinion that because of plant-specific differences in B&W-designed plants, this is an accept-able approach and if properly evaluated should provide more confidence that any potentially adverse systems interactions will be addressed.
In addition, the staff stated in Section 12.2 of the SER that it intends to audit the plant-specific implementation of the approved BWOG recommendations at selected B&W plants.
As part of this audit, the staff will assess, where appropriate, if systems interactions have been considered.
The staff believes that the combination of actions described above provides the assurance that significant systems interactions related to the recommendations approved by the BWOG will be effectively addressed.
NUREG-1231, Supp. 1 11-3
12 IMPLEMENTATION 12.2 NRC Audit of the BWOG Recommendation Approval Process In this section of the SER, the staff stated that the NRC intends to send an inspection team to selected B&W plants early in 1988 to audit plant-specific implementation of approved BWOG recommendations.
In a letter dated December 21, 1987, the BWOG stated that most of the recommendations require further evaluations that may require followup actions once the evaluations are completed.
Since most of the recommendations are not at a stage where the acceptability of the utilities' interpretation of the implementation of the recommendations can be evaluated by the staff, these audits will be deferred.
New dates will be estab-lished at a later time based on the status of the recommendations at the utili-ties selected to be audited.
Before the implementation audit, the staff plans to audit the program in place at selected utilities to evaluate the SPIP rec-ommendations for implementation at the plants.
This audit will be similar to that performed by the BWOG as describec in its letter of November 24, 1987, and discussed in Section 11.5 of this supplement.
This audit will provide the staff with additional assurance that the BWOG is properly monitoring the plans of its individual members.
12.3 Matrix of Approved Recommendations In the SER, the staff stated that Table 12.1, which identified each of the 207 approved recommendations, and Table 12.2, which identified the 22 recommen-dations rejected by the BWOG, will be updated in supplements to the SER as the staff completes its review of the remaining sections of the BWOG SPIP and the BWOG submits updated information in the recommendation tracking system (RTS) report.
Tables 12.1 and 12.2 have been updated as a result of the staff's completion of the review of integrated control system /non-nuclear instrumenta-tion and reactor trip initiating events reports, the addition of eight recommen-dations in the January 1988 update of the RTS report, and the reconsideration j
of two previously rejected recommendations.
12.4 Implementation of Recommendations In a letter dated December 21, 1987, the BWOG pre.vided comments on the staff's SER of the BWOG plant reassessment prograu.
The BWOG stated that it may not be possible for each utility to provide an implementation schedule by June 1, 1988, for most of the recommendations because further evaluation and potential followup actions may be required by the utilities.
In addition, it stated that 4
formal orders are not needed since the BWOG is fully committed to timely reso-4 lution of all appropriate i acommendations.
The BWOG also committed to provide periodic updates of the RTS, which show the status and progress and schedular information requested by the staff in the SER.
The staff believes that this RTS submittal will keep the staff cognizant of the overall implementation of the recommendations.
However, because the BWOG i
1 1
NUREG-1231, Supp. 1 12-1 q
.~
v-
l l
lacks the authority to enforce individual utility commitments to the NRC, the staff still believes that its request in the SER for each utility to provide its schtdule for implementation of the recommendations is appropriate.
Based on the BWOG's comment that it may not be possible to provide an implementation schedule for most of the recommendations by June 1, 1988, the staff recommends that the individual utilities provide, no later than that date, a schedule for implementation of those recomendations that have been evaluated, a tentative schedule for implementation of those recommendations requiring further evalua-tion, and justification for the delays in implementation.
The staff will re-view these submittals to determine whether the implementation schedules are ap-propriate.
The staff further recommends that these schedules be regularly updated to include the remaining recommendations not yet approved by the BWOG Steering Committee and that early notification of schedular slippages be made with adequate justification.
The NRC Project Manager for each B&W-designed utility would be responsible for tracking these implementation schedules.
The BWOG RTS updates will be reviewed regularly by the staff to determine the over-all program progress and the progress made by each individual utility.
Since the issuance of the SER, by letters dated January 5 and March 2, 1988, the BWOG has submitted two updated versions of the RTS report (November 1987 l
and January 1988).
The January 1988 report added eight new recommendations to the RTS.
These are the only additions since the September 1987 RTS report up-date was issued.
In addition, two of the previously rejected recommendations (T-34 and TIR-2) have been re-evaluated by the BWOG and added to the list of approved recommendations (TR-211-ICS and TR-213-ADM are two of the eight added in the January 1988 RTS report).
These reports contain an update on the status of those recommendations already in the RTS.
Summarized below is a comparison of the changes in the status of the recommendations as reported in Appendix J to BAW-1919, Revision 5, issued July 1987, and the September and November 1987 and January 1988 updates to the RTS.
All recommendations (%)
Key recommendations (%)
BAW-1919 RTS updates BAW-1919 RTS updates Recommendation July Sept Nov Jan July Sept Nov Jan status category 1987 1987 1987 1988 1987 1987 1987 1988 1
Evaluating 38 34 44 44 33 36 37 41 Implementing 17 18 19 12 22 31 31 22 Closed operable 16 15 17 20 22 22 23 28 Closed not applicable 15 17 18 18 7
8 8
8 Closed rejected 1
1 1
1 2
1 1
1 No report 13 15 2
4 14 2
While this table indicates some improvement in implementation of the recommen-dations, a more detailed review of the reports shows that from July 1987 to i
1 l
l NUREG-1231, Supp. I 12-2 l
y
January 1988, with the exception of Rancho Seco, little progress was made in implementing and closing recommendations currently in the RTS.
It appears that very little progress has been made recently toward approval of the remaining recommendations by the BWOG Steering Committee.
Although this may not be repre-sentative of actual progress made, the staff is concerned about the apparent lack of substantive progress.
Since it is the BWOG's intent to use the RTS to keep the staff informed of progress, the staff recommends that either more de-tail be provided in the RTS report or supplemental information be provided as an enclosure to the updated RTS reports so that the staff may more accurately evaluate program progress.
The staff will meet with the BWOG and the individual utilities to discuss the details of the schedular commitments for implementation of the recommendations and to identify the additional information needed to augment the updated RTS reports.
Should the staff find at a later date that this approach does not provide the desired results (i.e., if there were significant schedular slip-pages in the implementation of the recommendations), the use of regulatory action on a plant-specific basis to ensure that the recommendations are imple-mented in a timely manner will be considered.
12.5 Impact Assessment of Proposed BWOG Recommendations As discussed in Section 12.5 of the SER, the staff had not completed its review 1
of the BWOG SPIP evaluation of the ICS/NNI and thus was unable to complete its 4
j assessment of the SPIP recommendations on the safety of B&W plants. Within most of the sections of the SER and this supplement, the staff has evaluated the individual BWOG tasks and the recommendations developed from them.
In this section, specific safety issues have been defined and those RTS recommendations appropriate to that issue are collected.
This collection of recommendations is then examined to provide another perspective of the SPIP efforts for B&W plant safety.
The conclusions contained herein do not diminish any specific staff F
comments made in the individual sections of the SER or this supplement.
12.5.1 Reduction in Reactor Trips One of the goals of the SPIP is to develop recommendations that will decrease the frequency of reactor trips.
Areas of improvement and the associated recom-mendations in the RTS include those listed below.
(1) reduce the number of reactor trips to ICS/NNI that are caused by input signal failures (TR-001-ICS, TR-003-ICS, TR-005-ICS, and TR-104-ICS)
(2) improve power supplies to the ICS/NNI (TR-013-ICS, and TR-102-ICS)
(3) improve ICS performance by improved maintenance and tuning (TR-008-ICS, TR-009-ICS, TR-038-ICS, TR-107-ICS, TR-186-ICS, and TR-204-ICS) 1 (4) improve reliability of main feedwater system by minimizing trips to the MFW pump (TR-015-MFW, TR-016-MFW, TR-052-SFI, TR-053-SFI, TR-066-MFW, TR-067-MFW, TR-073-MFW, TR-076-MFW, TR-082-MFW, TR-083-MFW, TR-084-MFW, TR-170-MFW, TR-171-0PS, and TR-179-MFW) l 4
NUREG-1231, Supp. 1 12-3
(5) improve MFW control systems (TR-017-MFW, TR-021-MFW, TR-068-MFW, TR-069-MFW, i
TR-070-MFW, TR-072-MFW, TR-075-MFW, TR-085-MFW, and TR-088-MFW)
P (6) improve maintenance and operator training for the MFW system (TR-018-MFW, TR-020-MFW, TR-068-MFW, TR-074-MFW, TR-079-MFW, TR-173-MFW, and TR-202-HFW)
~
(7) reduce the number of reactor trips caused by the turbine systems (TR-025-MTS, TR-030-MTS, TR-100-MTS, TR-101-MTS, TR-168-MTS, TR-169-MTS, TR-200-MTS, TR-201-MTS, and TR-206-MTS)
(8) modify reactor trip set points to provide increased capability to 5
accommodate RCS response to upsets without resulting in a reactor trip (TR-027-ADM, TR-031-RPS, and TR-205-RPS)
(9) improve reliability of plant electrical systems (TR-112-PES, TR-113-PES, TR-117-PES, TR-119-PES, and TR-203-PES)
The staff expects that the actions proposed by the BWOG will, when implemented, have a significant impact on the frequency of reactor trips at B&W plants.
12.5.2 Challenges to Safety Systems Challenges to safety systems have been considered from several perspectives by the BWOG.
As noted above, the BWOG has proposed several actions that will reduce the challenges to the reactor trip system.
Additionally, several of the actions discussed above will result in enhanced reliability of the MFW system and thereby will decrease the need for actuation of the auxiliary /
emergency feedwater (AFW/EFW) system.
Other actions recommended to reduce unnecessary AFW/EFW actuations include those listed below.
(1) review and modify AFW/EFW actuation set points to prevent unnecessary challenges to the system (TR-022-EFW and TR-161-EFW)
(2) improve operator control of the MFW system following a reactor trip (TR-069-MFW, TR-070-MFW, TR-071-MFW, and TR-207-0PS)
To minimize actuations of the emergency core cooling systems (ECCS), those actions listed below have been propnsed to minimize the potential for overcool-ing of the RCS.
i (1) improve performance of the main steam pressure control systems i
(TR-023-MSS, TR-024-MSS, TR-048-MSS, TR-109-MSS, TR-174-MSS, and TR-176-POV) l l
(2) reduce overcooling of the RCS as a result of excessive MFW or AFW/EFW
{
l flow (TR-098-MFW, TR-099-0PS, TR-155-EFW, TP-159-0PS, and TR-176-POV)
(3) minimize RCS response to loss of ICS/NNI power events or restoration of I
ICS/NNI power following such events (TR-032-ICS, TR-036-ICS, TR-037-ICS, TR-096-MSS, TR-097-EFW, and TR-178-ICS)
NUREG-1231, Supp. 1 12-4 i
(4) reduce probability of PORV opening and ensure ability to isolate a stuck PORV (TR-051-0PS, TR-172-PRV, and TR-175-PRV)
The staff expects that the combinations of actions proposed by the BWM will
[
reduce the number of challenges to the plant safety systems and thus represent an improvement in the safety of B&W plants.
12.5.3 Operator Burden As noted in Section 5 of the SER, the NRC staff concluded that for reactor trips, B&W plants do not pose greater burden on the plant operator if steam and feed flows to the steam generator are controlled within reasonable bounds.
As dis-cussed above, SPIP proposed several recommendations to ensure that steam and feed flows would be properly controlled following a reactor trip.
Additionally, implementation of the acticns proposed, most notably TR-024-MSS and TR-174-MSS, should reduce the need for B&W plant operators to take manual control to ensure that the main steam safety valves fully reseat following a reactor trip.
- Thus, these recommendations should reduce operator burden.
One of the areas of increased operator burden at B&W plants, noted in Section 5 of the SER, is the operator's difficulty in diagnosing and responding to failures of automatic control systems (ICS/NNI).
