ML19269D918
| ML19269D918 | |
| Person / Time | |
|---|---|
| Site: | Crane, Black Fox  |
| Issue date: | 06/15/1979 |
| From: | Fate M PUBLIC SERVICE CO. OF OKLAHOMA |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| 6212DIN8-016-31, 6212DIN8-16-31, NUDOCS 7906200333 | |
| Download: ML19269D918 (80) | |
Text
6212 DIN 8-016-317 Fils 26212-125-3500-21L PUBUC SERVICE COMPANY OF OKLAHOMA A CENTRAL AND SOUTH WEST COMPANY P o Box 201/ TULSA. OKLAHOMA 74102 / (918) $83-3611 Wartier L fate, Jr.
becutive Vice Msdent June 15, 1979 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Re:
Public Service Company of Oklahoma Black Fox Station Response to TMI Event USNRC Docket STN 50-556 STN 50-557
Dear Mr. Denton:
As I discussed with you on June 13, Public Service Company of Oklahoma has already taken positive steps to study the significance of the Three Mile Island (TMI) event to our Black Fox Station project.
As the corporate ex-ecutive directly responsible for our nuclear project, I feel a deep sensitivity and need to respond to the events at TMI.
Therefore, we have proacted and directed that an assessment be conducted of the TMI accident in connection with PSO's pending application for construction permits for Black Fox.
The purpose of the assessment was to determine if any lessons learned from the TMI incident might be applicable to the design, construction, staffing, training and operation of the Black Fox Station.
We particularly wanted to as-certain whether any aspect of the accident had general ap-plicability to boiling water reactors.
The performance of this assessment is consistent with the suggestion of the ACRS in their interim report #3 dated May 16, 1979.
The ACRS noted that new mechanisms should be sought and imple-mented to expedite progress in addressing the new questions which have arisen because of Three Mile Island.
- Further, they suggested that where appropriate, licensees should perform suitable studies on a timely basis.
This would in-clude an evaluation of the pros and cons, and proposals for D\\
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Mr. Harold R. Denton June 5, 1979 possible implementation of safety improvements.
The Black Fox assessment is responsive to these objectives.
PSO has also established an in-house Technical Advisory Committee (TAC) whose function is to maintain continuous monitoring of developments stemming from the TMI investi-gations, and to assure that the " lessons learned" will be implemented during design construction and operation of the Black Fox Station.
A consultant, S. Levy, Inc., who is a participant in both the post event safe shutdown activities of TMI and in the EPRI investigation of TMI, has been retained by PSO to aid the workings of the PSO/ TAC.
Direct communica-tion has been established with EPRI and we will remain abreast of work performed by the Nuclear Safety Analysis Center, which was recently formed to investigate the TMI event at the request of the electric utility industry.
The assessment has been completed, and PSO has developed the attached program which will be implemented with respect to the Black Fox Station.
The details of the PSO program are described in the enclosure to this letter, and as a consequence, only the significant aspects will be provided here.
The program adopted by PSO was first reviewed to de-termine which activity, if any, would be foreclosed by the issuance of construction permits and the commencement of construction on safety-related systems at the Black Fox Station.
The small break analysis discussed in Plant Evaluation Issue 1, the failure mode and effect analysis for control systems discussed in Plant Evaluation Issue 3, the design review of RHR system performance discussed in Plant Evaluation Issue 8, and the control layout review discussed in Plant Evaluation Issue 15, are the only ac-tivities identified in the program that one could conceivably consider as having pre-construction permit design implica-tions.
Based on the TAC's assessment of these Issues, we conclude that none of these Issues involve activities that would be foreclosed by the issuance of construction per-mits for the Black Fox Station.
As the ACRS letter of April 7 clearly indicates, the recommendation for additional small break analyses (Issue 1) is prompted by the need to develop more knowledge and in-formation for reactor operators rather than a concern about the adequacy of reactor design.
Information of this type is best developed during the post-construction permit state, when the Black Fox design will be sufficiently complete that the sensitivity of reactor systems to various 2254 166
Mr. Harold R. Denton June 15, 1979 transients can be determined.
Consecuently, transient analyaes are properly left for the post-construction permit review state.
The same rational basis applies to the failure mode and effect analysis for control systems evalua-tion (Issue 3).
The RHR design review (Issue 8) involves an assessment of the adequacy of shielding and the ability to take coolant samples.
No pre-construction permit-type design changes are expected.
As indicated in the discussion of the control room layout review (Issue 15), the design of the NUCLENET control room concept is sufficiently flexible to accommodate any final design revisions at the final de-sign stage of the Black Fox Station.
A number of activities have been identified for action during the post-construction permit stage.
In addition to the foregoing, PSO intends to perform such tasks as:
1.
study adequacy and availability of reactor status information -- includinej core coolant inventory monitoring and operability of engineered safety features; 2.
provide assurance of containment isolation adequacy; and 3.
identify systems capable of transferring potentially radioactive fluids outside con-tainment and assure isolation and interlock when necessary.
PSO also intends to take a number of actions as a part of the operating license application for the Black Fox Station.
Examples include:
1.
full use of the BWR-6 training simulator in accident and abnormal event training; 2.
thorough description of the station's performance capabilities and limits in the operating procedures; and 3.
close coordination of the development of the Emergency Response Plan with local, state and federal jurisdictions.
4.
full delineation of qualifications, responsi-bilities, and authority of operations personnel in assuring safe plant operation at all times.
PSO executive management is committed to project in-volvement at every stage, from project conception through 2254 167
Mr. Harold R. Denton June 15, 1979 licensing and design into construction and operation.
We realize full well the personnel development responsibility that we bear in constructing and operating our first nuclear fueled generating station.
We have already amply demon-strated our desire to provide the best tools and education possible to our reactor operators.
The company made the decision several years ago to procure at a substantial increase in cost, a state-of-the-art control system which would give the operator correct, current, and understandable information regarding principal plant systems.
The NUCLENET control panel combines the most current data reduction techniques with space age man / machine interfaces to con-cisely display plant systems information.
As an adjunct, in order to superbly train and requalify these individuals, we successfully negotiated with General Electric to locate their BWR/6 training simulator near the BFS site.
Less than a mile from the station complex, the simulator will provide a unique opportunity to train opera-tors not only in normal plant operations but also to de-cisively respond to multiple failures and safety systems malfunctions incorporating the lessens of TMI.
The rigor-ous simulation permits repeatable real to life situations, to be displayed on control panels also identical to those at Black Fox.
Finally, we believe that our early establishment of an operating organization last year to plan station start-up and to fully participate in the development of station procedures assures an accurate translation of the AE/ vendor design into good operating procedures.
Through the review efforts of the TAC, this operating organization will under-stand and implement the lessons learned from TMI.
We believe the atta hed program constitutes a respon-sible assessment of the inplications of the Three Mile Island accident to the Bl ack Fox application.
My staff and I stand ready to sit down and discuss our program with you in order to expedite your understanding and concurrence.
On the basis of our establishment of and commitmenc to the program, we request that you remove the moratorium that has been placed on this project by the NRC Staff and support the issuance of construction permits for Black Fox Station.
Sincerely I
ccs:
BFS Service List
(
2254 168
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BLACK FOX STATION SERVICE LIST CERTIFICATE OF SERVICE I hereby certify that a copy of the foregoing PSO Response to the TMI Event has been served on each of the following persons by deposit in the United States mail, first-class postage prepaid, this 18th day of June, 1979.
L. Dow Davis, Esquire Mr. Joseph Gallo Counsel for NRC Staff Isham, Lincoln & Beale U. S. Nuclear Regulatory Commission 1050 17th Street N. W.
Washington, D. C.
20555 Washington, D. C.
20036 Dr. Cecil O. Thomas Joseph R. Farris, Esquire U. S. Nuclear Regulatory Commission Green, Feldman, Hall & Woodard Phillips Building 816 Enterprise Building 7920 Norfolk Avenue Tulsa, Oklahoma 74103 Bethesda, Maryland 20014 Docieting and Service Section Andrew T. Dalton, Esquire Office of the Secretary of the Comn.
1437 South Main Street, Suite 302 U. S. Nuclear Regulatory Commissicn Tulsa, Oklahoma 74119 Washington, D. C.
20555 (20 copies)
Mr. William G. Hubacek Mrs. Ilene H. Younghein U. S. Nuclear Regulatory Commission 3900 Cashion Place Office of Inspection and Enforcement Oklahoma City, Oklahoma 73112 Region IV 611 Ryan Plaza Drive, Suite 1000 Arlington, Texas 76012 Mr. Gerald F. Diddle Mr. Lawrence Burrell General Manager Route 1, Box 197 Associated Electric Cooperative, Inc.
Fairview, Oklahoma 73737 P. O. Box 754 Springfield, Missouri 65801 2254 169 Mr. Maynard Human Mrs. Carrie Dickerson General Manager Citizens' Action for Safe Nastern Farmers Electric CC?p.r?,:ive Energy, Inc.
P. O. Box 429 P. O. Box 924 Anadarko, Oklahoma 73bv5 Claremore, Oklahoma 74017 Michael I. Miller, Esq.
Charles S. Rogers, Esq.
Isham, Lincoln & Beale Assistant Attorney General Onc lst National Plaza 112 State Capitol Building Suite 4200 Oklahemg C ty Ok homa 73105 Chicago, Illinois 60603 Q
VaugNL.Conrad Manager, Licensing & Compliance Public Service Company of Oklahoma
TMI Assessment Program Black Fox Station, Units 1 and 2 Public Service Company of Oklahoma USNRC Docket STN 50-556 and 50-557 Public Service Company of Oklahoma has established a Technical Advisory Committee (TAC) to assess the events at Three Mile Island, Unit 2 and apply the lessons learned to its Black Fox Station project.
This committee has established at the direction of the President and Chief Executive Officer of the Company and reports findings and recommendations, directly to the Review and Audit Committee for Black Fox Station.
These finding and recommendations will then be implemented by the Review and Audit Committee.
The TAC is chaired by the Manager, Black Fox Station, who is in charge of all station cperations.
Other present members include the Manager, BFS Engineering; Manager, Nuclear Training; Manager, Nuclear Fuels, and the Manager, Licensing and Compliance.
The TAC has been directed to utilize PSO and consultant resources to fully review the interim and final results of the various investigations.
These p asently include:
USNRC's " Lessons Learned Task Force" The President's Commission on Three Mile Island EPRI - Nuclear Safety Analysis Center Generic vendor programs Atomic Industrial Forum TMI Policy Committee Rogovin.u"ritigation
.- The TAC and its consultants have already assessed issuances of the ACRS and regulatory staff.
It is aware of the activities of various other legislative and regulatory investigations and will assess future recommendations from them.
The assessment and resulting program was predicated on the advice and guidance set forth in the various letters from the ACRS (particulu."'t their letters of April 7 and May 16, 1979), and IE Bt letin No. 79-08, dated April 14, 1979.
In addition, S. Le ry, Inc., a participant in both the post event safe shutdown activities of TMI and the EPRI investigation, has been retained to keep PSO continuously informed of any new developments arising from the ongoing investigations by EPRI and other organizations.
The objective of the TAC and its consultants is to ensure that the Black Fox Station design, construction, operat-ing procedures, staffing and training program, and emergency response plan incorporates the benefits of the TMI investiga-tions to the fullest extent practicable.
As the first task-the TAC created a TMI Assessment Matrix for Black Fox Station, which is attached hereto.
This matrix was developed from all known regulatory guidance in an attempt to specify those issues that had been identified from the TMI experience which could possibly affect the Black Fox Station.
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.- After the issues were identified, an assessment was made of their import and applicability to Black Fox.
At that point a commitment to action within a specified time frame was made as appropriate.
The TAC has made a finding that no actions identi-fied in the program will be foreclosed by the onset of safety-related construction at Black Fox.
We believe that our initial assessment in review-ing the implications on our own operation illustrates our understanding of the salient concerns arising from the Three Mile Island incident.
Moreover we have demonstrably committed ourselves to an ongoing evaluation program for the added protection of the public health and safety during the operation of Black Fox Station.
2254 172
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'IMI Assessment M trix Ebr Black Ebx Station, thits 1 and 2 Issue Source Inpact EctionBy:
Action Timing PIAtIP EVALUATICri 1.
Evaluate anticipated IUREG-0560 All IWR's
- Li nsee/
Transient analysis with
function of equipnent shutdown including snall l
malfunction and/or break IOCA's and loss of htsnan error.
offaite and onsite 1C power.
2.
Identify safety aspects IUREG-0560 B&W B&W (bvered under iten 1 for i
of OPSG that deal with IMR generic analysis.
transients and anall breaks.
Examine EW transients 10 REG-0560 IHR's Vendors mitigation, and reli-ability of all IW systems.
i 3.
Evaluate control systens tUREG-0560 All IWR's Vendors ENEA's for control sys-Prior to Final and impact on plant safety tems that inpact plant Design via ENEA.
safety during transients.
Cbnsider design criteria for non-class IE systems inportant to safety.
4.
Evaluate monitoring sys-IUREG-0560 All IWR's Licensee /
Identify available Prior to Final tens for adequacy of ves-ACRS Vendor instrumentation, its
~ Design sel coolant inventory IE79-G3 reliability and redun-detennination.
dancy, and assure direct N
indication with back-up N
confirmation.
Ln 5.
PORV operability and IUREG-0560 All INR's Licensee /
Examine reliability of Prior to Final verification.
Vendors FORV's and upgrade where Design N
necessary; up3rade u
position detection capability.
The division of work between Licensee and Vender has not yet been defined.
Final design is understood to be the time of FSAR docketing.
~
'IMI Assessment Matrix Ebr Blac'c Ebx Station, Units 1 and 2 (cont'd)
Issue Source Inpact Action By:
Action Timing 6.
Containme:.t Isolation in NUREG-0560 All IWR's Licensee /
Strengthen procedures for During writing an accident.
IE79-08 Vendors manual containnent isola-procedures and tion valves. Evaluate prior to final initiation for validity design and completeness and reliability to assure isolation where needed.
7.
Potential transfer of NUREG-0560 All IWR's Licensee /
Identify systems Prior to final radioactive fluids IE79-08 Vendor capable of transferring design j
outside containment.
potentially radioactive l
fluids outside contain-ment and assure isolation and interlock when neces-sary.
8.
IUIR systen performance NUREG-0560 All IWR's Licensee /
Review design adequacy Prior to final IE79-08 Vendor with contaminated coolant, design review necessary actions-auto and manual in assure appropriateness.
9.
Natural circulation ACRS PWR's Licensee /
Detailed analysis and inplementation.
Vendor procedures.
10.
Pressurizer heaters.
ACRS IMR's Licensee /
Put pressurizer heaters Vendor on qualified onsite power supply and arrange for N
redundant capability.
N Ln
- 11. Post-Accident >bnitoring. ACRS IWR's Licensee Review reg. guide 1.97 re-Generic and A
quirements and assess ade-prior to final quacy of range of nonitor-design ing and ability to pull y
sanples to determine plant 4
condition.
'IMI Assesstrent Matrix Ebr Black Fox Station, Units 1 and 2 (conhi )
d Issue Source Inpact Action By:
Action Timing
- 12. 11 Ceneration IE79-08 IHR's ' '
Licensee /
Ibview equipnent and pro-Prior to final 2
Vendor cedure for dealing with design large anounts of hydrogen gas inside primary con-talment.
- 13. Power supplies ACRS IHR's Licensee /
Ib-examine adequacy of Prior to final Vendor offsite and onsite AC &
design DC power supplies for support of almormal transients and small break IOCA's.
- 14. Engineered Safety ACRS IWR's Licensee /
(bnsider adequacy of moni-Prior to final Features Systms Status Vendor toring of status of ESF design Fbnitoring.
systems inportant to safety to assure aware-ness and availability of systems to operator.
- 15. I!uman Factors in control NUREG-0560 IWR's Licensee /
Review control rocxn lay-Prior to final rocxn layout.
Vendor out and design to provide design operator response improve-ment and operator focus on core cooling needs with ability for easy instrument verification.
16.
Remote control venting ACRS IWR's Licensee /
Assure capability for re-Prior to final Vendor actor control venting of design N
non-condensible gases in N
RPV and primary system.
LJ1 A
- 17. Cbntainment purging TCRS IWR's Industry Perform design study of Within the next pros and cons of filtered 12 months
[
venting or purging of con-tainment in event of g
accident.
'IMI Assessment Matrix For Black Fox Station, Units 1 and 2 (cont'd) 8 Issue Source Inpact Accion By:
Action Timing
- 18. Core thernoccuples 1CRS IWR's Industry Cbnsider whether the core Prior to final exit tmperature measure-design ments might be utilized.
- 19. Decontamination and ACRS IWR's Industry Optimally plan the decon-Prior to final Recovery tamination and recovery design of major nuclear power plant systans.
I i
i N
N LD 4
Ch
'IMI Assessaant Matrix Ebr Black Ebx Station, Units 1 and 2 (cont'd) 4 Issue Source Inpact Act' ion By:
Action Timing OPERATIO2S 1.
Cbre cooling priority PURDG-0560 IWR's Licensee Establish through train-Prior to oper-ing, procedure, testing ator license that operator places first priority on naintaining core cooling. Consider real time operator feed-back on status of coolant inventory.
2.
Simulator Training NUREG-0560 INR's Licensee /
Add operator training on During opera-ACRS training sinulator dealing with tor training.
supplier human error and nultiple failure and safety system malfunctions.