Several of the recommendations discussed above were developed by the BWOG to address this issue.
Additional recommenda-tions related to the adequacy of instrumentation available and the avai' ability of emergency operating procedures for a loss of ICS/NNI power include TR-012-ICS, TR-034-ADM, and TR-154-ICS.
Beyond these specific areas, the BWOG made additional recommendations to improve operator performance and reduce operator burden including those listed below.
(1) improve emergency notifications (TR-058-0PS, TR-059-0PS, and TR-156-0PS)
(2) identify high priority tasks during emergencies for specific training (TR-061-0PS, TR-064-0PS, TR-157-0PS, and TR-177-0PS)
(3) improve information available to operator in control rcom (TR-062-0PS, TR-063-0PS, and TR-158-0PS)
While the staff noted in Section 7 of the SER that the BWOG evaluation of oper-ator burden would have been enhanced by the use of human factors expertise, the staff believes that the implementation of the SPIP recommendations will have a positive effect on reducing the operator burden at B&W plants.
12.5.4 Decay Heat Removal B&W plants have several means of ensuring decay heat removal.
First of these is the use of the MFW system.
Actions taken by the BWOG to improve the reli-ability of the MFW system have been discussed above.
In the event that all main feedwater is lost, the AFW/EFW system is used to en-sure the steam generator heat sink.
Several recommendations developed for this system have been discussed previously, Additional recommendations developed to improve the reliability of the AFW/EFW system include those listed below.
NUREG-1231, Supp. 1 12-5
(1) improve performance of motor-operated valves (TR-041-M0V, TR 042-MOV, TR-043-MOV, TR-044-MOV, TR-045-MOV, TR-046-MOV, and TR-047-MOV)
(2) improve operational reliability of EFW system (TR-160 EFW and TR-166-EFW)
(3) improve maintenance, testing, and surveillance of EFW system (TR-163-EFW, TR-164-EFW, and TR-165-EFW)
The staff also notes that it has an ongoing activity under GI-124 to evaluate the reliability of certain plants having only two AFW/EFW pumps, including Rancho Seco, Crystal River Unit 3, and ANO-1.
The staff will develop evalua-tions/ recommendations for the AFW/EFW systems in these plants, separate from the BWOG SPIP, to ensure that they demonstrate acceptable reliability for a loss of MFW event.
In the event that all sources of feedwater are unavailable, B&W plants have an alternate way of removing decay heat by feed-and-bleed cooling, as discussed in Section 5.3.8 of the SER.
No specific recommendations have been developed by SPIP to enhance feed-and-bleed systems at B&W plants.
However, the staff noted that the Davis-Besse plant, which unlike the other B&W plants has a low shutoff head for the high pressure injection system, is pursuing separate plant-specific modifications to enhance its feed-and-bleed capability.
To provide additional assurance that the feed-and-bleed system would be manually actuated by the plant operators when required, SPIP recommended, as TR-177-0PS, that operators be given specific training to reinforce the need to take action when specified by the emergency operating procedures (EOPS).
The NRC staff finds that implementation of the recommendations developed under SPIP will provide added assurance that adequate decay heat removal will be pro-vided at B&W plants.
12.5.5 Pressurized Thermal Shock Pressurized thermal shock (PTS) is a situation wherein the RCS overcools while it either remains pressurized or repressurizes.
The BWOG proposed several actions aimed at reducing the potential for overcooling the RCS, as discussed in Section 12.5.2 above.
Examination of previous analytical results of PTS for B&W plants shows that one of the most frequent, and the most severe, PTS events is the plant response to a loss of ICS/NNI power.
SPIP recommendation, TR-178-ICS, is directly re-sponsive to this transient scenario and should have a significant effect on the PTS risk at B&W plants.
As noted in Section 9 of the SER, PTS risk at B&W plants is not a significant contributor to core damage even before the implementation of the recommendations.
The staff believes that the actions proposed by the BWOG will further reduce any J
PTS risk posed at the B&W plants.
l NUREG-1231, Supp. I 12-6 i
2 C::o rn CD e
H Nw."
Table 12.1 SPIP Recommendations m
Cv V
Recommenda-5[R/55tR tion No.
Subject Source: BAW-1919 and other reports section w
1R-001-ICS Replace RC flow signal input to ICS with RC pimap status.
Vol. 1, pp. IV-14 to IV-16, $A.5.c.1
$$[R 6.1.3 TR-G02-ICS Autoratically detect invalid RC temperature inputs to Superseded by Recommendation TR-104-ICS 55tR 6.1.3 ICS (superseded by Recommendation TR-104-ICS).
TR-003-ICS Remove startup IW flow correction to main FW flow Vol. 1 pp. IV-17 & 18, $A.S.c.3 SSE R 6.1. 3 function from the ICS.
TR-004-ICS Automatically detect an invalid input to ICS of turbine Superseded by Recommendation TR-104-ICS SSER 6.1.3 header pressure (superseded by Recommendation IR-104-ICS).
IR-005-ICS Remove neutron fluu signal auctioneering circuitry f rom Vol. 1, pp. IV-19 & 20, $A5.c.5, &
SSER 6.1.3 g
N RP5 and relocate in the ICS.
Appendin J. p. 3-51 e
TR-006-ICS Delete FW temperature correction to FW demand from 105.
Vol. 1, pp. IV-20 & 21, $A.5.c.6. &
$5t R 6.1. 3 Appendia J, p. 3-51 TR-007-ICS Remove Btu limits from ICS.
Vol. 1, pp. IV-9 & 10, 5A.4.6.1 SSER 6.1.4 TR-008-ICS Improvements to reactor runback capability.
Vol. 1, pp. IV-11 & 12, $A.4.c.2
$$tR 6.1.4 TR-009-ICS Improvements to ICS tune control circuits.
Appendix J. $1.n., p. 24
$$[R 6.1.4 TR-010-ICS Modify ICS contrcl circuit.
Appendix C, p. 2-5, 52.4.2 &
SSER 6.1.4 Appendix R. la., p. 7, 52.1.6 TR-Oll-ICS Determine if the grid frequency error circuit has Vn1. 1, p. IV-13, SA.4.c.5 55tR 6.1.4 been detuned.
TR-012-ICS Determine if operator has necessary information from Appendix J. %IX.d(1.3.3.f)
$$tR 6.1.6, 6.1.9 procedures, indicators, etc. to detect loss of NMI power.
m.m___
_.m_
M i
a z
C i
x rn
+
O eW to Table 12.1 (Continued) f w
C V
Recommenda-5tR/55tR F
tion No.
Subject Source: BAW-1919 and other reports Section TR-013-ICS Prevent loss of power to the ICS or fel.
Appendia J, 51.a.5 & 51X.d 55tR 6.1.3 TR-014-W W Install monitoring system on N W pumps to document Appendia H pp. 3-44 & 3-53 stR 6.2/551R 6.1.12 I
causes of pump trips.
1 T R-015-NW Determine if a low N W pump suction pressure is needed Appendim 8-4, p. 87, %.2 & Appendin H.
SER 6.2
- p. 3-45 T R-Olb-NW Investigate oil system pressure in N W pump.
Appendim 8-4 pp. 90 & 91, 56.3 SIR 6.2/55tR 6.1.12 T R-017-WW Evaluate N W pump control systems.
Accendim 8-4, pp 90 & 93, %.4 & %.4.2 SER 6.2/55tR 6.1.12 w
TR-018-W W Provide training on N W system components.
Appendia B-4, p. 90, 56.2 & Appendin H,
$[R 6.2, 7,3/55tR 6.1.9
'f
- p. 3-35 a>
T R-019-NW Ensure there are sufficient annunciator and trip signals Document No. 47-1158723-00 "8WOG Oper - SER 6.2/55tR 6.1.12 for N W supply system.
atirug Emperience Summary Report for i-1984," dated June 1985, p. 19 i
T R-020-NW Establish procedures for switching of N W pump oil Document No. OC-1-85-02, "8WOG TAP 5[R 6.2 supply.
Report for Oconee 1 Reactor Trip on December 3, 1984." $1, p.1 TR-021-M W Identify causes for N W pump control problems.
Document 86o. 08-1-84-05 (84W) Document
$tt 6.2/55ER 6.1.12 1
No. 12-1151048-00), BWOG TAP Report, i
"D-8 Rm Trip on 1/8/84," $1.03, i
- p. 2 & Sil.E
- p. 7 l
TR-022-N W Review EFIC system low SG 1evel set points.
Document No. D-8-1-85-03, "D-B Events SER 6.3 of 3/16/85 and 3/21/85." $1.D.1, p. 2 I
TR-023-H55 Deterisine need to replace M55V release nut cetter pins.
Document No. 12-1151244-30, *D-B TAP SER 6.4 Report. D-8 Reactor Trip of 3/2/84."
511.E.1, p. 13 l.
l 1
m.
t t
zcx l
rn 1
a l
9W fu lable 12.1 (Continued) m E
Recommenda-SE R/SSE R y
tion No.
Subject source: BAW-1919 and other reports section es IR-0?4-M55 Determine causes to correct anomalous post-trip Document No. OC-1-85-01 "BWOG TAP 5[R 6.4 perf ormance of M55Vs.
Report, Oconee 1 Reactor Trip on December 2, 19R4," %Ill.B.1, p. 13 TR-0?5-MIS Review fHC system for loss of input power.
"EWOG TAP Operating Experience Summary
$[R 8.3/55tR 6.1.12 Report, Jan-Dec 1983." p. 2-4 TR-026-OPS Operability of SG shell themocouples.
Document No. 12-1151244-00 "EWOG 1AP
$[R 4.2 Report, Davis-Besse Reactor Trip on March 7, 1984, " $11. E.7, p. 13 TR-027-ADM Calibration techniques for power range imbalance.
"PWOG 1ransient Assessment Program
$[R 4.2 Operating Experience Summary Report, g
,o Jan-Dec 1983 " p. 2-5 e*
TR-028-ADM Training on power / imbalance control.
[xperience Summary Report, January-December 1983,* pp. 3-1 & 3-2 TR-029-ADM Include human error infomation in TAP reports.
Oncument No. 77-1156349-00, "BWOG 5[R 7.3 Report, A Study of Human Interface Events at B&W Operating Plants,"
february 1985, %5.1, p. 27 T R-030-M15 Raise ART on turbine trip arming point.
Appendin B-1, p. 6-1 & Vol. 1,
$[R 8.1 pp. IV-3 to IV-5, %A.1 TR-031-RPS Increase setpoint for high pressure reactor trip.
Appendix B-3, p. 3-2 & Vol. 1, SER 8.2 pp. IV-7 & 8, %A.3 TR-012-IC51
[ valuate restoration of ICS/NNI power.
Appendix R VII.B.1, p. Vil-4 SSER 6.1.6 TR-033-IC5t Ensure that plant will go to a safe state on loss of Appendia R VII.B.1, pp. Vil-4 & VII-5 55fR 6.1.6 ICS/NNI power.
1 Recommendations marked with a daqqer are commitments to the NRC.
z E8 U
Table 12.1 (Continued) mj Recosamenda-SER/S$tR v
tion No.
Subject sourc e: BAW-1919 and other reports section w
TR-034-IC5t Training for loss of ICS power.
Appendix R VII.B. I. p. VII-5 55tR 6.1.9 TR-035-IC51 Familiarire operators with Rancho Seco evect.
Vol. 1, p. IV-29, %5 SSER 6.1.9 TR-036-IC51 Evaluate turbine bypass valve position on loss of ICS.
Appendix R VII.B.I. p. VII-5 55tR 6.1.6 TR-037-IC51 Evaluate MFW pump speed control on loss of 1C5 power.
Appendin R VII.B.1, p. VII-5 551R 6.1.6 TR-038-ICS Develop and implement a preventive maintenance program Appendix R V?l.B.1, p. VII-5 SSER 6.1.8 for the ICS/NNI.