3.
Senior Operator FUREG-0560 IHR's Licensee /
Upgrade senior operator's During opera-qualification training selection, qualification tor selection supplier and ability to direct and training emergency operations.
for senior
'Ihis may require college operator.
level technical education.
4.
Operator training and ACRS IWR's Licensee Examine operator qualifi-During opera-qualification and NBC cation, training and tor training.
licensing to inprove abil-ities; effectiveness of requalification training N
and assinulation of criti-N cal information.
Ln b
N
'IMI Assess:nent Matrix For Black Box Station, Units 1 and 2 (cont'd)
Issue Source Inpact Action By:
Action Timing 5.
Operating Proentres tamEG-0560 IWR's Licensee Increase refresher train-During develop-PERS and IRC ing on energency proce-ment of proce-dures including technical dures, staff. Review energency transient procedures by interdisciplinary teams of operation and systens ex-perts. Standardize among utilities. Write eraergency procedures in real time and and not just as a func-tional checklist. Address sinple and nultiple failure degraded conditions.
6.
Bnergency Planning ICRS IWR's Licensee Ebrmalize emergency plan-Prior to oper-IE79-08 and 1HC ning including procedure, ation advisory assistance, equipnent availability, reporting procedures to NBC for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> notifica-tion. Include designa-tion of a single point
" clearing house" for in-formation and public statenents.
7.
Overriding ESF systans IE79-08 IWR's Licensee Review operating proce-During prepa-N operation dures and training in-ration of N
structions to avoid over-operationg riding needed ESF system procedure operation and assure con-sideration of all plant parameters before over-y riding autanatic actuation.
a3
$4I Assesment Ltrix For Black Fox Station, Units 1 and 2 (conEId}
I Issue Source Inpact Action By:
Action Timing 8.
Safety-related systcm IE79-08 IHR'n Licensee Review procedure and in-Prior to final aligrments ACRS strunentation i.e. CRP design and dar-NUREG-0560 display formats to confirm ing preparation valve position and align-of procedures ments of safety related systans to assure aware-ness by operator of systen status and assure return of all systems to operable aligment following system returned to service.
9.
Assessment of operating NUREG-0560 IRR's Licensee Pcview reporting and data Generic program experience ACRS and NRC assembly process to accu-mulate and assess operat-ing data significance.
Ebrmalize use of IER's in operator licensing and requalification.
- 10. Technical Specifications NUREG-0560 IWR's Licensee Surveillance and testing During prepara-ACRS and NRC requirements, to ensure tion of tech-operability and proper nical specifica-l aligment of safety sys-tions tens should be factored into technical specifica-tions. Reporting require-i ments for unplanned events that don't exceed technical specification limits slould y
be identified. Other N
changes frun 'IMI assessment Ln must be factored into tech-A nical specifications and proposed by Licensee.
Re-view technical specifica-N tions to avoid overly re-C ctrictive requirements that would inhibit operator im-provisation during degraded plant conditions.
'IMI Assessr.ent.ht-ix Ebr Bla&_ Ex Scation, thits 1 ard 2 (cont'd) 4 Issue Source Iapact Action By:
Action Timing
- 11. Incident Response Center NRC IWR's Licensee Establish an incident During final response center away frun design the main control room to serve as a strategic cen-ter from which plant status can be indepen-dently assessed, future actions can be planned, and off-site commnica-tions can be established.
l
~
i N
Cn 4
CD
'IMI Assessment Ebtrix Pbr Black Pax Station, Units 1 ard 2 (cont'd)
Issue Source Inpact Action Dy:
Action Timing LIO2JSI!K; AND REXIIIATICN 1.
Quantitative Safety (bals ACRS IWR's NBC IEC determine quantita-Unknown tive safety goals by which design changes can be evaluated.
2.
Reactor and Ebel Cycle ACRS IWR's NBC Develop capability in Unknown Chcznistry reactor and coolant chem-istry to evaluate and establish acceptability of results of unplanned
~
event handling.
3.
Adequacy of single ACRS IMR's NBC t@C examine adequacy of Unknown failure criteria single failure criteria.
l 4.
Audit Calculation IUREG-0560 IMR's NBC NBC establish its.sn Unknown capability capability to perfonn quick engineering calcu-lations and perform audit calculations on IMR and IMR.
5.
BWR Analysis Fbdels IUREG-0560 IHR's G
G slould review Unknown its codes to ensure capability to perform realistic full spectrum analysis and provide nodel verification to NBC.
N N.
Standard Review Plan tUREG-0560 IWR's NBC Review and revise SRP's Unknown W
dealing with transient and anall break IOCA analysis based on 'IMI experience.
m
'IMI Assess:nant Matrix For Black Fox Station, Units 1 and 2 (cont'd)
Issue Source Inpact Action By:
Action Timing 7.
Interpretation of NUREG-0560 IWR's NIC NIC should develop Regu-Unknown General Design Criteria latory Guides to give greater guidance on design l
requirments for transients j
and degraded plant condi-tion analysis.
8.
Safety Research ACRS IWR's NIC IEC should put high Unkrown priority on safety re-l search of plant behavior during anmalous tran-sients sinulating a wide range of postulated events and accidents, including low probability mechanical and electrical failures and htrnan errors. }bre exploratory research should be performed with freedom i
frun inmediate licensing requirements.
N N
LM 4
N
PLANT EVALUATION ISSUE 1 Evaluate anticipated transient events as a function of equipment malfunction and/or human error.
REFERENCE:
NUREG-0560, p.
8-2.
ACRS letters dated April 7, p.
1, and dated May 16 (interim report 3), p.
3.
DISCUSSION:
Current licensing transient analysis begins with ab-normal operational transients but does not consider poten-tial for multiple operator errors and does not extend to stable cold shutdown conditions.
It is desirable to con-sider a wide range of anomalous transients and degraded accident conditions so that methods of controlling or pre-venting such conditions might be identified and incorporated into operator training programs.
Results of these studies may identify specific instrumentation and priorities that the operator should observe to most effectively preserve core coolant inventory.
Special attention should be directed toward understanding the sensitivity of anticipated transients as a function of equipment malfunction or human error.
Feedwater transients with additional equipment failures or human error should be addressed specifically to assure that the plant operator is trained to observe and recognize through the available instru-mentation the ongoing status of the event and to take the appropriate mitigating steps on a timely basis.
As the transient is extended to stable cold shutdown conditions, the requirements for performance of all safe 2254 183
Plant Evaluation Issue 1 Page 2 shutdown systems will be identified and the sensitivity to performance of these systems can be determined.
Results of these studies can be used to develop real time procedures that the plant operator can be trained to follow in the event of a degraded plant condition.
Although the specific TMI-2 event does not represent a real threat to BWR system performances, expanding this study to additional events will provide assurance that adequate understanding of all potentially degraded plant conditions are well enough understood to develop a proper operator training program.
General Electric analysis of the loss of feedwater flow transient combined with a stuck open safety relief valve has shown that the critical limit of suppression pool tempera-ture is not exceeded with reasonable operator initiation of the residual heat removal system.
The statement of concern for the small break LOCA analysis does not apply to the BWR; however, additional studies will be performed to reconfirm this fact.
The normal sequence of events for all BWR small break LOCAs is to initiate automatic isolation, automatic depressurization and automatic water level maintenance.
There are no trans-ition zones as a function of small break size that result in possible confusion to the operator or lead to potential conflicting demands for operator and safety system 2254 184
Plant Evaluation Issue 1 Page 3 performances.
The severity of the small break accident decreases monotonically with decreasing break size.
PSO ASSESSMENT:
PSO believes, that in view of the TMI event, it is important in the final design stage to re-examine the transient analysis and small break LOCA analysis for in creased understanding of the consequences of operator and safety system actions.
Transient analysis and small break LOCA analysis with multiple equipment failures and operator errors will be completed during the final design stage.
These additional analyses will be selected following a full assess-ment of potential mitigating system failure modes and effects analysis the most significant multiple failure scenarios will form the basis for detailed transient investigations.
Sensi-tivity to operator action and availability of mitigating systems will be examined in the final design to assure that operator training and transient procedures adequately prepare the operator to take proper and timely actions to protect the reactor core in the event of degraded conditions.
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PLANT EVALUATION ISSUE 2 Identify safety aspects of once-through steam genera-tors that deal with transients and small breaks.
Examine feedwater transients mitigation, and reli-ability of all feedwater systems.
REFERENCE:
NUREG-0560, p.
8-3.
DISCUSSION:
This issue is not applicable to the Black Fox Station as its nuclear steam supply system is a boiling water re-actor.
Feed water analysis for BWRs is included in Plant Evaluation Issue 1.
PSO ASSESSMENT:
No action is required.
2254 186
PLANT EVALUATION ISSUE 3 Evaluate control systems and impact on plant safety via failure mode and effect analysis.
REFERENCE:
NUREG-0560, page 8-4.
DISCUSSION:
Plant control systems play an essential part in plant operations and the control of transient situations.
These control systems are designed to assure a high availability of plant operation and are reviewed for impact on the safe operation of the plant.
Failure of controls could initiate a transient or could inhibit the control of a transient.
Transients have been evaluated to determine the consequences of feedwater control system failure, pressure control system failure and recirculation flow control system failure.
To date, no additional specific failure mechanisms have been identified with sufficiently high frequency to represent a concern for plant safety.
Each of these systems has suf-ficient reliability that the need to combine control system failure with other equipment failures or transients has not been felt.
As a result of the TMI event in which inadequate reactor water level control was provided, it appears that a re-evaluation of control systems via failure mode and ef-fects analysis (FMEAs) should be completed. FMEAs will identify realistic plant interactions resulting from failures in non-safety systems, safety systems and operator actions.
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Plant Evaluation Issue 3 Page 2 This evaluation can provide more insight into the rate at which the critical plant safety systems are challenged and can lead to a better understanding of acceptable criteria for final design of control systems.
This assessment should also be applied to monitoring systems relied upon to measure water inventory in the re-actor vessel.
Alternative monitoring methods can be iden-tified so that the operator can be better prepared to deal with abnormal monitoring system performances.
For equipment and systems that are important to safe plant operation but not required to be Class IE, this effort would provide confidence that those systems have the neces-sary reliability to be expected to function during abnormal transients.
For limiting systems, operating procedures can be developed to assist the operator in taking innovative 7 tion to account for the failure of a particular control system.
PSO ASSESSMENT:
In view of the TMI event, the seriousness of operator interactions with control systems, and the necessity of the operator to systematically analyze plant conditions and parameters to take appropriate corrective action, PSO be-lieves that a careful examination of the plant control systems via FMEAs should be completed.
This evaluation 2254 188
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Plant Evaluation Issue 3 Page 3 will be completed during final design of the control systems and the results will be made available to prepare the operator to accommodate transients with potential control system malfunctions.
Because of the extent of past control system design evaluations and the transient evaluations performed with control system failures, it is felt that no substantial modifications to control and monitoring systems are re quired.
The examination of systems at the time of final design will provide confirmation and provide useful guidance in completing operator training and preparing transient operating procedures for the most likely control system failures with the greatest consequences, and any modifica-tions can be incorporated Et that time in BFS.
2254 189
PLANT EVALUATION ISSUE 4 Evaluate monitoring systems for adequacy of vessel coolant inventory determination.
REFERENCE:
NUREG-0560, page 8-2.
ACRS letters dated April 7, 1979, pages 1,2 and dated May 16, 1979 (interim report 3), page 1.
DISCUSSION:
An overriding priority must be ercabliched for plant conditions following a transient or accident so that, regard-less of other concerns, no actions on the part of the oper-ator or automatic systems are contrary to maintaining core cooling.
The BWR has direct water level detection capabil-ity with diverse and redundant monitoring.
Many other parameter it.dications are also available to independently determine status of core cooling.
Prior to the completion of operator training and preparation of plant operating procedures, PSO will assemble a complete accounting of reactor core coclant inventory monitoring information, including an assessment of the reliabili,-- of each piece of information.
This evaluation should provide complete iden-tification of available coolant inventory information and identify the desired priority of dependence (by the operator) on this information.
Having completed this study, operator guidance and training can be developed to make the operator a more effective recovery agent or incident / accident mitigator.
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Plant Evaluation Issue 4 Page 2 Experience with the BWR demonstrates that sufficient instrumentation exists for the operator to continuously monitor reactor vessel inventory.
Means by which the operator would best employ the multiplicity of information provided to him should be co'isidered to prevent inappro-priate actions and promote productive intervention.
PSO ASSESSMENT:
PSO understands the importance of giving the highest priority to maintaining core cooling.
In view of the TMI event and the operator actions taken which were detrimental to maintaining core cooling, a special study will be under-taken at the time of final design.
The study will identify the available instrumentation related to maintaining core cooling and assure sufficient reliability and redundancy so that the operator is fully capable of assessing coolant status under emergency conditions.
The results of this study will be factored into operator training and plant operating procedures to make the operator a more effective recovery ar 7t or incident / accident mitigator.
An element of this program will also include proper operator preparation for steps to be taken prior to over-riding automatic action of engineered safety features.
These steps will take advantage of redundant core coolant inventory status information in order to increase assurance 2254 191
Plant evaluation Issue 4 Page 3 that engineered safety features will only be overriden if continued operation will result in unsafe plant conditions.
This program will also identify available confirmatory indications so that operational decisions will not be based solely on a single plant parameter indication.
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PLANT EVALUATION ISSUE 5 Power operated relief valve operability and verifica-tion of position.
REFERENCE:
NUREG-0560, page 8-5.
DISCUSSION:
The combination of a stuck open relief valve and a transient changes the transient into a small LOCA accident.
Although the consequences of this event for a BWR are not severe, it is of value to limit the frequency and severity of those events.
The frequency of abnormal transient events combined with stuck open safety relief valves depends, of course, on the reliability of the safety relief valve.
An assessment of past experiences with safety relief valves should be completed to determine the expected frequency of failure to close following actuation.
The consequences of failure of the relief valve to re-close are felt most severely in rising suppression pool temperature.
The maximum pool temperatures can be effectively limited by early operator detection of the valve remaining open.
Direct and positive indication of pool temperature and valve vent pipe temperatures give positive means for the operator to detect the status of the valve.
If the frequency of valve failure to re-close is as-sessed to be sufficiently large, then operator training and procedures should be examined to assure timely operator action.
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Plant Evaluation Issue 5 Page 2 PSO ASSESSMENT:
Although a stuck-open relief valve does not represent a severe transient event for the BWR, in view of the TMI event PSO recognizes the importance of assuring proper operator action under all transient conditions.
For this reason, a relief valve failure assessment will be completed at the time of final design and operator training and pro-cedures will be reviewed to assure proper operator actions to limit the consequences of this transient.
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a PLANT EVALUATION ISSUE 6 Containment isolation in an accident situation.
REFERENCE:
NUREG-0560, pages 8-6, 8-7.
DISCUSSION:
In order to minimize direct radioactivity released to the environment, it is important to assure that the reactor containment is isolated soon after the occurrence of an accident involving potential fuel damage.
Containment isolation in BWRs is initiated concurrently with emergency core cooling system actuation.
Since the design basis for BWR operation assures that any event which potentially results in fuel damage also initiates the emergency core cooling system.
Prior to final design, a study should be completed to assess the impact on containment isolation system perform-ance of the resetting of engineered safety features actu-ation signals following an accident and to assure proper operator actions to maintain containment isolation when necessary.
The plant parameters to be monitored for the initiation of containment isolation have been selected for adequate protection.
Since a number of containment isolation valves are not used during normal operation, these valves are under administrative control.
The procedures applying to these isolation valves need to be reviewed to ensure 2254 195
Plant Evaluation Issue 6 Page 2 correct positioning of all manual and remote manual con-tainment isolation valves.
PSO ASSESSMENT:
In view of the TMI event in which automatic containment isolation was not initiated by safety injection actuation, PSO believes that it is important to review the position indication and automatic and administrative controls in containment isolation prior to final design.
This study will reflect the importance of assuring prompt and complete containment isolation in the event of significant radio-activity build-up in the reactor containment.
This study will review the administrative procedures for the position-ing of manual and remote manual containment isolation valves, the selection of plant parameters to be monitored for initiation of automatic and manual isolation, and the impact of resetting engineered safety features actuation signals has on containment isolation system performance.
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PLANT EVALUATION ISSUE 7 Potential trcnsfer of radioactive fluids outside containment.
REFERENCE:
NUREG-0560, page 8-7.
IE Bulletin 79-08, page 3.
DISCUSSION:
During normal operation, several systems transfer liquids and gases outside of containment.
In the event of an accident involving fuel damage and release of radio-activity to the reactor coolant, excessive amounts of radioactive material may be transferred outside the con-tainment inadvertently if provisions are not made to pro-cedurally or automatically isolate these transfer systems.
For BWRs of the Black Fox design, system designs and proce-dures will prevent the inadvertent release of radioactive effluents.
To verify that this is satisfactory, a study should be completed at the time of final design to identify all systems that are capable of transfering potentially radioactive liquids and gases out of the containment in order to assure that inadvertent transfer of fluids will not occur.
This study should confirm that sufficient interlocks exist to prevent automatic transfer from occuring when a high radiation level exists.
PSO ASSESSMENT:
In view of the inadvertent transfer of radioactive fluids out of the primary containment during the TMI event, 2254 197
Plant Evaluation Issue 7 Page 2 PSO believes it is important to verify that sufficient interlocks exist to prevent inadvertent transfer of radio-active fluids from its BWR.