1R-039-1C5 Wire the power supply monitor in the ICS/NNI directly Appendix R VII.B.1, p. VII-5 55ER 6.1. 7 to the output bus af ter the auctioneering diodes.
4 TR-040-ADM Use the TA Comunittee's Trip Intestigation/ Root Cause Appendix H. p. 4-2 Action Item I SER 11.1 o
Determination Program TR-041-MOV Confirm by field inspection data required to site Appendix H
- p. 4-2. Action Item 3.1 SER 8.4 operators and valves for motor-operated valves.
TR-042-MDV Obtain analytic methods used by valve and operator Appendix H
- p. 4-2, Action item 3.2 SIR 8.4 vendors.
TR-043-sev Ensure that torque switch bypass limit switch is set to Appendix H, p. 4-2 Action Item 3.3 SER 8.4 open after valve is unseated.
TR-044-MOV For wedge seating valves, position open direction torque Appendin H
- p. 4-2. Action item 3.4 StR 8.4 switches to the highest allowable setpoints.
TR-045-MOV
[nsure that maintenance procedures provide for properly Appendix H. p. 4-2 Action item 3.5
${R 8.4 setting torque switches and bypass limit switches.
1Recosamendations mar 6ed with a dagger are commitments to the feRC.
z C
- u rn a
e HU lable 12.1 (Continued) w tny Recomunenda-SIR /55IR t.3 tion No.
Subject Sourc e: BAW-1919 and other reports Section 9
T R-046-MOV Challenge valves to open and close under dif ferential Appemiin H. p. 4-2, Action Item 3.6
$[R 8.4 pressures that simulate worst operational arvi auident conditions.
TR-047-NOV Institute formal traitting programs on motor-operated Appendia H, p. 4-2, Action Item 3.7
$[R 8.4 valves.
TR-048-M55 Revise turbine bypass valve preventive maintenance Appendin H
- p. 4-3, Action Item 6.1 SER 6.4 program.
T R-049-M55 Review a w1 revise steam trap preventive maintenance Appendia H. p. 4-3, Action Item 6.2 SIR 6.4 program.
U 1R-050-M55 Include in plant operating procedures provisions for Appendix H. p. 4-3. Action Item 6.3 SIR 6.4 opening steam trap bypass valves <;uring startup, and 8
y draining turbine bypass header valves before startup or cooldown.
TR-051-0P5 Conduct postmaintenance and surveillance PORV testing.
Appendia H. p. 4-4. Action Item 7.1 SER 8.4 TR-052-5FI filter steam generator level signals in steam feedwater Appendin H
- p. 4-4, Action Item 10.1 55[R 6.1.12 rupture control system.
TR-053-STI Correct overheating problems that can lead to electronic Appendia H., p. 4-4, Action Item 10.2 551R 6.1.7 power supply malfunctions.
TR-054-5FI Redesign M51V pneumatic hardware to ensure this equipment Appendix H, p. 4-4, Action item 10.3 SIR 8.4 is cuercised during survelliance testing.
T R-055-ADM Coordinate the activities of plant operations, security, Appendix H
- p. 4-4. Action Itee 11.1 SI R 7. 4 and radcon personnel to f acilitate time;y auess to critical equipment.
TR-056-ADM Move chain link fences to provide better access to Appendix H
- p. 4-5, Action Item 11.2
$I R 7. 4 critical cneponents.
2 Cm en cn e
WU fable 12.1 (Continued) w
<n C
Recommenda-SIR /55tR E.
tion No.
Subject Sourc e: BAW-1919 and other reports Section T R-057-ADM Consider ways to improve access to critical rnerenents.
Appentix H. p. 4-5, Action Item 11.3 SI R 7. 4 TR-058-0P5 Use highest emergency cla.sification Icvel when making Appendia H, p. 4-5, Action Item 12.1
$I R 7. 4 initial notification to NRC.
TR-059-0P5 Training for personnel who make emergew y notificatinns.
Appendix H, p. 4-5, Action Item 12.2
$[R 7.4/55 R 6.1.9 TR-060-0P5 Stress in personnel training that drastic actions Wall Appendin H. p. 4-5, Action Item 12.3 SIR 7.4 be taken if regoired by procedures (superseded by Recom-mendation IR-177-OPS).
TR-061-0P5 For specific training, identify high priority operator Appendix H, p. 4-5, Action item 12.4 SE R 7.4 tasks during emergencies.
ea e
TR-062-0P5 Maintain a high $PDS availability by corrective anrt Appendix H, p. 4-5, Action Item 12.5
${R 7.4/551R 6.1.6 U
preventive maintenance.
TR-063-0P5
[nsure that P/T graphs are provided in the control room.
Appendix H, p. 4-5, Action Item 12.6
${ R 74 TR-064-0P5 Training for resetting turbine driven IFW purts.
Appendix H, p. 4-5 Action Item 12.7 SIR 6.3, 7.4 TR-065-3P5 Improve communications between control room and certain Appendin H
- p. 4-5. Action Item 12.8 SIR 7.4 plant areas at Rancho Seco.
T R-066-M W Ensure that a single electrical failure will not cause a Appendix H. p. 4-6, Action item 14.1
$(R 6.2/55ER 6.1.12 loss of both feedwater trains.
T R-067-M W Wherever possible, eliminate automatic M W pump trip Appendix H, p. 4-6, Action item 14.2 SER 6.2/S$[R 6.1.9 functions.
TR-068-M W Develop a postmaintenance testing program for the NW Appendix H. p. 4-6, Action Item 14.3 SIR 6.2/55tR 6.1.8 pump terbines and gow?rnor controls.
T R-069-M V Eliminate automatic control of the NW block valve Appendix H. p. 4-7, Action item 14.6 SIR 6.2/55tR 6.1.9 e= cept during a reactor trip.
6.1.12
z C
r$
O e
c*
90 c*
Iable 12.1 (Continued) m C
V Recommenda-SE R/551R 7
tion No.
Subject Source: RAW-1919 and other reports Section IR-0 70-Mf W Provide capability to override a "close" signal in the Appendin H, p. 4-7, Action item 14.7 SIR 6. 2/5',[R 6.1.9, MEW bloc k walve.
6.1.12 T R-0 71-Mf W Install valve position indication for the startup and M W Appendin H, p. 4-7, Action item 14.R
$[R 6.2/55f R 6.1.12 requiating valves.
T R-072-MW
[liminate the transfer f rom the startup to the MIW Appendin H
- p. 4-8, Action Item 14.9 SIR 6.2/55tR 6.1.12 flowmeter when the MfW block valve opens.
T R-073-Mf W (liminate high MFW pump discharoe pressure trips as a Appendix H, p. 4-8, Action Item 14.10 SI R 6. 2 common occurrence.
c*
1 R-0 74-M W Schedule I&C calibration and inspection work to minimite Appendix H. p. 4-8. Action item 14.11 SIR 6.2 7
the number of times the MFW pq and turbine y
instrumentation and controls are distorted.
TR-q75-MFW Modify control scheme for the heater drain puny recircu-Appendin H. p. 4-8 Action Item 14.12 SIR 6.2 lation control valves (for ANO-1 only).
T R-076-M W
[liminate automatic trip of the "preferred" N V pump Appendix H. p. 4-8, Action Item 14.13 SER 6.2 af ter a reactor trip (for AND-1 only).
T R-077-MF W Review and upgrade preventive maintenance on auxiliary Appendin H, p. 4-9, Action item 14.18
$[R 6.2 boilers.
1R-078-MFW Add an indication near the NW pump controls for MFW Apperwfin H
- p. 4-9, Action Item 14.19 SIR 6.2, 7.4 pump discharge pressure.
TR-079-MFW Put MW regulating valves, main block walves and startup Appendia H. p. 4-9. Action Item 14.20 STR 6.2/551R 6.1.8 control valves on a refueling frequency for an operational check.
T R-ORO-MF W Instrumentation to determine performance of MFWPi shaf t-Appendia H. p. 4-9. Action Item 14.21 SER 6.2 driven oil pump,
z C
- n mo eW fV lable 12.1 (Continued) us C
V Recomumenda-5( R/55t R
?
tion No.
Subject Source: BAW-1919 and other reports Section T R-081-M W Movem control room W W flow indication from back panel Appendia H, p. 4-9 Action item 14.22 Sit 6.2 to apron (for Rancho Seco only).
TR-082-MfW Add automatic bypass of Powdex (or condensate Appendin H, p. 4-10. Action item 14.24 SER 6.2 demineraliser) units on high differential pressure.
T R-083-MF W Add M W pump turbine lube oil purifiers.
Appendin H, p. 4-10. Action Itm 14.25 SLR 6.2 TR-084-M W Correct feed pump turbine shaft sealing problems.
Appendia H, p. 4-10. Action item 14.26 SIR 6.2 TR-085-M W Modify main TW pump recirculation valve for automatic 4 pendia H
- p. 4-10 Action Item 14.27
$[R 6.2/55tR 6.1.12 control during startup and shutdown.
w7 T R-086-MIW leproper draining of first-stage IW heaters Appendix H. p. 4-10 Action Item M.18 SIR 6.2 T R-087-Mf W Add capability for flushing the feed pump turbine Appendia H, p. 4-10. Action Item 14.29 SIR 6.2 gnvernor control oil system.
TR-088-MIW Eliminate automatic plant runback on low MIW pump Appendix H. p. 4-10 Action item 14.'A SER 6.2/55(R 6.1.12 discharge pressure or establish setpoint to achieve a successful runback.
T R-089-M W
[liminate potential for physical damage to condensate Appendia H. p. 4-10. Action item 14.31 5(R 6.2 and MfW pneumatic valve operator air supply lines.
TR-090-MfW Add valve position indication in control room for daer-Appendia H, p. 4-11 Action item 14.32
$(R 6.2/55tR 6.1.12 rator f eedwater tank inlet valves (for Davis-Besse only).
TR-091-MrW Eliminate need for an auniliary operator to open a Appendin H. p. 4-11. Action Item 14.33 SIR 6.2/55TR 6.1.9 deaerator feedwater tank drain line af ter reactor trips (for Davis-Besse only).
TR-092-MFW Assess the cause for f requent feed boc4ter pump Inw-
% pendia H, p. 4-11, Action item 14.34 SIR 6.2 suction pressure alarms (for Davis-Besse only).
0 6
6 1
9 6
I.
1 1
1 6
6 6
2 1
R R
R R
1 f
f t
(
5 5
5 5
5 5
5 5
6
/
/
/ /
4 3
2 3
R t
6 6
6 7
5 7
7 3
5 R
/
1 1
1 1 n 2
2 2
2 2
2 3
3 3
6.
6 5 o 6
5i 6
6 6
4 4
4 6
8 8
/t R
R R
R c R
R R
R R
R R
R R
I f.
f, e
(
I I
E I
E
(
I I
S MS 5
S S
S S
S 5
S S
S 5'
'5 5
6 7
3 3
3 3
II s
4.
4 4
II t
1 1
1 II r
V o
e m
e 2
4 p
e e
e l
R p 8
C e
t t
t C
5 a,
i r
I i
I i 1 D
H 2
3 5
5 5
D r
n n
n P
d P
2 2
e o
o o
na o
I I
I I
I I
i i
i ei h
t t
t t
o.
pd o M 8
8 V
V V
o c
c c
M pn N
5 4
A A
A Ae d
p 9
p p
p n
2
- p 5
1 4
5 a
1 1
1 4A 1
1 1
l I
I 8
8 1
I I
9 l
I I
B.