This study will be completed during final design, and proper operator training and pro-cedures will be implemented to confirm that proper transfer system interlocks will function during periods of high radiation inside the containment.
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PLANT EVALUATION ISSUE 8 Residual heat removal system performance.
REFERENCE:
NUREG-0560, page 8-7.
IE Bulletin 79-08, page 2.
DISCUSSION:
Long term heat removal following reactor isolation is carried out using the Residual Heat Removal (RHR) system.
The BWR system is designed to be self-contained and self-actuated.
Short term high pressure residual heat removal is initiated automatically via the Reactor Core Isolation Cooling System based on reactor vessel low water level signals.
Prior to low water level, actuation of RCIC heat is removed from the vessel via steam discharge to the suppres-sion pool through the safety relief valves.
Backup to the RCIC is provided by the High Pressure Core Spray System (HPCS) or automatic depressurization and use of the Low Pressure Coolant Injection (LPCI).
This system continues the core cooling function until the long-term shutdown cooling systems are placed into service by the operator.
Because there exist multiple redundant paths for shutdown cooling, the operator has many options to choose from in order to minimize distribution of highly contaminated coolant in the case of an event resulting in significant core damage.
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l Plant Evaluation Issue 8 Page 2 Although the plant design basis assures an extremely low likelihood of core damage, an evaluation should be performed to identify the optimum shutdown cooling mode in this highly unlikely event.
PSO ASSESSMENT:
In view of the TMI event and the subsequent complica-tions of long-term heat removal by the RHR because of high levels of contamination in the reactor coolant system water, PSo believes that during final design an optimization study should be completed to establish a decision process to be used by the operator in selecting the optimum long-term shutdown cooling path.
This path would be incorporated into the operator training program and the transient operating procedures to take maximum advantage of the automatic, redundant, diverse systems available to maintain core cooling.
PSO will also perform a design review of the RHR systems to assure long-term operability using contaminated coolant.
It is not expected that this assessment will result in major design modifications but rather, additional shielding and sample points.
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PLANT EVALUATION ISSUE 9 Natural circulation implementation procedures.
REFERENCE:
ACRS letter dated May 16, 1979 (interim report 2),
pages 1, 2.
DISCUSSION:
This issue is not applicable to Black Fox Station as its nuclear steam supply system is a boiling water reactor.
PSO ASSESSMENT:
No action is required.
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PLANT EVALUATION ISSUE 10 Use of pressurizer heaters to maintain suitable over-pressure on the reactor coolant system.
REFERENCE:
ACRS letter dated May 16, 1979 (interim report 2),
page 2.
DISCUSSION:
This issue is not applicable to the Black Fox Station as its nuclear steam supply system is a boiling water re-actor.
PSO ASSESSMENT:
No action is required.
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i PLANT EVALUATION ISSUE 11 Post-accident monitoring instrumentation.
REFERENCE:
ACRS letter dated May 16, 1979 (interim report 2),
page 3.
DISCUSSION:
The ability to follow and predict the course of an ac-cident is essential for mitigating it and also for accurately and reliably predicting its potential off-site consequences.
The BWR incorporates redundant and diverse means for tran-sient or accident monitoring by the operator, and the majority of accident sequences require minimal operator action to achieve a safe shutdown condition.
Most instrumentation and control systems equipment is housed inside a Class I secondary containment building environment, providing added assurance that the instrumentation will be able to follow the course of an accident.
Instrumentation to follow the course of an accident is the subject of Regulatory Guide 1.97 and is the subject of an NRC Staff Task Action Plan for the resolution of generic issues.
PSO ASSESSMENT:
PSO believes the lessons learned from TMI should provide a basis for reviewing the positions of Regulatory Guide 1.97, including a potential redefinition of the Task 2254 203
J Plant Evaluation Issue 11 Page 2 Action Plan.
Although review and reexamination of existing criteria may take some time, the BWR with the NUCLENET con-trol room provides adequate tiexibility to accommodate reasonable resolution of Regulatory Guide 1.97 requirements during the course of final design.
During construction permit hearings for the Black Fox Station, the NRC Staff acknowledged that the plant design could be modified during construction to accommodate any additional instrumentation required by the Task Action Plan.
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PLANT EVALUATION ISSUE 12 Operating modes and procedures to deal with significant amounts of hydrogen gas that may be generated during an event.
REFERENCE:
IE Bulletin 79-08, page 3.
DISCUSSION:
During the unlikely event of severe loss of core cool-ing and elevated fuel cladding temperatures, significant amounts of hydrogen gas may be generated.
During normal operation, the BWR processes hydrogen through the main con-denser.
During isolation of the main condenser, hydrogen in the reactor pressure vessel is vented to the containment via several paths.
Inside the containment, the hydrogen is controlled, treated, and contained via several systems.
Following final plant design, operating procedures can be established to determine the best combination of automatic and manual treating and venting of hydrogen as a function of the severity of hydrogen generation.
PSO ASSESSMENT:
In view of the severity of fuel damage at TMI and the consequent build-up of hydrogen, PSO believes that during the course of final design, the best utilization of hydrogen venting and treatment systems should be studied so that operator training and procedures can be developed to deal with this unlikely hydrogen build-up condition.
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PLANT EVALUATION ISSUE 13 Reliability of electric power supplies, specifically off-site and on-site AC and DC power supplies.
REFERENCE:
ACRS letter dated May 16, 1979 (interim report 3),
page 3.
DISCUSSION:
The ACRS believes that it is important that a compre-hensive examination be made by the NRC technical staff and reactor licensees of the adequacy of design, testing and maintenance of off-site and on-site AC and DC power supplies.
It is also suggested that some type of failure modes and effects analyses be made.
Finally, it is recommended that consideration be given to systematic testing of power system reliability during anomalous system transients and to development of ways to improve the status monitoring and control of power systems.
Loss of off-site auxiliary power has long been a design basis transient which BWRs are designed to withstand without any fuel damage.
Because of the potential for loss of normal off-site power to the power plant, substantial redundance and diversity exists between off-site and on-site emergency AC and DC power supplies.
Many studies have been performed of the adequacy of design, testing and maintenance of off-site and on-site AC and DC power supplies.
Systematic testing, quality assurance requirements and status monitoring of power supply systems are imposed to assure high reliability 2254 206
Plant Evaluation Issue 13 Page 2 of power supplies required for safe reactor shutdown.
FMEAs have been performed as a guide to understand potential power supply failure modes.
PSO ASSESSMENT:
The TMI event illustrated dependence on availability of particular components needed for long-term cooling (this is addressed in our discussion of Plant Evaluation Issue 1).
PSO believes that as part of its efforts to assure long-term cooling capability that the reliability of standby power supplies should be re-evaluated to demonstrate the high reliability expected in the existing design.
The assessment of reliability of power supplies during final design will be used as a basis for establishing testing and surveillance intervals.
A considerable amount of detailed work has already been done, as documented in the Preliminary Safety Analysis Report, and the reliability of off-site sources and the stability of the electrical system was a subject of intense staff review.
In addition, this issue was a topic of the subcommittee and full committee ACRS meetings for Black Fox Station.
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PLANT EVALUATION ISSUE 14 Provisions for additional status monitoring of various engineered safety features and their supporting services.
REFERENCE:
ACRS letter dated May 16, 1979 (interim report 2),
page 4.
DISCUSSION:
Status monitoring of engineered safety features is critical to their achieving satisfactory performance.
The much more serious consequences of the unavailability of engineered safety features can be of concern if the status monitoring is inadequate.
Status monitoring, not dependent chiefly on administrative control and thus less subject to human error, might help increase the availability of essential safety features.
A study should be performed at the time of final design to examine the advantages and disadvantages of relying on status monitoring through administrative control.
PSO ASSESSMENT:
In view of the TMI e'ent, PSO believes that efforts need to be made during final design to verify the effective-ness of status monitoring of engineering safety features.
Although the closed auxiliary feedwater system valves may not have contributed directly or significantly to the core damage at TMI, the potential for errors in the status of safety systems has been highlighted.
The NUCLENET control 2254 208
Plant Evaluation Issue 14 Page 2 room feature at the Black Fox Station provides considerable status monitoring improvement over earlier BWR designs and provides the flexibility to accommodate additional automatic status monitoring if studies show this to be necessary.
Results of additional studies of the need for improved status monitoring and of the availability of this informa-tion to the operator can be factored into operator training
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and operating procedures.
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PLANT EVALUATION ISSUE 15 Human factors engineering; emphasis in the design and layout of control rooms.
REFERENCE:
NUREG-0560, page 8-12.
DISCUSSION:
The operator has been trained to rely on his instrumen-tation.
He will continue to do so until he suspects an erroneous reading; however, he must be trained not to rely solely on a single indication since it may be erroneous or misleading under certain conditions.
If the operator has too many additional manual func-tions to perform, he may reduce his observations on other system parameters, which may lead him to have " tunnel vision."
Human factors engineering should be emphasized in the design and layout of control rooms.
The location of in-struments and controls in many power plants may reduce the likelihood of operator error or, at least, help the opera-tor to efficiently carry out the normal, abnormal, and emergency actions required of him.
The NUCLENET control room design takes advantage of recent improvements in control room layout for improved operator action.
Studies completed during final design may indicate that operator and technical staff training should be revised to improve the operator's understanding 2254 210
f Plant Evaluation Issue 15 Page 2 of his responsibilities during abnormal and emergency con-ditions.
It is recognized that the operator has been trained to believe his instrumentation, but he must not do so blindly.
Almost every monitored parameter of interest can be validated by appropriate checking of other instru-mentation.
The operator must be trained to perform this crosscheck to verify instrument display and must not develop
" tunnel vision" in which one display is relied on exclu-sively.
PSO ASSESSMENT:
In view of the TMI event where it appears that the operator relied on the high pressurizer level and thus er-oneously inhibited core cooling, PSO believes that special studies should be completed during final design to assure that the control room is instrumented and laid out ade-quately.
Human factors should be examined to assure that the operator can rapidly verify that appropriate safety systems actuations have taken place, and that, if appro-priate actuation has not occurred, he is capable of in-terceding to perform necessary actions.
The NUCLENET control room concept is flexible enough to accommodate final design revisions coming out of these studies. In fact, it is unlikely that anything other than 2254 211
Plant Evaluation Issue 15 Page 3 software changes would be required.
These studies would also be used to assure proper operator training and tran-sient procedures preparation, including requirements for the operators verifying performance status with diverse in-strumentation.
Emphasis must continue to be placed on maintaining and verifying core cooling.
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PLANT EVALUATION ISSUE 16 Provision for remotely venting high points in the reactor system to eliminate non-condensible gases.
REFERENCE:
ACRS letter dated April 7, 1979.
DISCUSSION:
In the event of formation of non-condensible gases in the primary coolant system of the reactor, means need to be available to vent these gases from the primary system to assure the gases do not interfere with normal coolant flow pathways.
The BWR normally operates in the nonsolid, vented condition with steam and non-condensible gases vented through the steam line to the main condenser.
If the re-actor is isolated from the main condenser, the reactor pressure vessel dome can be vented to the suppression pool via the safety relief valves.
In addition, a remotely operated vent valve at the top of the reactor pressure vessel dome vents directly to the containment.
PSO ASSESSMENT:
In view of the TMI event, PSO understands the impor-tance cf having the ability to remotely vent gases from the primary system.
Because the BWR provides both automatic and remotely operated means of venting high points in the reactor coolant pressure boundary, this issue is not likely to have design implications for the Black Fox Station.
However, a design review will be performed to confirm this conclusion.
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PLANT EVALUATION ISSUE 17 Ability to vent or purge the containment through filters which could retain particulates and most iodine in the event involving degraded conditions.
REFERENCE:
ACRS letter dated May 16, 1979 (interim report 3) page 6.
DISCUSSION:
Filtered, vented containment provides additional pos-sible means to mitigate the consequences of serious reactor accidents.
This concept represents a serious departare from the current BWR design.
Although the controlled filtered venting of a containment which could retain particulates and the bulk of the iodine generated in a serious accident has been recognized for more than a decade, little progress has been made on the development of sufficiently detailed design information on which to determine the efficacy and the factors relevant to the possible implementation of such a design concept.
Although this approach is premature for application to BWRs currently being planned and constructed, a generic study might be sponsored by the NRC to determine the prac-ticality, pros and cons, the cost, and the potential for risk reduction of a filtered venting or purging containment design.
PSO ASSESSMENT:
The TMI event suggests that a filtered vented con-tainment could be useful in reducing the risks of a serious reactor accident, because of the uncertainties of this issue PSO believes that it is premature to consider this concept 2254 214
Plant Evaluation Issue 17 Page 2 for the Black Fox Station.
Adequate means for controlling radioactive releases are presently provided in the Black Fox Station design.
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s PLANT EVALUATION ISSUE 18 Core exit thermocouples for additional indication on the status of the core, employing full range indication.
REFERENCE:
ACRS letters dated May 16, 1979 (interim report 2) page 3 and dated April 17, page 1, 2.
DISCUSSION:
The PWR uses core exit temperature measurements to provide indication regarding the status of the reactor core and core cooling.
The BWR does not use thermocouples be-cause the BWR operates in a saturated steam environment
~
where the coolant temperature remains constant.
Thermal performance of the core is based on fluid energy balances and in-core neutron flux detectors.
In addition, direct water level instrumentation provides adequate status of core coolant inventory.
In view of the potential additional information that might be gained under accident conditions, the potential use of thermocouples should not be discarded without studying the pros and cons further prior to final design.
PSO ASSESSMENT:
In view of the TMI event and the importance of complete information on core fuel and coolant inventory status with diverse assessment, PSO believes that a study of the pros and cons of utilizing thermocouples in the Black Fox Station should be completed prior to final design.
If necessary, thermocouple instrumentation could be added durin$4 216 final 22 design.
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PLANT EVALUATION ISSUE 19 Design changes necessary to facilitate the de'.Jn-tamination and recovery of major nuclear power plant systems.
REFERENCE:
ACRS letter dated May 16, 1979 (interim report 3),
page 4.
DISCUSSION:
Following a serious accident with highly contaminated primary system coolant, substantial equipment decontamina-tion of the major power plant systems will be required be-fore recovery and return to power is possible.
Prior to final design, it is desirable to have completed an assess-ment of the survivability, behavior and failure modes of systems important to long-term recovery.
Results of this study will determine the optimum use of various systems during the long-term shutdown cooling mode following a serious accident such that return to full power operation can be effectively achieved.
Operating procedures can then be generated which recognize the sensitivity of those systems to the presence of contaminated fluids and the decontamina-tion required in order to maintain system reliability.
PSO ASSESSMENT:
PSO believes that r.uch can be learned from the TM1 event and post-event recovery process.
In view of the TMI event, PSO plans to maintain awareness of the progress of the decontamination and recovery process to factor any poten-tial improvements into the final systems design whereever 2254 217
s Plant Evaluation Issue 19 Page 2 possible.
This program is not essential to the public health and safety, but will serve to best prepare the plant to effectively recover from a serious accident.
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OPERATIONS ISSUE 1 Establish overriding priority for maintaining core cooling.
REFERENCE:
NUREG-0560, pages 8-8, 8-10.
DISCUSSION:
The most basic, fundamental aspect of nuclear safety and licensing philosophy has been protection of the core by providing adequate core cooling.
On the basis of the events at TMI, more effective translation of this message needs to be made to the reactor operators and station management.
PSO ASSESSMENT:
We believe that new emphasis in various aspects of exist-ing personnel selection and training methodologies, may be required to instill the fundamental nuclear safety philoso-phies into the reactor operator so he instinctively reacts correctly to an anomalous situation consistent with the integrated design.
In the development of our training program, we will look closely at the background and educa-t tien requirements to see if an upgrading would provide a positive, reliable measure of improvement.
Development of station operating procedures will be lirected toward the priority of core cooling.
Once we have selected the remainder of the station management and operator candidates, extensive use of the General Electric BWR/6 Training Simulator will provide a 2254 219
s Operations Issue 1 Page 2 means of giving these people experience in responding cor-rectly to anomalous events.
This simulator will be less than a mile from the Black Fox site and will be an exact replica of the Black Fox Station control room, and will regorously model the Black Fox units.
In addition, a pro-gram of periodic drills in "what if" situations is judged to be essential with the reactor operations at the actual control boards.
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t OPERATIONS,ISSU2 2 Simulator training is an essential tool to address human errors and plant operation during transient conditions.
REFERENCE:
NUREG-0560, page 8-9.
ACRS letter dated May 16, 1979 (interim report 3) page 2.
DISCUSSION:
Simulator training is seen as an essential element in assuring that station operations personnel are prepared to consistently respond in an appropriate way to plant con-ditions during transients.
It offers unique opportunities to create situations in real and accelerated time frames to assure complete understanding on the part of the trainee.
Retraining can be done at any time to address a particular situation and to research the most practical solution.
PSO ASSESSMENT:
We will reevaluate our previously planned training exercises using the General Electric BWR/6 simulator.
Several years ago, PSO top management illustrated their belief in the importance of simulator training to the safe operation of the Black Fox Station by successfully negoti-ating with General Electric to locate its BWR/ 6 training simulator near the Black Fox site.
This simulator, located less than a mile from the site, is a replica of the Black Fox Station control room.
Additionally, the system utilizes 2254 221
s Operations Issue 2 Page 2.
data display and interpretation for the operator.
This is designed to markedly reduce the incidence of human error by utilizing superior data acquisition, reduction and display techniques in coordination with enhanced man-machine interfaces.