B.
p B.
p I
I p
1 4
4 4
I 9
e3 l
t I
1 p
p p
S -
p p
l i
I p
W
.I 2
2 V
V V
A I
P H
H H
I RI I
I B
B R
R R
V2 n
a n
a n
n a
a a
n a
n t
i
.C
)
e:
i d
id id iM pD id i
id i
i id it r
d d
d u
d c
n n
n n
,P n
n n
n n
n m
e r
e e
e e
e e
e e
e e
e p
p p
p p1 p
p p
p p
p 5
u W
p p
p p
p
.o.
p p
p p
p p
n A
A A
A A(N A
A A
A A
A A
i tn o
C
(
s r
p e
is e
1 a
t v
p d
g 1
C 1
m.
a e.
a n
a P
P 2
p e
D s.
a t
A F
1 h
d l
r l
r e
o r
r o
e I
f ic W
n v
o o
f W
o e
l T
i f
f tb w
(
r A
l
(
(
s a
e e
n t
n o
Y s
u I
t g
s g
i i
t p
n a
a r
p i
a t
1 e
n h
t e
m n
i M
ib d
s l
r a
s l
o o
e e
t d
t w
h o
t v
t i
t t
c a
l o
d c
r a
i r
a r
n
=
o c
l y
u r
3 v
h a
e f
a r
l o
o r
l
,s.
l r
s t
l v
m n
f t
a o
o o
m o
o e
f f
m s
r e We r
t i
/
o g
o s
t t
F r t
a l
S n
p a
n s
Mu n
r p
C n
n p
o y
d o
e.
p I
o g.
a i
s n) t y
c s
ee c
u s
g it u
iy p
t vc hl w
W i o l
n n
sn e
e o
F sr e
r i
ce M
sp v
n e
n o
ao t
n l
l i.
l a
i f
e e
i w
r e
t) f e s
t r
ct l
a.
p e
i o
a s
ay s
n r
W o
x n m
rl f s e
u F
f ea a
n l
c en oe d
t I
l n
no c
i po B
n n
np a oi d
r t
o s-o f
f o o t
t e
a i
n ga t
w m
s t s c
o o
~
rt ci t
ei n
nr n
o o
ei ev e
n n
c c
ia.
io n
l a
e a
p t
i wn f a h
g g
e n1 np d u
ou fD t
i i
t eP r n p
e s
s o
dH do a
u a
e r
h) e e
r i
l rd d
n e
l c eo sy d d
p ug Rt t n e r
t l n hf ul gn uo t(
l n e. e l
i Sa a
r e
a Mr r
t r
t f -
f o t s t
l el u
ol l
n oI.
c 0
es /
am a
i dt wg t o l
. e.
rM pM e
w cn ne ue u
f ut ei ar a)
)
j nr ui as l t l
r l o if rt t y ey o/
b i o da es as a
e cr vn en sl sl cS u
l f er l e vy v
v nh eo po nn un nC S
A(
Rd CB
( s t
O It R c Oc I o F o II
-a W
W W
5 W
W 5
S 5
S S
S d.
M M
M M
E M
f f
f 5
T F
P T
T C
C C
eo 0
M M
n I
I I
eN 3
4 5
6 7
8 9
0 1
2 3
4 n
n 9
9 9
9 9
9 9
0 0
0 0
0 on 0
0 0
0 0
0 0
1 1
1 1
1 co ei R
R R
R R
R R
R R
R R
R Rt T
T T
T T
T T
T T
T T
T 2C::5. U o
.yww
<E? "
n l
2C
- ms m
CD e
e*
Nw
,c
- Table 12.1 (Continued) vs C
E Recommenda-51R/551 R tion W.
Subject Source: B W-1919 and other reports Section c4 TR-105-ICS Perform field verification of ICS/NNI drawings.
4pendia R Yll.B.1, p VII-5 55tR 6.1.1. 6.1.8 TR-106-ICS Remove unused hardware from ICS/M41 cabinets.
Appenrtin R VII.B.I. p. VII-6 551R 6.1.4, 6.1.8 TR-107-ICS Improved maintenance and tuning of ?"
Appendia R VII.B.I. p. VII-6
$$ft 6.1.8 Appendia I. P. 111-10 Mo. PM-582; and Appendim I, p. 111-12. No. fH-061.
TR-108-M55 Investigate using manistan allowable set pressure for the Appendia M. p. 12-14. Item 11, 112.5 SC 6.4 lowest set M55Vs (or for TMI-I only).
U TR-109-M55 Ensure that relief valves not automatically isolated f rom A tendiv N
- p. 12-13. Item 10. 512.5 SIR 6.4 main steam system post trip are in a preventive main-e c'
tenance and test progrm.
m T R-110-M55 Davis-Besse should prow?de continuous [f W flow as a Appendia M. p. 12-13, item 6, 512.5 &
SER 6.4 function of level.
Appendia N, p. 12-14. Item II, 412.5 T R-111-RPS Review safety system surveillance procedures for checking Appendin K. p. III-1, $1 SER 4.4/551R 6.1.8 which channel is available for testing twfore initiation of test.
TR-112-PES Review switchyard maintenance procedures to ensure there Appendin K. p. III-l. 52 5tt 4.4 is no mechanism for less of offsite power.
TR-113-Pf5 Review breaker control power distribution to determine Appendin K. p. 111-2. %3 STR 4.4/55ft 6.1.7 effects of a loss of the battery bus.
TR-114-r[5 Evaluate hardware to ensure diesel generators cannot te Appendin K p.
!!I-2, $5 SER 4.4 synchronised to grid out of phew.
T R-115-f*t 5 Test diesel ge-trators to assure they will carry loads Appendia K. p.
III-2, %
SIR 4.4 under expected sequential loading conditions.
4 I
sc C
'A3 rn o
e p.a to 00 Table 12.1 (Continued) la (n
C
]$
Recommenda-SER/SSER tion No.
Subject Source: BAW-1919 and other reports Section na TR-116-PES Review dc charging system and ensure the charging voltage Appendix K, p. III-3, 57 SE R 4.4/SSE R 6.1. 7 does not exceed plant equipment voltage rating.
TR-117-PES Notify inverter overcurrent protection to ensure the Appendix %, p. Ill-3, 58 SER 4.4/SSER 6.1.7 breaker / fuses open on overcurrent before inverters fall.
TR-118-PES Evaluate loadings on ac and dc vital buses to ensure Appendix K p.
III-3, 59 SER 4.4/SSER 6.1.7 adequate margins exist without trip of equipment.
TR-119-PES Implement preventive maintenance for elettrical buses.
Appendix K
- p. III-3, 510 SER 4.4/SSER 6.1.7 TR-120-IAS Check 0-rings in r_ritical air-operated valves.
Appendix M. p. 5-2, 55.2.1 SER 6.5 s*
Sf TR-121-IAS Make appropriate personnel aware of importance of Appendix M, p. 5-2, 55.2.2 SER 6.5/SSER 6.1.9
["
instrument air system, prohibition of use for tools, and need to report air system damage.
TR-122-IAS Systematically inspect instrument air systems for leaks.
Appendix M, p. 5-3, $5.2.4 SER 6.5 TR-123-IAS For instrument air systems, protect against failures Appendix M
- p. 5-3, 55.2.5 SER 6.5 possible with desiccant-type driers.
TR-124-IAS Identify metal lines with high vib ition in instrument Appendix M. P. 5-4, 55.2.6 SER 6.5 air systems, and when cracks are $ und, replace metal lines with flexible tubine.
TR-125-IAS Test critical air-operated valves for operability in Appendix M. p. 5-4, 55.2.8 SER 6.5 the preventive maintenance program.
TR-126-IAS Compare instrument air system configuration with Appendix M, p. 5-5, 55.2.10 SER 6.5 functional target criteria.
TR-127-IAS for instrument air system, review preventive maintenance Apprndix M, p. 5-5, 55.2.11 SER 6.5 program, identifying parameters for trending to deter-eine maintenance requirements.
N=?M Table 12.1 (Continued) en5 Recommenda-SER/SSER T3 tion No.
Subject
$ource: BAW-1919 and other reports Sec tion -
6*
TR-128-IAS Review training and loss of air response procedures for Appendix M, p. 5-5, 65.2.12 SER 6.5/55ER 6.1.9 instrument air system.
TR-129-IAS Install automatic bypass line around driers and filters Appendix M. p. 5-5, %5.3.1.1 SER 6.5 (for ANO-1 only).
TR-130-IAS Expand procedure for the loss of instrument air (for Appendix M. p. 5-6, 55.3.1.2 SER 6.5/55ER 6.1.9 AND-1 only).
IR-131-IAS Investigate feasibility of routing instrument air com-Appendix M. p. 5-6, $5.3.2.1 SER 6.5 pressor intakes to the exterior (for Oconee units only).
TR-132-IAS Add an af ter-drier to the instrument air line (for Oconee Appendix M, pp 5-6 & 5-7, 55.3.2.2 SER 6.5 g
to units only).
5 TR-133-IAS Add a filtration system downstream of the last drier in Appendix M. p. 5-7, $5.3.2.3 SER 6.5 the instrument air system (Oconee units only).
TR-134-IAS Install control room-operated isolation valves with Appendix M
- p. 5-8, 55.3.2.4.1 SER 6.5 mancal bypass at the key line feeding each unit's auxiliary building instrument air system header (for Oconee units only).
l TR-135-IAS Install automatic isolation valves that could limit Appendix M. p. 5-8, 55.3.2.4.2 SER 6.5 instrument air system leaks (for Oconee units and CR-3).
TR-136-IAS Install a dewpoint mordtor downstream of instrument air Appendix M. p. 5-8. 55.3.2.5; SER 6.5 I
system driers (for Duke FPC and TED only).
Appendix M. P. 5-11, 55.3.3.3; &
Appendix M, p. 5-19, 55.3.7.1 T R-137-I AS Check accumulators in instrument air system for water Appendix M, p. 5-8, $5.3.2.6 SER 6.5 buildup. Install drain valves where necessary (for all Appendix M, p. 5-12, %5.3.3.6; &
operating plantt, except ANO-1).
Appendix M. p. 5-15, %5.3.4.3
i C
PD rn c) e 5
oo j'
Table 12.1 (Continued) o, 4
2 c
l$
Recommenda-SER/SSER tion No.
Subject Source: BAW-1919 and other reports Section n'
TR-138-IAS Install a check valve after each compressor aftercooler Appendix M. p. 5-9, $5.3.2.7 &
SER 6.5 in instrument air system (for Duke and frC only).
Appendix M, p. 5-13, 65.3.3.7 TR-139-IAS Install on/off status and remote start of instrument air Appendix M. p. 5-9, $5.3.2.8 &
SER 6.5 compressors in the control room (for Duke and FPC only).
Appendix M, p. 5-10, 65.3.3.1 TR-140-IAS Assign high maintenance priority to an out-of-service air Appendix M, p. 5-9 & 5-10, 55.3.2.9 SER 6.5 i
compressor and maintain suf ficient spare parts to repair a compressor within a week (for Oconee units only).
TR-141-I AS Install an automatic bypass valve to bypass driers Appendix M pp. 5-10, 5-11, 65.3.3.2; SER 6.5 filters upon loss of instrument air header pressure (for Appendix M, p. 5-18, $5.3.6.1; &
Appendix M, p. 5-20. S5.3.8.1 B
l
[$
TR-142-IAS Design the components of the instrument air system to Appendix M. p. 5-11, 55.3.3.4 SER 6.5 withstand maximum flow oenerated by all the compressors (for FPC only).
TR-143-1AS Inspect accumulators and their check valves in the in-Appendix M, p. 5-12, $5.3.3.5; SER 6.5 strument air system (for FPC, GPUN, SMUD, and TED only).
Appendix M. p. 5-15, $5.3.4.2; &
Appendix M, pp. 5-16 & 5-17, S5.3.5.2 TR-144-IAS Develop or upgrade a loss of-instrument air procedure Appendix M
Appendix M. p. 5-17, $5.3.5.4; &
Appendix M, p. 5-22, 55.3.8.4 TR-145-IAS Install automatic isolation valves in instrument air-Appendix M, p. 5-14, 55.3.3.9; SER 6.5 lines (for CR-3 only).
TR-146-IAS Loss of air procedure for instrument air system should Appendix M
- p. 5-14, 55.3.4.1 SER 6.5 note importance of quickly bypassing driers and filters whee excessive flow rates are experienced (for TMI-1 only).
TR-14 7-I AS Normally open or closed positions are recommended for Appendix M, p. 5-16, %5.3.4.4 SER 6.5' l
certain valves (for TMI-1 only).
~ -
2 C$?
M w
H Table !?.1 (Continued) v?
Recommenda-SE R/SSE R tion No.
Subject Source: BAW-1919 and other reports Section g
IR-148-IAS Install automatic isolation valves at specified points Appendix M, p. 5-16, 55.3.5.1 SER 6.5 in instrument air system (for SMUD only).
TR-149-IAS Design instrument air system components to withstand Appendix M
- p. 5-19, $5.3.7.2 SER 6.5 maximum flow generated by all compressors (for 1[D only).
TR-150-IAS Eliminate the ESFAS signal to close specified valves and Appendix M
- p. 5-21, 55.3.8.2 SER 6.5 isolate service and control air (for IVA only).
TR-151-IAS Eliminate apparent inconsistencies in instrument air Appendix M
TR-152-IAS Establish same run time for the various compressors in fppendix M, p. 5-22, $5.3.8.5 SER 6.5 g
the instrument air system (for (IVA only).
TR-153-IAS Evaluate a plant-specific air system failure.
Appendix M, p. 6-10, 66.7.1 SER 6.5 TR-154-ICS Provide operator with unambiguous status of indicators Appendix R, VII.B.1, p. VII-6 SSER 6.1.3, 6.1.6, 6.1.9 and recorders in main control room on loss of ICS/NNI power or signal.
TR-155-EFW Limit maximum flow rate delivered by the EFW system.
Appendie P, Vol. II, p. B-12 $3.A; SER 6.3, 4.2, 5.4/
Appendix Q, p. 2-2, $2.1.3; &
55ER 6.1.10 Appendix Q, p. 2-3, 52.1.4 TR-156-OPS Provide a sesignated "phone talker" to relay Emergency
' Appendix 5, p. 3-2, $3.1.2 SER 7.5/55ER 6.1.9 Plan messages.
TR-157-OPS Validate E0Ps to determine if adequate staffing and Appendix S, p. 3-2, $3.4.2 SER 7.5 prioritization exists TR-158-OPS Re-evaluate annunciator designs to ensure key alarms do Appendix 5, p. 3-5, $3.6.2 SER 7.5/SSER 6.1.6 not go unnoticed.
m m-e
"'wT'&
e ie e
a*w e
h s
i b
a;
?::
w f*
Table 12.1 (Continued) on C
l3 Recommenda-SER/SSER tion No.
Subject Source: BAW-1919 and other reports Section
>4 IR-159-0PS Evaluate secondary system controls to achiese remote Appendix N, p. 12-13. Item 7, $12.5 &
SER 4.2/SSER 6.1.6 manual control in the main control room of all post-trip Appendix S. p. 3-8, $3.9.2 steam flow paths, MFW and EFW.
TR-160-EfW Evaluate ability to extend the time to achieve design EFW Appendix Q, p. 2-1, 92.1.2 SER 6.3 flow.
l 1R-161-EFW Evaluate the margin between the EFW and MFW low-level Appendix Q
4
~
TR-162-EfW Modify EFW flow control to provide smoother flow control Appendix N. p. 12-13 Item 6, $12.5; SER 6.3 rather than on-off control.
Appendix N. p. 12-14, item 11,
$j Appendix Q, p. 2-4, % 2.1.6; &
s Appendix S, p. 3-9. 53.10.2 na TR-163-EFW Review EFW survelliance and test procedures to ensure Appendix Q, p. 2-5, S2.2.2 SER 6.3/SSER 6.1.8 r
that components used in the E0Ps are included in the test program.
TR-164-EFW Review EFW Preventive Maintenance Program, including Appendix Q
- p. 2-5. 52.2.3; Appendix Q.
SER 6.3/SSER 6.1.8 minimizing potential for common cause failures arising
- p. 2-6, 52.2.6; & Appendix Q, p. 2-6, from maintenance and testing procedures.
52.4 TR-165-EFW Review EFW maintenance and testing procedures to elimi-Appendix Q, p. 2-6, S2.2.4 SER 6.3/SSER 6.1.8 nate conflicting and confusing instructions.
t TR-166-EFW Implement a program to improve and maintain the avail-Appendix Q, p. 2-6, $2.3 SER 6.3/SSER 6.1.8 ability and performance of the EFW systems.
T R-167-PES Include in operating procedures guidance on restoration Document No. OC-1-85-06, "TAP Report,"
SSER 6.1.6 of power to electrical buses, especially if the ICS or
- p. 7; Document No. 1MI-86-02, "TAP i
j ICS controlled equipment is affected.
Report," p.
19; & Dccument No. RS-86-01, 4
"TAP Report," p. 46 3
1 I
4 l
4 i
~
.=
I z
hh nn
??
w S$
d" Table 12.1 (Continued)
- i La E
}3 Recommenda-SER/SSER tion No.
Subject Source: BAW-1919 and other reports Section w
TR-168-MTS Provide guidance in procedures when troubleshooting the Document No. OC-2-85-02, "TAP Report,"
- p. 6 & Document No. OC-2-86-01, "TAP Report," p. 6 TR-169-MTS Evaluate possibility for defeating the high vibration Document No. D-B-1-85-04, "TAP Report, SER 8.3/SSFR 6.1.12 trip during main turbine valve testing (for GE turbines
- p. 5 & Document No. 47-1168891-00, only).
"Review of Reactor Trip Initiating Events at the BWOG Plants, 1980-1985." pp. V-3, V-7, & V-8 TR-170-MFW Evaluate placing orifice snubbers in the MFW pump control Document No. ANO-85-06, "TAP Report,"
SER 6.2/SSER 6.1.12 pa oil system.
- p. 4 i
na
/o TR-171-OPS Evaluate alare setpoints to determine if adequate time is Document No. OC-3-86-02, "TAP Report,"
SSER 6.1.9 ha provided for operator response.
- p. 6 TR-172-PRV Evaluate PORV circuitry to determine if momentary loss of Document No. TMI-86-02, "TAP Report."
SSER 6.1.6 power or restoration of power can cause PORV to open.
- p. 27 & Document No. 0C-3-86-04 "TAP Report " p. 12 TR-173-MrW Ensure in procedures that MFW pump status to ARTS /RPS is Document No. ANO-85-05, "TAP Report,"
SER 6.2/SSER 6.1.12 reset af ter each MFW pump is operational.
- p. 5 TR-174-MSS Improve response of modulating turbine bypass valves.
Appendix N, p. 12-14, Item 11, 512.5 SER 6.4 TR-175-PRV Fnsure the PORV block valve functions as designed under "B&W Owners Group Review of June 9,1985 SER 8.4 transient conditions.
Davis-Besse Loss of Feedwater Transient,"
BWOG-1154 Task For2e, August 1986,
- p. 3-2fs & Document No. 47-1165733-00, Review of Category B and C Events at the B&WOG Plants 1980-1985," 57.0, Item 2.0 1
m z
L ER?
L ff Table 12.1 (Continued) t us 45 Recommenda-SER/SSER l'
tion No Subject Source: BAW-1919 and other reports Section TR-176-POV Ensure power-operated valves employed in controlling Appendix H. "BWDG Review of the June 9 SER 8.4 post-trip energy balance between primary and secondary 1985 Davis-Besse loss of Feedwater system are subjected to confirmation of ability to Transient," B&WOG lask Force 1154 &
perform their function.
Appendix I, "Review of Category B and C Events at the B&WOG Plants 1980-1985,"
Document No. 47-1165733-00 TR-171-OPS Review E0Ps to ensure that wherever drastic actions are Appendix S
- p. 3-5, 53.5.1 SER 7.5 I
specified, plant conditions require the action.
TR-178-ICS Ensure plant goes to a known safe state on loss of power B&W b...e.s Group I&C Committee Meeting SSER 6.1.6, 6.1.9 to the ICS/NNI.
Minutes, March 20, 1986; B&WOG SPIP-1A8 sa Meeting Minutes, March 31-April 1,1987; 93
& Appendix R, ICS/NNI Evaluation Fir.al na Report, p. LC-1 (Known Safe State on the 6*
Loss of ICS/NNI Power)
TR-179-MFW Identify areas for enhancing the eliability of the MIW Appendix P, Vol. II. p. B-8, $2.A &
SER 6.2 and condensate systems and controls.
- p. B-9, $2.B.7 & $2.C TR-180-MTS Provide a monitoring capability for the EHC system for Appendix B-2, p. 8-3, 58.2.1 SER 8.3/SSER 6.1.12 purpose of root cause determination.
T R-181-OPS Verify adequacy of instrumentation and displays used to Appendix I, p. III-14, No. HI-C2 SER 7.5/SSER 6.1.6, 6.1.9 assess and control the ATOG stability parameters.
TR-182-ICS Evaluate installing automatic bus transfer switches of Appendix R VII.B.2, p. VII-6 SSER 6.1.7 MFW pump controllers (or Davis-Besse only).
TR-183-ICS Preventive maintenance and testing for ABT switches.
Appendix R VII.B.2, p. VII-6 SSER 6.1.7, 6.1.8 TR-184-ICS Provide separate fuses for hand stations that use ac Appendix R VII.E, p. Vll-11 SSE R 6.1. 7 power.
TR-185-ICS Power feedwater flow recorders directly from NNI.
Appendix R VII.E. p. VII-11 SSER 6.1.6
t i
sc C
r0 c)
Table 12.1 (Continued) us s
i La Recommenda-SER/SSER 45 tion No.
Subject Source:. BAW-1919 and other reports Section o
f TR-186-ICS Minimize access to ICS/NNI cabinets during operation, Appendix R VII.B.2
- p. VII-8 SSER 6.1.7, 6.1.8 t
E' and train maintenance (ersonnel on location of power distribution components.
TR-187-ICS Install current and voltage meters for NNI power Appendix R VII.B.2
- p. VII-8 SSER 6.1.7 supplies (for Davis-Besse only).
TR-188-ICS Maintain dc power supply current balance and perform a Appendix R VII.B.2, p. VII-8 SSER 6.1.7, 6.1.8 periodic full load test for each power supply.
TR-189-ICS Set selector switches to select maximum NNI dependence.
Appendix R VII.B.2, p. VII-7 SSER 6.1.3 j
TR-190-ICS Develop backup controls for puressurizer level and Appendix R VII.B.2, p. VII-10
$$ER 6.1.4, 6.1.6 i
pressure control.
pa to I
TR-191-ICS Separate condensate flow from NNI power (for CR-3 only).
Appendix R VII.B.2, p. VII-7
$$ER 6.1.6 TR-192-ICS Remove / modify NNI power supply and signal select logic Appendix R VII.E. p. VII-12 SSER 6.1.7 (for Rancho Seco only).
TR-193-ICS Review / test pressurizer heater low-low level interlock Appendix R VII.E. p. VII-12 SSER 6.1.4 logic.
TR-194-ICS Buffer hand powered indicators and recorder inputs t rom Appendix R Vol. 1, pp. 5-3 & 27-2 SSER 6.1.4 automatic power signals.
l TR-195-ICS Supply hand and automatic powered circuits from separate Appendix R VII.B.2, p. VII-6 SSER 6.1.6 panels.
TR-196-ICS Set pressurizer level signal select relays to automatic Appendix R VII.B.2, p. VII-7 SSER 6.1.3 powered transmitters.
TR-197-ICS Provide automatic power transfer for the modulating Appendix R VII.B.2, p. VII-7
$$1R 6.1.6 pressurizer heater converters.
i I
2c:
i pp on o
e 1
he BJ La f*
Table 12.1 (Continued)
La C
l3 Recommenda-SER/SSER tion No.