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OPERATIONS ISSUE 3 Qualification and ability of senior reactor operators to direct activities during abnormal or emergency operations.
REFERENCE:
NUREG-0560, page 8-9.
DISCUSSION:
It is desirable to determine better ways of evaluating a senior operator's ability to direct activities during abnormal or emergency operations.
These ways include a reevaluation of personnel educational requirements as well as specialized training.
The specialized training must include a significant amount of simulator experience, both in normal requalification sequences as well as in an "as needed" basis.
PSC ASSESSMENT:
We are reevaluating our educational and experience re-quirements.
One way is by a careful study of the proposed Revision 2 to Regulatory Guide 1.8,
" Personnel Selection and Training which endorses ANSI /ANS 3.1-1978 - Selection and Training of Nuclear Power Plant Personnel".
The NRC Staff has recently requested more comments be submitted by the industry on the basis of considerations stemming from TMI.
Items being considered by NRC and evaluated by PSO include:
2254 223
e Operations Issue 3 Page 2 Staffing, training, initial qualification, and re-qualification of operating personnel; Staffing, training, initial qualification, and re-qualification of supervisory personnel; Staffing, training, initial qualification, and re-qualification of technical support personnel; Use of plant simulators for training, initial qualification, and requalification; Training of plant operating staff following extended shutdown; Content of programs for training nuclear power plant personnel; Use of operating experience information in training nuclear power plant personnel; We will continue to carefully evaluate these aspects of personnel training as well as others resulting from the various investigations.
We also recognize that the ability of senior personnel to direct activities in stress situations is demonstrated through selection on the basis of previous experience, training, eesting, job performance and evaluation as well as evaluation during periodic drills.
The senior personnel in charge in the control room must be dedicated to a command and control function with a minimal amount of administrative duties, so that he can always be able to assure that adequate core cooling is provided.
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9 OPERATIONS ISSUE 4 Examine operator qualifications, training and licensing to improve abilities and effectiveness, especially in stress conditions.
REFERENCE:
ACRS letter dated May 16, 1979 (interim report 3) pages 1, 2.
NUREG-0560, page 8-10.
DISCUSSION:
This issue is an adjunct to Operations Issue 3, con-cerning the abilities of the senior reactor operator.
PSO ASSESSMENT:
We believe that our assessment of Operations Issues 2 and 3 adequately addresses this subject and illustrates the PSO commitment to action.
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OPERATIONS ISSUE 5:
Analysis and development of operating procedures to promote proper understanding of event sequences, margins available to the operator, and critical decision points.
REFERENCE:
NUREG-0560, page 8-10.
DISCUSSION:
It is important that the station operations personnel and the'ir written procedures fully and honestly reflect the capabilities and design needs of the integrated plant systems under a variety of conditions.
The critical element is the transfer of knowledge from the designer / vendor to those responsible for operation of the facility with due regard for the public health and safety.
PSO ASSESSMENT:
We believe that our early establishment of an operating organization was essential to assure their complete understanding of station design.
These people are routinely involved in design operability review and equipment selection.
A full time, senior manager was assigned as -- The Manager, Nuclear Training over two years ago to ensure that a firm basis for the BFS training program would be set.
.The Manager, Black Fox Station has been a vital functioning member of the PSO staff since the Fall of 1977.
It is expected that many of the members of the BFS Engineering staff, who perform a third level review of the design, will be on the operations team.
A BFS mechanical the advanced NUCLENET control boards which incorporate a significant step forward in 2254 226
s Operations Issue 5 Page 1 engineer has recently been named a Test Engineer for Start-up.
Additionally, we currently employ an outside start-up consultant who has one man full time in our offices.
General Electric has a contract engineer who works for PSO in the start-up and operations area.
We have authorized the hiring of the Station Superintendent, Station Engineer, and Operations Supervisor, prior to the end of 1979.
The incorporation of these wide varieties of per-i sonnel background ensures an interdisciplinary review.
We recognize the need to close the cultural gap between plant designers and plant operators.
Both groups can benefit from training in the other's area to provide a basis for feedback and feedforward of fundamental philosophies.
This includes not only station operations personnel but also decision-makers on a corporate level.
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OPERATIONS ISSUE 6 Emergency planning reassessment to assure adequate off-site advice and assistance as well as prompt notification of appropriate regulatory authorities.
REFERENCES:
ACRS letter dated May 16, 1979, (interim report 3),
pages 2, 3.
IE Bulletin 79-08, page 3.
DISCUSSION:
The experiences at TMI graphically illustrate the need for prompt coordinated notification of the appropriate regulatory authorities in the event of a radiological emer-gency.
Each licensee must also have defined the inside and outside resources available to meet the needs of various levels of emergency situations.
PSO ASSESSMENT:
We are evaluating the current regulatory requirements for emergency planning colored by the events at TMI.
Since April 1, we have had several meetings with Oklahoma State Department of Health, Division of Occupational and Radio-logical Safety personnel who have been designated by the Governor, State of Oklahoma, as the prime respondent.
Our health physics, operations, and licensing personnel are now reviewing a draft State Emergency Response plan.
We are fully prepared to assist the State in timely development and submittal of the plan to the NRC for approval.
Con-currently, PSO is establishing target tasks for the BFS 2254 228
Operations Issue 6 Page 2 Emergency Response Plan development to be submitted in support of an application for operating licenses.
Reporting procedures will be established to ensure notification of NRC within one hour of an emergency event.
These procedures will also include designation of a single point of contact for information and public statements.
PSO has had a Manager, Nuclear Information for several months for this very function.
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t f
OPERATIONS ISSUE 7 Overriding of engineered safety feature operation.
REFERENCE:
IE Bulletin 79-08, page 1.
NUREG-0560, page 8-8.
DISCUSSION:
As previously discussed in Operations Issues 1, 2 and 3, the preservation of adequate core cooling is paramount and fundamental to nuclear safety.
An overriding priority for plant conditions to be pursued following a transient or accident, must be established so that, regardless of other concerns, no actions on the part of the operator or automatic systems should be contrary to maintaining core cooling.
Oper-ational personnel should be instructed to not override auto-matic action of engineered safety features unless continued operation of these engineered safety features will result in unsafe plant conditions.
PSO ASSESSMENT:
As stated in our assessment of Operations Issues 1, 2
and 3, new ways of selecting, educating and training opera-tions personnel are required.
The PSO plan of action has been delineated in those earlier analyses.
However there are times when it is appropriate to override automatic ESF operations.
An assessment will be done to establish proper abnormal and emergency procedures allowing necessary over-riding without endangering the adequacy of core cooling.
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OPERATIONS ISSUE 8 Safety related system alignments; status monitoring.
REFERENCE:
ACRS letter dated May 16, 1979 (interim report 2),
Page 4.
IE Bulletin 79-08, page 1.
NUREG-0560, page 8-16.
DISCUSSION:
Safety related systems are provided with redundant capacities so that portions of these systems may be main-tained during power operation.
Maintenance and test pro-cedures and technical specifications should be established to assure proper _ alignment of safety systems prior to re-moval from service.
Upon return to service folloving main-tenance or testing, procedures should require explicit notification of reactor operational personnel whenever a safety-related system is removed from and returned to service.
Surveillance procedures should be established to ensure that safety-related valves are returned to their correct positions following necessary manipulations and are maintained in their proper positions during all operating modes.
Procedures for maintenance and testing of safety-related systems should be required to ensure verification of operability of safety systems when they are returned to service.
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Operations Issue 8 Page 2 PSO ASSESSMENT:
In view of the improper alignment of safety-related systems at TMI, increased emphasis will be placed on re-viewing technical specifications, maintenance procedures and test procedures for safety-related systems to assure proper verification of responsibility, alignment and operator notification.
This effort will be undertaken during the preparation of those specifications and procedures.
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OPERATIONS ISSUE 9 Assessment of operating experience from similar plants, including Licensee Event Reports (LER).
REFERENCE:
ACRS letter dated May 16, 1979 (interim report 3),
page 2.
DISCUSSION:
The LERs can provide a significant source of early warning of potential plant equipment and system failure or malfunction.
The problem is one of establishing a method of constant update and easy data retrieval for design and operations assessment.
~ PSO ASSESSMENT:
We have previously committed (PSAR S 14.1. 6. 4) to use LER's in similar plants in the preparation of the BFS startup programs.
Additionally, we are presently investigating the feasibility of taking the LERs in magnetic tape form from the NRC and placing it on a digital computer.
We will utilize our already implemented ATMS/ STAIRS text management system to aid data retrieval and extend the usefulness of the data base.
PSO also participates in the Nuclear Plant Reliability Data (NPRD) effort which provides similar valuable indicators of failure trends.
We anticipate that similar useful information will be available for the EPRI-Nuclear Safety Analysis Center.