Subject Source: BAW-1919 and other reports Section TR-198-ICS Automatic powered reactor inlet and out temperature Appendix R VII.B.2, p. Vll-7 SSE R 6.1. 3 -
sensors should be selected, or logic changed to auto-natically select auto powered sensor on loss of hand power.
TR-199-ICS Failure of inputs to RC pump interlocks must not prevent Appendix R VII.E
- p. VII-12 SSER 6.1.4 pump testart.
TR-200-MIS Install a time-delay relay or an orifice between the EHC Document No. 47-1168891-00, Sept. '87, SER 8.3/SSER 6.1.12 oil system and the ARIS sensing line to prevent oil "Review of Reactor Trip Initiating pressure perturbations.
Events at the B&WOG Plants 1980-1986."
p.
IV-4, %IV.B.I.C
$j TR-201-MIS Review EHC ove ++ peed and fast control and intercept Document No. 47-1168891-00, Sept. '87, SER 8.3/SSER 6.1.12 e
valve circuits.
"Review of Reactor Trip Initiating C3 Events at the B&WOG Plants 1980-1986,"
p.
IV-4, $lV.B.I.D T R-502-MFW Review FW system procedures to determine if switchover Document No. 47-1168891-00, Sept. '87, SER 6.2 of steam supplies can be made at lower power levels.
"Review of Reactor Trip initiating I
Events at the B&WOG Plants 1980-1986,"
- p. IV-5, $1V.B.2.F TR-203-PES Establish preventive maintenance to increase reliability Document No. 47-1168891-00, Sept. '87, SSER 6.1.7, 6.1.8. 10.3 of inverters.
"Review of Reactor Trip Initiating Events at the B&WOG Plants 1980-1986 "
- p. IV-6, %IV.B.3.b TR-204-ICS Eliminate or reduce automatic ICS runback rate on Document No. 47-1168891-00, Sept. '87, SSER 6.1.4 asymmetric rod conditions.
"Review of Reactor Trip Initiating Events at the B&WOG Plants 1980-1986 "
i p.
IV-6, %IV.B.S.c i
l TR-205-RPS Evaluate lowering the low RCS pressure trip setpoint Document No. 47-1168891-00. Sept. 'a7 SSER 10.3 i
(for Davis-Besse only).
"Review of Reactor Irip In'tiating Events at the B&WOG Plants 1980-1986,"
- p. IV-7, $1V. B.6. f i
I l
t t
,;i' 6i!
l'l i
,ft 9
1 6.
R E
8 4
8 6
9 S
3 3
S 4
1 1
R
/
1 1
6 6
1 1
4 1
0 0
E n 3
2 6
6 6
S o Si 8
6
/t R
R R
R R
R R
R R c R
R E
E E
E E
E E
E E e E
E S
S S
S S
S S
S SS S
S S
S S
S S
S S
S
~
t 7
m c
7
",a re 8
.n 8
o 9
6 9
6r oj E. i E.
1 8
1 8g t o s
t I t I
g9 g9 o ar r
1 I c r
I
.n1
.n1 r rP o
8 V e e
1 V
ti -
ti P2 e
p 5
S pt0 pt0 8
pw n
ea8 ea8t9 Oe e
n c
r I
o.
c o
l o
Si9 Si9n1 r
t -
D V
t
,i
,i m. ",i H
i8 o
D o
i t1 t1 e vr c
0ns 0nsst 0e e
3 cV l
o eI q
e 0I t8 0I t sp 0 t "s h
t 8
N S
l p
S n-nee
- n o
5 p
p 1 paV 1 pass 8I e
,p p
u c
9il 9il s 6
s 7l n d
2 2
u S
o 2
8rPd 8rPA n
6 2
1 S
D 1
8T n
8T 1
4 eo 6
Ga 6
Gt0 6 np 5
a 3
I 1 rO 1 rOn-1 ns 8
1 I
9 1
I 3 8
l I
1 oW,
1 owe 3 1 oe 1
1 V$
1 1
p V
- t&7 t&i8
- sR p
7 cB -
7 cBs-7 r 9
p u
4 a V
4 a nO 4 ef 1
pd.
p p
S p
ee eeaN
.P o p
W 2
.Rh
.Rh rA A
o t3 o
tT oen B
8 I
R c R
R R
R Nf Nf
. Nco o
otV ot o
ni
)
e i
i ixD i
x x
x t
a t
a
.N t at x
x x
nw
, nw 2
nna d
c n
n nl.
d id i
i ees2 ees8t eel d
d d
d d
n n
n n
mit -
mit - r mti e
r e
e e p A.
e e
e e
uvnV uvnVo unp u
u p
p pp p
p p
p cee cee p
cim oR v
.e oao n
o p
p puV p
p p
p oR v.
i S
A A
ASI A
A A
A D"
E p D"
E pR DMC tno C(
ss.
nn oo 1
e m
/
ii e
2
.l o
n d
l r tt t
1 mp r
o e
eo cr a
l ss u
s a
ip iz v
ue l
em l
f e
e t a a
r ce rs o
g i n t n s
b y
n l
r r
i si i
r T
e l b n
a sd a
a ia S
e cu u
o T
d m
e p
C ot eo n
r n
ar i p t
La g
H e
m I
t l
dr Ed h
e t
e. c e
ri v
e n
t t
n r
c n
g s
i m
ea s
i h
o y
da i t
e od o
t,
t s
l rr t
e vt ro d
k o
ag a
d ol pr l
r oo m u
n nn d
ia r
o t
br o
o ds ol o
t a
r n
p t
t ee t o i
l g
t o
t u
cr r
t il e
n c
ig a
2 a
et a
oi r
o un 5
l y rn v
o c
t ci e
pl uo i
g h
n rn v
d t
dc t
ng t
r e
ii o
n ec e
s in o
v ca m
a be cd a
ti i
t e
r e
r oe w
1 di rd S
at
. i r
rt r.
l d pn E.
ca gp n
p o
e i c ni o
n.
f e or u
e s
ri ir m
oo c
t u s. on nm ee br nt ti sn l
es ha om rc ub. i -
o t
da ai h n sc io ui l uk at t
sa rn ca co tc sv l n rs rr ae if ti st ce sr f
a to m
eu dt t
it ea nr ee y
m oW g
itt nn ar wi ch u
rs ont p
a i
r r
a ai ce ss it f p p
w o
v l o p
ms t a i sm f o I p ess at Cm t eo tF o
i
~
p N dep ms R u i pt aM r
l o i
rit r
p t
t n d
Nd hi p
upo ef l n ai oe /e ecr Do w
pdb po h
ai h
ml Sp vtt Rt ol n
o s
n td g
C p ii C
l a gg e
pn Ii t wr w
i t
eat l
i c
vta wo l
i n yd oi r
cso eo et e
os er b
si f u l s l t e
t sh rn j
r e
it a
o il e
e t pc i
ue oi a vr ss b
pd n vn t
dg rc vI b
u mni eo s
dn en eN as rre eo ns S
I al R c E
Ao Vi DN L v Pt r Rf E e a
S S
S S
S S
S M
D P
T P
C C
C C
C D
R C
d.
M O
I I
I I
I A
C R
neo mN 6
7 8
9 0
1 2
3 4
5 m
0 0
0 0
1 1
1 1
1 1
on 2
2 2
2 2
2 2
2 2
2 co ei R
R R
R R
R R
R R
R Rt T
T T
T T
T T
T T
T c:Dn :
o?
0j1~
o 4j3 d
, n g,o
,a n
3C3r e
j' i!?
1 4
2 il i; j 4
I l
- - _ =.
=
i l
2 C
t
- o rn O
e H
ro W
l Table 12.2 SPIP reconunendations rejected by the UWOG steering communittee Note:
The 21 recosamendations listed here were rejected by the BWOG steering conunittee; thus, permanent numbers were never assigned y
to the recommendation and working (temporary) numbers identify the rejected recommendation. The source document cited by the steering committee as its basis for rejection, the date of the steering comunittee meeting at which the recommendation was rejected, and the SER or 55ER section that treats the rejection are given.
g Basis for rejection SER/SSER No.
Subject Source document (date)
Section 16 Evaluate the adequacy of plant comununi-BAW-1919, Appendix I, "Review of This item was addressed on a plant-SER 4.2 cation systems, to include:
Category 'B' and Category 'C' Events by plant basis for the 1154 Task at B&WOG Flants, 1980-1985,"
Force and is documented on p. 3-40 47-1165733-00, September 1986, of that report.
a.
The ability to contact plant aux-
- p. III-18.
iliary operators following a trip.
U b.
The availability of comununications 4
equipment at key operating locations, w
such as the EFW and TBV control sta-tions and other components / locations isportant to the control of FW and steam flow where manual control actions may be required.
22 Establish a program to reduce reactor Ibid, p. III-25.
This is the purpose of SPIP and the SER 4.2 trips and improve plant post-trip response. program is being conducted.
The program should focus on two key areas:
(October 28-29,1986)
^
a.
Identification and correction of recurring problems at their plant.
b.
Preclusion of problems with poten-tially severe consequences based on the collective experience of the BWDG plants.
i
.. =_
B 2
E8 U
Table 12.2 (Continued)
]$
on 45 Basis for rejection SER/SSER 13 No.
Subject Source document (date)
Section
>^
23 The Transient Assessment Committee should Ibid, p. III-25 This review is within the capabil-SER 4.2 review the TAP Report guidelines to deter-ity of the Transient Assessment mine what additional information is Committee.
necessary to document and communicate the transient experience and lessons learned.
(October 28-29, 1986)
Areas for potential inclusion in TAP Reports are:
a.
Listing of events of a similar nature, b.
Discussion of transient response had the event occurred under different sa initial conditions, such as: startup y
mode vs. full power; low decay heat na vs. high; backshift vs. day shift.
oo 25 The BWDG should undertake a study of Ibid, p. 111-28 This study is within 'ba capabil-SER 4.2 reactor trips following outages or pre-ity of the Transient Assessment vious reactor trips (repeat reactor trip)
Committee.
to identify the causes of, and recommen-dations that could be made to reduce the (October 28-29, 1986) number of, such reactor trips.
A-1
- 1. Establish a BWOG Task Force to develop BAW-1919, Appendix N, "B&WOG Avail-A Valve Task Force was established S[R 6.4 and implement an MSSV performance ability Committee: Main Steam to handle these and other valve-improvement program as described in Pressure Control Review," Section related issues.
Section 12.1 of the SPIP report.
12.1, 47-1165943-00, p. 12-2.
"Main Steam Pressure Control Review."
(January 14-15, 1987)
- 2. Based on the results of Phase 1, raise M55V design pressure if necessary..
- 3. Based on monitoring and evaluation of future performance, initiate work to reduce lifting of MSSVs.
P
5=?
w N
p Table 12.2 (Continued)
(n E
Basis for rejection SER/SSER
?
No.
Subject Source document (date)
Section 5.2.7 The compressors should be isolated and/or BAW-1919, Appendix M "Instrument Other recommendations deal with SER 6.5 separated as part of the redesigning and Air System Review Report," Section improving response on loss of air.
upgrading of the air supply system.
5.2.7. 47-1165 % 5-00, December 1986, This is not a feasible solution.
- p. 5-4 52.9 The status of leakage in the air system Ibid, Section 5.2.9, p. 5-4.
This is not feasible or practical.
can be determined by observing the run times of the compressors and the discharge (January 14-15, 1987) from the receivers. However, some plants have the compressors located in different places throughout the buildings. This layout makes it difficult to quickly e
determine the total air flow in a loss-of-P,0 air event. Flowmeters should be installed ro to enhance system diagnostics.
so 3b If EFW flow-limiting means are not adopted, BAW-1919, Appendix P, "A Compara-Reconnendation is not justified, SER 5.4 analysis and tests should be performed to tive Study of the Sensitivity of since it is impofiart to limit the demonstrate, if possible, that a head ade-B&W Reactor Plants," Vol. II, MPR an3ximum EFW by such mean, as for quate tc w intain natural circulation of 948. March 1987, p. B-13.
example,t;avitating venturls, an reactor coolant is developed by injection automatic control based on obtain-of emergency feed, without raising the ig a pre-selected EFW flowrate, or level of condensed liquid in the steam incorporation into existing auto-generator above its normal low level value matic level controls, of feedback (roughly 5% full).
from one of the existing EFW flow instruments, (May 5-7, 1987) 2.1.7 Although each utility has taken action to BAW-1919. Appendix Q, "Auxiliary /-
14t necessary beyond what was done SER 6.3/55ER 6.1.12 reduce the propensity for turbine overspeed Emergency Feedwater System Review by the 1154 Task Force based on trips, further actions are recosamended to Final Report," 47-1168159-00, operating history and AP&L and T[D.
enhance resistance. It is recommended that May 1987, p. 2-4.
the BWOG undertake swe or all of the (May 5-7, 1987) c following activities:
2 C-30 rn Q
s 9a r0il Table 12.2 (Continued) vi 55 Basis for rejection SER/55tR 3
13 No.
Subject Source document (date)
Section
- a. Sponsor an R&D* effort to determine if more reliable controls are available, including an investigation of non-traditional sources such as aerospace turbo pump controls.
i
- b. Develop an information exchange program that provides each owner with results of successful changes made to improve the system arrangement and configuration.
fj 2.2.5 Training of plant personnel should be ccn-Ibid, p. 2-6 Although this is a good idea, it is SER 6.3/55ER 6.1.9 ducted specifically in areas where repeti-not limited to EFW but is included e
@j tive problems have occurred (for example, in the SPIP recommendations covering motor-operated valve (MOV) testing and a broader range of plant areas.
maintenance, and turbine controls testing and maintenance). Repetitive problems may (May 5-7, 1987) also be indicative of design weaknesses, and design reviews may be necessary.
6 CR-3 experienced a steam generator over-TAP Report CR-85-01, p. 6.
This is included in general train-SSER 6.1.10 fill due to the resetting of the EFIC ing at the utilities.
system trip by maintenance personnel after its inadvertent actuation. Review this (May 5-7, 1987) event as part of the Operations, Mainte-nance, and I&C personnel training programs.
Emphasize the need to establish communica-tion with the control room before taking action on safety system equipment if an
]
unplanned actuation occurs during mainte-nance, testing, etc.
- Research and development.
I
z E8 13 H
Table 12.2 (Continued) t)
1
?
Basis for rejection SER/SSER No.
Subject Source document (date)
Section
(
g 11 Provide indication of the status of the TAP Reports DC-2-82-01, p. 3 Not justified SSER 6.1.12 MFW pump and turbine trip bistables into OC-3-84-04, p. 5; & OC-2-87-03 ARTS /RPS in the main control room.
- p. 2.
(May 5-7, 1987)
T-5 Implement a program that identifies the BAW-1919, Appendix B-?. "An Evalua-This is part of the Transient SER 8.3 cause of turbine trips and records data tion of Turbine Generator Ciused-Assessment Committee's charter.
for identification of pattern development.
Reactor Trips in B&W Nuclear Power Plants," Pickard, lowe, and Garrick (October 28-29, 1986)
Inc., PLG-0444, Section 8, December 1985, p. 6.
N TIR-7 Include the plant response for a turbine Ibid, pp. VI-31 & VI-32 Standard industry practice is to SSER 10.3 7
trip runback in the operator training retrain after modifications W
program.
(September 9-10, 1987)
TIR 8 Review ICS tuning to ensure control Ibid, pp. VI-31 & VI-32.
ICS tuning is not a function of SSER 6.1.8. 10.3 settings are compatible with a turbine ARIS setpoint. Retuning is trip runback from less than 45% power.
accomplished by TR-107-ICE.
(September 9-10, 1987)
TIR-19 The plants that trip both MFW pumps on Ibid, pp. VI-29. VI-36, & VI-37.
Trip is the correct consequence for SSER 6.1.10, 10.3 higti SG 1evel should evaluate sequential this situation.
vs. simultaneous MFW pump trip.
(September 9-10, 1987)
TIR-24 The BWOG I&C Committee should develop a Ibid, pp. v-70 & V-71 This has no benefit to reduce trips SSER 6.1.6 post-trip or post-transient troubleshoot-or complex transients ing procedure to aid in diagnosing ICS module failures.
(September 9-10,1987) l
2 q
c m
1 m
i O
a
?
w (n
C "U?
V i
L Table 12.2 (Continued)
Basis for rejection SER/SSER 1
No.
Subject Source document (date)
Section TIR-26 Review and upgrade maintenance procedures Ibid, pp. V-82 & V-83.
Not a feasible or practical solu-SSER 6.1.8, 10.3 as necessary to ensure that proper check-tion to the problem.
out of replacement modules and assemblies in the CRDS is made prior to their use.
(September 9-10, 1987) t emO This recommendation should also be i
N extended to ICS components and EHC com-ponents, circuit boards, etc.
r i
TIR-28 Each BWOG plant should review it CRD Ibid, pp. V-80 through V-84.
This industry issue is being SSER 10.3-I cabinets for proper labeling and con-addressed on a braoder plant-j sistency with procedures.
wide basis.
?
(September 9-10, 1987) 6 I
k
APPENDIX A BIBLIOGRAPHY BAW-1919 Babcock & Wilcox Owners Group, "Safety and Performance Improvement Program," BAW-1919, Revision 4, April 1987, and Revision 5, July 22, 1987.
BAW-1564 Babcock & Wilcox, "Integrated Control System Reliability Analysis," BAW-1564, August 1979.
BWOG 47-1168891-00 Babcock & Wilcox Owners Group, "Review of Initiating Events at the BWOG Plants, 1980-1986," BWOG Report 47-1168891-00, September 1987.
Generic Letter U. S. Nuclear Regulatory Commission, Generic Letter 85-13 85-13 transmitting NUREG-1154 regarding the Davis-Besse loss of main and auxiliary feedwater event, August 5,1985.
IE Bulletin 79-27 U.S. Nuclear Regulatory Commission, "Loss of Non-Class 1E Instrumentation and Control Power System Bus During Operation," IE Bulletin 79-27, November 30, 1979.
IE Bulletin 80-04 U.S. Nuclear Regulatory Commission, "Analysis of a PWR Main Steam Line Break With Continued Feedwater Addition,"
IE Bulletin 80-04, February 8, 1980.
IEEE Standard Institute of Electrical and Electronics Engineers, "Cri-279-1971 teria for Protection Systems for Nuclear Power Generat-ing Stations."
i NUREG-0667 U.S. Nuclear Regulatory Commission, "Transient Response of l
Babcock & Wilcox-Designed Reactors," NUREG-0667, May 1980.
j NUREG-0737 U.S. Nuclear Regulatory Commission, "Clarification of TMI Action Plan Requirements," NUREG-0737, November 1980.
NUREG-0800 U.S. Nuclear Regulatory Commission, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants," NUREG-0800, July 1981.
NUREG-0844 U.S. Nuclear Regulatory Commission, "NRC Integrated Program for the Resolution of Unresolved Safety Issues A-3, A-4, and A-5 Regarding Steam Generator Tube Integrity," draft report for comment, April 1985.
NUREG-1154 U.S. Nuclear Regulatory Commission, "Loss of Main and Auxiliary Feedwater Events at the Davis-Besse Plant on June 9, 1985," NUREG-1154, July 1985.
NUREG-1231, Supp. I 1
Appendix A
NUREG-1177 U.S. Nuclear Regulatory Commission, "Safety Evaluation Report Related to the Restart of the Davis-Besse Nuclear Power Station, Unit 1, Following the Event of June 9, 1985," NUREG-1177, June 1986.
NUREG-1195 U.S. Nuclear Regulatory Commission, "Loss of Integrated Control System Power and Overcooling Transient at Rancho Seco on December 25,1985," NUREG-1195, February 1986.
TAP Report Babcock & Wilcox Owners Group, Transient Assessment Report No. ANO-1 85-06 (TAP) for Arkansas Nuclear One, Unit 1, "Reactor Trip on MFW ARTS Trip," January 23, 1986.
l NUREG-1231, Supp. 1 2
Appendix A
APPENDIX B ABBREVIATIONS ABT automatic bus transfer ACRS Advisory Committee on Reactor Safeguards AFW auxiliary feedwater ANO-1 Arkansas Nuclear One, Unit 1 AP&L Arkansas Power & Light Co.
i ARTS anticipatory reactor trip system ASME American Society of Mechanical Engineers ATOG Abnormal Transient Operator Guideline ATWS anticipated transient without scram B&W Babcock & Wilcox B0P balance of plant Btu British thermal unit BWOG Babcock & Wilcox Owners Group CE Combustion Engineering CR-3 Crystai River Nuclear Plant, Unit 3 CRD control rod drive CST condensate storage tanks D-B Davis-Besse Nuclear Power Station ECCS emergency core cooling system EFIC emergency feedwater initiation and control EFW emergency feedwater EHC electrohydraulic control E0P emergcncy operating procedure ESF engineered safety feature (s)
ESFAS engineered safety feature (s) actuation system FMEA failure modes and effects analysis (analyses)
F0GG feed only good generator FPC Florida Power Corporation FSAR Final Safety Analysis Report FW feedwater GE General Electric GI generic issue GL generic letter GPUN General Public Utility Nuclear Corp.
HPI high pressure injection HSPS heat sink protection system I
l NUREG-1231, Supp. 1 1
Appendix B
I&C instrumentation and control ICS integrated control system IE NRC Office of Inspection and Enforcement IEB IE Bulletin IEIN IE information notice IGSCC intergranular stress corrosion cracking KSS known safe state LER licensee event report MCR main control room MDFWP motor-driven feedwater pump MFW main feedwater MFWP main feedwater pump.
M0V motor-operated valve MPR MPR Associates MSIV main steam isolation valve MSSV main steam safety valve NNI non nuclear instrumentation (system)
NRC U.S. Nuclear Regulatory Commission NSAC Nuclear Safety Analysis Center ORNL Oak Ridge National Laboratory OTSG once-through steam generator PIC previously identified concerns PORV power-operated relief valve PSM power supply monitor PTS pressurized thermal shock PWR pressurized-water reactor RC reactor coolant RCIC reactor core isolation cooling RCP reactor coolant pump RCPPM reactor coolant pump power monitor RCS reactor coolant system RHR residual heat removal RPS reactor protection system RTB reactor trip breakers RTS recommendation tracking system (BWOG)
SAIC Science Applications International Corporation SASS smart analog signal selector r
i SER safety evaluation report SFAS safety features actuation system SFRCS steam and feedwater rupture control system SG steam generator i
SGTR steam generator tube rupture SMUD Sacramento Municipal Utility District NUREG-1231, Supp. 1 2
Appendix B
SPDS safety parameter display system SPIP Safety and Performance Improvement Program SPRIG Safety and Performance Recommendation Integration Group SRP Standard Review Plan STA shift technical advisor TAP Transient Assessment Program TED Toledo Edison TMI Three Mile Island Nuclear Station ULO unit load demand USI unresolved safety issue UT ultrasonic testing W
Westinghouse NUREG-1231, Supp. 1 3
Appendix B
APPENDIX C NRC STAFF CONTRIBUTORS AND CONSULTANTS This suppl.ement is a product of the NRC staff and its contractors.
The staff members listed below were principal contributors to this report.
A list of contractors follows the list of staff members.
NRC STAFF Sections Name Title, Organization Reviewed M. Goodman Engineering Psychologist, Human Factors 7
Assessment Branch R. Jones Section Chief, Reactor Systems Branch 3, 5.5.3, 10, 11.4, 12.5 j
R. Kendall Senior Electrical Engineer, Instrumentation and 6.1, 6.3.3 Control Systems Branch 1
W. LcFave Senior Reactor Engineer, Plant Systems Branch 10.3 B. Siegel Lead Project Manager, B&W Reassessment,
- i., 2,
Project Directorate IV 4.5.4, 11.5, 12.2, 12.3, 12.4 M. Rubin Technical Assistant, Assistant Director for
- 9. 6 Systems CONTRACTORS Name Company Sections Reviewed J. T. Beard J. T. Beard Inc.