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OPERATIONS ISSUE 10 Technical specifications development.
REFERENCE:
ACRS letter dated May 16, 1979 (interim report 3),
page 2.
NUREG-0560, page 8-16.
DISCUSSION:
The concern is that the existing regulatory philosophy of reliance on technical specifications to assure safety may not adequately cover all potential abnormal events.
Specific concerns center on surveillance and testing requirements to ensure that operability and alignment of safety systems are factored into technical specifications.
Reporting require-ments for unplanned events that do not exceed technical specification limits should be identified.
Other changes from TMI assessment must be factored into technical specifica-tions and proposed by reactor licensees.
Technical specifications should be reviewed to avoid overly restrictive requirements that would inhibit advan-tageous operator improvisation during degraded plant conditions.
PSO ASSESSMENT:
This issue is closely allied with Operations Issue 8.
In addition to the action described therein, PSO is partici-pating in industry technical specifications groups seeking to standardize safety, environmental and radiological technical specifications.
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a OPERATIONS ISSUE 11 Establish an Incident Response Center away from the main control room as a strategic planning center for reaction to incidents such as the TMI event.
REFEFENCE:
NRC Staff presentation to ACRS and BWR owners on June 14.
DISCUSSION:
It is desirable to establish a command center away from the main control room which can be activated in the event of an unplanned anomalous event.
This center should contain telemetered information from the main control room to assess current plant status and analyze performance trends.
Detailed and up-to-date plant configuration draw-ings should be maintained in this center, including accurate layouts of all piping systems, drain lines, valves, pumps, connecting lines, sample locations, etc.
This center will relieve potential congestion in the main control room and avoid potential diversion of the plant operators.
In establishing the Incident Response Center, proce-dures should be established to identify specific technical personnel on call at all times to respond to emergency con-ditions.
Criteria for selection of on-call personnel should be established to identify necessary qualifications and training to adequately equip these personnel for response to anomalous transient and accident events.
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Operations Issue 11 Page 2 In addition to the first line emergency on-call per-sonnel, identification of specific technical reserve per-sonnel shall be established to provide additional expertise in ar_eas potentially necessary to cover a broad range of possible emergency situations.
PSO ASSESSMENT:
In the Black Fox Station Preliminary Safety Analysis Report, S 13.3.3, PSO identified the secondary Emergency Control Center located away from the generator complex within the site boundaries.
This center will serve as the relocation center for station personnel as well as the coordinating point for outside agencies involved.
The above discussion items will be incorporated into the Emergency Control Center.
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LICENSING AND REGULATION ISSUE 1 Quantitative safety goals.
REFERENCE:
ACRS letter dated May 16, 1979 (Report on Quantitative Safety Goals.)
DISCUSSION:
The Advisory Committee on Reactor Safeguards has recom-mended that consideration be given by the NRC to the es-tablishment of quantitative safety goals for overall safety of nuclear power reactors.
Recognizing the difficulty and uncertainties in the quantification of use and understand-ing that engineering judgment may be the only basis for a decision, the ACRS believes, nevertheless, that the existence of quantitative safety goals and criteria can provide important yardsticks for such judgment.
PSO ASSESSMENT:
We concur that it is time to place the discussion of risk, nuclear and non-nuclear, on as quantative a basis as feasible.
Management by objective is only possible when one knows the objective and has given thought to its worthi-ness.
Ground rules must be established beforehand that will allow us to determine, in agreement with the regulatory staff, when safety goals are met.
We will support the establishment of quantitative safety goals through the comment process when the regulatory staff makes such a proposal.
Specific action by PSO is not deemed appropriate prior to that time.
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s LICENSING AND REGULATION ISSUE 2 Develop staff expertise in reactor and fuel cycle chemistry.
REFERENCE:
ACRS letter dated May 16, 1979 (interim report 3),
page 5.
DISCUSSION:
The ACRS has recommended that the capability of the NRC Staff to deal with basic and engineering problems be augmented.
The areas include such important technical areas as the behavior of PWR and BWR coolants, and other materials under radiation conditions; generation, handling, and disposal of radiolytic or other hydrogen; performance of various chemical additives in containment sprays; processing and disposal techniques for low and-high level radioactive wastes; cheemical operations in other parts of the nuclear fuel cycle; and in the chemical treatment operations involved in recovery, decontamination, or de-commissioning of nuclear facilities.
PSO ASSESSMENT:
We believe such an addition to the technical staff would improve the effectiveness of regulatory guidance in these areas and ensure more uniform practice throughout the industry.
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n LICENSING AND REGULATION ISSUE 3 Adequacy of the single failure criteria as a licensing basis.
REFERENCE:
ACRS letter dated May 16, 1979 (interim report 3),
page 5.
DISCUSSION:
Operating experience has suggested to the ACRS that multiple failures and common mode failures are encountered with sufficient frequency that they need more specific consideration.
The focus is to be on establishing an ap-propriate level of reliability for reactor safety si' stems.
PSO ASSESSMENT:
~ ~
This concern seems closely allied with Licensing and Regulation Issue 1, establishment of quantitative safety goals, as a determination of failure rate must be tied to the probable consequences of that failure.
PSO would support, through the public comment process, such an ef-fort on the single failure criterion if it were done con-currently with the establishment of safety goals.
In and of itself, we believe that the current conserva-tive design analysis methods and margin in design limits provide protection beyond the postulated single failure accident.
This additional margin provides, in our opinion, adequate tolerance against many multiple failure scenarios.
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e LICENSING AND REGULATION ISSUE 4 Improvement of the audit calculation capability of the regulatory staff.
REFERENCE:
NUREG-0560, pages 8-14, 8-15.
DISCUSSION:
It has been noted that the NRC presently has only a limited independent capability to perform audit calculations for LOCA events and transients.
This capability is limited to PWRs with U-tube-type steam generators, with reliance placed on hand calculations for the balance of the event.
The regulatory staff should have the ability to indepenently perform quick engineering calculations for transients and small break LOCAs.
PSO ASSESSMENT:
s We certainly support the proposition that the regula-tory staff be able to verify the adequacy of the licensee calculations on a more timely basis by developing full in-house capability rather than relying on piecemeal inputs.
This capability will provide greater assurance that the models do represent appropriate, bounds on real world events to ensure that the health and safety of the public is ade-quately protected during reactor operation.
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c LICENSING AND REGULATION ISSUE 5 Review of BWR analysis codes to ensure that full spectrum analyses using. realistic models can be performed.
REFERENCE:
NUREG-0560, page 8-14.
DISCUSSION:
The computer codes generally used for transient and small LOCA analyses are complex and do not always include provisions for extending the calculations to cover the event duration through the time period until stable cooling (e.g.,
cold shutdown) is achieved.
In some cases, conservative bounding types of assumptions and models are used that may mask out realistic system and equipment behavior.
GE should review and modify as appropriate its computer codes to ensure that they can perform, using realistic models, TAC type, full spectrum analyses identified in Plant Evaluation Issue 1. Furthermore, the codes together with their experimental verification should be submitted for review by the NRC. Until the above verification is complete, existing codes should be used. Actual input parameters and model assumptions should be used to ensure proper tracking of the events.
PSO ASSESSMENT:
PSO will encourage GE to work closely with NRC to review the experimental verification of the realistic transient model codes which vill be used for extending the 2254 241 calculations to stable shutdown cooling.
i LICENSING AND REGULATION ISSUE 6 Modification of standard review plans as appropriate to deal with transient and small break LOCA analysis based on TMI-2 experience.
REFERENCE:
NUREG-0560, page 8-15.
DISCUSSION:
Based on the TMI-2 experience, more explicit guidance is desirable for the calculation of feedwater transients.
This guidance should be placed in regulatory guides and, as appropriate, in the standard review plans.
PSO ASSESSMENT:
We concur that more guidance is desirable with regard to the evaluation of transients and small break LOCA.
Our plan for action is contained within Plant Evaluation Issue 1.
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LICENSING AND REGULATION ISSUE 7 Development of regulatory guides to provide greater guidance in interpretation of General Design Criteria 10, 13, 14 and 15 which relate to anticipated transients.
REFERENCE:
NUREG-0560, page 8-15.
DISCUSSION:
General Design Criteria (GDC), by their very defini-tion, are not specific in nature, although they do reason-ably encompass the necessary requirements for plant design features.
The possible broad interpretations in delineating specific requirements can lead to misunderstanding of the actual event in transient and accident analysis.
An example is the matter of defining a pressure failure, as noted in Appendix A to 10 C.F.R. 50 and its application to such failures as the power operated relief valve.
PSO ASSESSMENT:
As a result of the TMI experience, we are reviewing our interpretation of these particular GDCs against the in-tegrated design of our Black Fox Station.
Our commitment to ongoing review and implementation of the lessons learned will include this area and we support the development of regulatory guidance to assure better predictability of actual events.
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