6.1.2, 6.1.3, 6.1.4, 6.1.6, 6.1.10 R. Haroldson EG&G/INEL 6.1.8 R. McCormick EG&G/INEL 6.1.12 M. Spencer EG&G/INEL 6.1.5 M. Yost EG&G/INEL 6.3.3 M. Waterman EG&G (Rockville) 6.1.5 T. Leivo Parameter Inc 6.1. 7 4
J. DeBor SAIC 7
R. Liner SAIC 7
C. Tobert SAIC 7
NUREG-1231, Supp. 1 1
Appendix C i
- =
APPENDIX E PREVIOUSLY IDENTIFIED ICS/NNI CONCERNS
- Source Concern Document Guidance / Recommendation 1.
Adequacy of power SMUD report Fuse all loads powered from distribution system June 1978 NNI-X and NNI-Y buses and fault protection evaluate the possibility of using lower size fuses.
2.
Failure of controls that NUREG-0560 Re evaluate control systems and could initiate transient May 1979 their significance to safety, or inhibit mitigation of transient 3.
Steam generator overfill BAW-1564 Review for possible changes to and overcooling resulting Sec. 3.2.c enhance reliability and safety from ICS failure Aug 1979 by preventing or mitigating the consequences of an inadver-tently opened feedwater startup valve and by providing means to prevent or mitigate the consequences of an open tur-bine bypass valve.
4.
Loss of power to controls IE Bulletin Review buses supplying power and to control room 79-27 to I&C systems that could instrumentation Nov 1979 affect the ability to achieve cold shutdown; identify I&C loads and evaluate loss of power to these loads; describe l
proposed design modifications.
5.
Consequences of loss of NRC Generic Address the proposed corrective NNI power Letter actions identified following Har 1980 the Feb 26, 1980 loss of NNI power event at CR-3 in terms of applicability to other B&W plants and expand the IE Bulletin 79-27 review to include implications of the CR-3 event.
- These are concerns typically common to the more significant B&W plant i
transients involving the ICS/NNI systems.
NUREG-1231, Supp. 1 1
Appendix E
Source Concern Document Guidance / Recommendation 6.
Loss of power supply Industry Investigate the need for review of backup or bus transfer CR-3 event capabilities upon loss of NSAC-3/INP0-1 instrumentation and control Rec. III.A.1 power supplies resulting Mar 1980 from faults.
7.
Close electrical NSAC-3/INP0-1 Investigate and take corrective coupling Rec. III.A.2 action regarding coupling of Mar 1980 indication, control, and computer input signals.
8.
Spurious control system NSAC-3/INP0-1 Investigate and take actions Rec. III.A.4 corrective action regarding Mar 1980 control system sesceptibility to spurious / inappropriate actions based on incorrect information (caused by elec-trical faults).
9.
Electrical cross-NSAC-3/INP0-1 Investigate (including field contamination of Rec. III.A.5 tests) to identify problems instrument channels and Mar 1980 concerning cross-contamination also of controls of power supplies affecting redundant channels of instru-ment and control functions.
10.
Effects of loss of NSAC-3/INPO-1 Investigate and take correc-single bus Rec. III.A.7 tive action regarding assign-Mar 1980 ment of instruments to specific buses to ensure as much re-dundancy as possible.
j 11.
Impact of single NUREG-0667 power buses and signal paths bus failure Rec. 5.a for ICS and NNI should be May 1980 separated and channelized.
12.
Undesirable failure modes NUREG-0667 ICS should have provisions (of power source, signal Rec. 5.d for detecting gross failures source, and ICS itself)
May 1980 and taking appropriate causing substantial defensive action.
plant upsets 13.
Implementation of NUREG-0667 Prompt action should be existing recommendations Rec. 5.h taken on IE Bulletin 79-27.
Hay 1980 14.
Avoiding unnecessary IE Information Consider scheduling calibra-failure mechanisms Notice 84-80 tion of NNI and ICS de power Nov 1984 supply alarm and trip set points every refueling outage, i
NUREG-1231, Supp. 1 2
Appendix E
u i
Source I
Concern Document Guidance / Recommendation 15.
Availability of minimum IE Bulletin Ensure that indications for indicators for safe 79-05A the minimum set of plant shutdown Apr 1979 variables needed for safe shutdown will be available.
16.
Alarm on loss of I&C IE Bulletin Review buses supplying power bus power 79-27 to I&C systems that could Nov 1979 affect ability to reach cold shutdown; identify and review for each bus the alarm /
l indications provided in the l
control room to alert op' era-tors to the loss of power to the bus; describe any proposed design modifications.
17.
Loss of control room NRC Generic Address information available indications Letter to the operator, including how Mar 1980 t!,s opertor determines which information is relic 51e, and identify what information is needed to br'ng the plant to cold shutdown.
18.
Indications of NRC Orders Make changes in equipment lost control Apr 1980 and control systems to give functions clear indications of functions that are lost or unreliable.
19.
Numerous and NUREG-0667 Modifications should be made complicated operator Rec. 11 to reduce post-reactor trip action necessary May 1980 immediate manual actions.
20.
Confusing or NUREG-0667 Power supply arrangements misleading
~ Rec. 5.b should be reconsidered to indications to May 1980 eliminate "mid-scale" failure operator mode for instrumentation.
21.
Backup NUREG-0667 Multiple instrument failures instrumentation Rec. 5.c should be unambiguously May 1980 indicated to guide operator selection and use of alternate instrumentation unaffected by the failure (s).
22.
Adequacy of NUREG-0667 Establish the minimum set of remaining or backup Rec. 6 parameters needed (recommended instrumentation May 1980 list include); instrumenta-tion for these should be reliable, redundant, and meet i
NUREG-1231, Supp. 1 3
Appendix E
l Source Concern Document Guidance / Recommendation safety system standards; at least one channel for each parameter shall be recorded for trending etc.
P 23.
Procedures needed for IE Bulletin Ensure adequate emergency loss of power to I&C 79-27 procedures exist to achieve bus Nov 1979 cold shutdown upon loss of power to each bus supplying power to I&C systems; include:
(a) alarms; and indicators of the loss, (b) use of alternate instruments / controls powered by other buses, (c) restora-tion of power to bus; describe any design modifications or administrative controls ne-cessary to implement such procedures.
24.
Implementation NUREG-0667 Develop and promptly of existing Rec. 14 implement plant-specific recommendations May 1980 emergency procedures for loss of NNI or ICS power; B&W development of AT0G endorsed; full utility support for AT0G recommended.
25.
Adequacy of safe SMUD Report Procedures should be developed shutdown capability June 1978 for safe shutdown upon total loss of NNI-X or NNI-Y power.
i l
NUREG-1231, Supp. 1 4
Appendix E 1
1 APPENDIX F NRC LETTERS TO THE BWOG CONCERNING THE ICS/NNI REASSESSMENT 1.
Letter dated June 24, 1986, from D. M. Crutchfield (NRC) to C. Doyel (BWOG I&C Committee).
Tnis letter provides staff feedhack to the BWOG concerning the ICS/NNI reassessment portion of the SPIP based on information presented by the BWOG during meetings with the staff on April 29 and May 21, 1986.
4 2.
Letter dated August 1,1986, from D. Crutchfield (NRC) to H. B. Tucker (BWOG Executive Committee).
This letter requested clarification of the BWOG reassessment program with regard to retrospective issues raised by the Incident Invcstigation Team in its report on the loss of ICS power event at Rancho Seco on December 26, 1985 (NUREG-1195).
3.
Letter dated September 12, 1986, from D. M. Crutchfield (NRC) to C. Doyel (BWOG I&C Committee).
This letter provided staff feedback to the BWOG concerning the ICS/NNI reassessment based on information provided in preliminary documents developed by the I&C Committee and discussed during a meeting with the I&C Committee on August 26, 1986.
4.
Letter dated December 17, 1986, from D. M. Crutchfield (NRC) to H. B. Tucker i
(BWOG).
This letter provided a list of concerns identified during the staff's review of the BWOG SPIP, as documented in BAW-1919 and its revi-4 sions, and requested that the BWOG address these concerns in a future update of BAW-1919.
5.
Letter dated February 13, 1987, from D. M. Crutchfield (NRC) to L. Stalter (BWOG I&C Committee).
This letter provided staff feedback to the BWOG concerning the ICS/NNI reassessment based on information provided in revised documents developed by the I&C Committee and discussed with the I&C Com-mittee during a meeting on December 10, 1986.
6.
Letter dated April 7,1987, from D. M. Cruchfield (NRC) to G. R. Skillman (BWOG).
This letter provided staff comments and questions concerning the BWOG ICS/NNI reassessment and requested that the BWOG address the associated concerns in a future update of BAW-1919.
7.
Letter dated November 24, 1987, from J. A. Calvo (NRC) to R. P. Rogers (BWOG).
This letter requested additional information needed for the staff to complete its review of Appendix R, "ICS/NNI Evaluation Final Report," to BAW-1919, "B&W Owners Group Safety and Performance Improvement Program (SPIP)."
1 NUREG-1231, Supp. 1 1
Appendix F
1 APPENDIX G ERRATA The following errata apply to the "Safety Evaluation Report Related to Babcock
& Wilcox Owners Groep Plant Reassessment Program."
Page 9-43, Table 9.22 Change first column:
"C" to "Cs" "T" to "T "
4 "D" to "D "
2 "T" to "T "
2 "C" to "Ca" "T" to "T "
i "D" to "D "
3 "C" to "C "
2 "D" to "C "
i "C" to "C "
3 "T" to "T "
3 "D" to "D "
3 Change Davis-Besse Score "39" to "9" i
NUREG-1231, Supp. I 1
Appendix G
geo.===
v s ve6.. asevs.toa, :
rr
....ot v., w.,m.,on v.,
.,.+
0'i,g, BIBUOGRAPHIC DATA SHEET NUREG-1231 us i navet.o. o%, i ni=u Supplement No. 1
,, a ti. o.., a ti
,6i.
i.6.==
Safety Evaluation Report related to Babcock & Wilcox Owners Group Plant Reassessment Program
.o..
j 5.wt-oa'n' March 1988
. o.u..o.i.mio Byron L. Siegel and others
.,o.a.
g March 1988
>.... o....a o. s.. ;,. i.o.... e...u s s 2 o. a,... i. c,
e wecen.
.oa v o =v=ei.
Office of Nuclear Reactor Regulation
.,,,o.....,.v U.S. Nuclear Regulatory Commission Washington, Le 20555
.,o.o.. s o. s. <,., o... 2... u o.c o. u,,.. <. <.,,,
ii.v...o....o.i Same as 7, above.
,,,,,o o co...,,_.,,,,,
,,......i,... osi.
,4..u.. c,.m m
Supplement No. 1 to the Safety Evaluation Report (SER) related to the Babcock & Wilcox Owners Group (BWOG) Plant Reassessment Program has been prepared by the Office of Nuclear Reactor Regulation of the U.S. Nuclear Regulatory Commission. This supplement contains the NRC staff's evaluation of the BW0G reassessment of the integrated control system /non-nuclear instrumentation system, the emergency feedwater initiation and control system, reactor trip initiating events, several additional open items identified in the SER, and the BWOG comments on the SER.
8 %OLlO800F/
.. o _ i,,................,.......
safety evaluation report (SER)
Babcock & Wilcox Owners Group (BWOG)
B&W-designed plant reassessment Unlimited
'4 SiCV.ity Ct.tilpac.teog
,Ta.. apes
. ei o...
. o.i % *ei o """'
Unclauified
, r..
Unclassified o -ei a o....u
,....c a
-