ML17297B305
| ML17297B305 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 02/28/1982 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-0857, NUREG-0857-S01, NUREG-10857-S1, NUREG-857, NUREG-857-S1, NUDOCS 8203020137 | |
| Download: ML17297B305 (80) | |
Text
NUREG-0857 Supplement No.
1
~e~ Ewelluatmon Report related to the operation of Palo Verde Nuclear Generating Station, Units 1, 2, and 3 Docket Nos. STN 50-528, STN 50-529, and STN 50-530 Arizona Public Service Company, et al ~
U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation February 1982
I
TABLE OF CONTENTS 1
INTRODUCTION AND GENERAL DISCUSSION...
- l. 1 Introduction.
1.9 Summary of Outstanding Issues.....
1.10 Confirmatory Issues...
- 1. 11 License Conditions 2
SITE CHARACTERISTICS 2.5 Geology and Seismology.
2.5.4 Stability of Subsurface Materials and Foundations....
2.5.4.3 Foundation Stability..
3 DESIGN CRITERIA - STRUCTURE, COMPONENTS, EQUIPMENT AND SYSTEMS..
3.9 Mechanical Systems and Components..
3.9.2 Dynamic Testing and Analysis of Systems, Components, and Equipment.......
5 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS 5.2 Integrity of Reactor Coolant Pressure Boundary.....
5.2. 1 Compliance with Codes and Code Cases 5.2. 1. 1 Compliance with 10 CFR 50.55a.
5.2.4 Reactor Coolant Pressure Boundary Inservice Inspection and Testing.
5.2.4
~ 1 Evaluation of Compliance for Palo Verde Unit 1 with 10 CFR 50.55a(g)......
',2.4.2 Evaluation of Compliance for Palo Verde Units 1 and 2 with 10 CFR 50.55a(g).........
6 ENGINEERED SAFETY FEATURES 6.6 Inservice Inspection of Class 2 and 3 Components.........
6.6.1 Evaluation of Compliance for Palo Verde Unit 1 with 10 CFR 50.55a(g).
6.6.2 Evaluation of Compliance for Palo Verde Units 2 and 3 with 10 CFR 50.55a(g).
Pacae l-l 1-1 1-2 1-3 2-1 2-1 2-1 3-1 3-1 5-1 5-1 5-1 5-1 5-2 6-1 6-1 6-1 Palo Verde SSERl
TABLE OF CONTENTS (Continued)
,7 INSTRUMENTATION AND CONTROLS 7.3 Engineered-Safety Features Actuation System.
7.3.3 Conformance with IE Bulletin 80-06...
7.6 Interlock Systems Important to Safety..
7.6.2 Low Temperature Overpressure Protection..............
9 AUXILIARYSYSTEMS
- 9. 1 Fuel Storage Facility.........
- 9. 1.2 Spent Fuel Storage.
- 9. 1.4 Fuel Handling System...
9.6 References 13 CONDUCT OF OPERATIONS..
13.2 Training Program.
13.2. 1 Training Program for Licensed Plant Staff.........
- 13. 3 Emergency Planning...............
13.6 Physical Security.
18 REPORT OF THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS...........
20 FINANCIAL QUALIFICATIONS 22 TMI-2 RE(UIREMENTS
- 22. 2 Evaluation of TMI Requirements I.A.l.1 Shift Technical Advisor.;
I.D. 1 Control Room Design Review.
II.B. 4 Degraded Core - Training.
~Pa e
7-1 7-1 7-1 7-1 9-1 9-1 13-1 13-1 13-1 13-3 13-4 18-1 20-1 22-1 22-1 22-1 22-1 22-3
TABLE OF CONTENTS (Continued)
APPENDICES
~Pa e
APPENDIX A APPENDIX B APPENDIX C
CONTINUATION OF CHRONOLOGY OF RADiOLOGICAL REVIEW......
REPORT BY THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS, DATED DECEMBER 15, 1981.........
EMERGENCY PREPAREDNESS EVALUATION REPORT..
A-1 B-1 C-1 APPENDIX 0 PRINCIPAL CONTRIBUTORS APPENDIX E ERRATA TO SAFETY EVALUATION REPORT 0"1 E-1 Palo Verde SSER1
7 1
- 1. INTRODUCTION AND GENERAL DISCUSSION
- 1. 1 Introduction On November. 13, 1981, the Nuclear Regulatory Commission (NRC) staff issued its Safety Evaluation Report (SER) relating to the application for licenses to operate the Palo Verde Nuclear Generating Station, Unit Nos.
1, 2 and 3
(PVNGS 1-3).
The application was submitted by the Arizona Public Service Company (APS or the applicant) initially on behalf of itself, the Salt River Project Agricultural Improvement and Power District, Southern California Edison
- Company, El Paso Electric Company, and Public Service Company of New Mexico.
By letter dated July 31, 1981, the application was amended to include the Los Angeles District of Mater and Power and South California Public Power Authority.
By letter dated November 6, 1981, the application was further amended to include the M-S-R Public Power Agency.
In the SER, the staff identified certain issues where either further informa-tion was required of the applicant or additional staff effort was necessary to complete the review of the application.
The purpose of this supplement is to update the SER by providing (1) the evaluation of additional information sub-mitted by the applicant since the SER was issued, (2) the evaluation of the matters that the staff had under review when the SER was issued, and (3) the response to comments made by the Advisory Committee on Reactor Safeguards in its report, dated December 15, 1981 (Appendix B).
Each of the following sections of this supplement is numbered the same as the section of the SER that is being updated
- and, unless otherwise noted, the discussions are supplementary to and not in lieu of the discussions in the SER.
Appendix A to this supplement is a continuation of the chronology.
Appendix B
is the report by the Advisory Committee on Reactor Safeguards, dated December 15, 1981.
Appendix C is the staff evaluation report on emergency preparedness.
Appendix 0 is the list of principal contributors to the staff review.
Appendix E
is the errata to the SER.
1.9 Summar of Outstandin Issues Section 1.9 of the SER contained a list of outstanding issues.
Two of those issues were resolved after the issuance of the SER.
These are listed below, along with the section of this supplement wherein their resolution is discussed.
(1)
Seismic and LOCA loads (Section 3.9.2)
(2)
Secur ity (13. 6)
At this time, a number of outstanding issues remain to be resolved.
These are listed below along with the section of this supplement and/or the SER wherein each issue is discussed.
The resolution of these remaining issues will be addressed in a subsequent supplement to the SER.
Palo Verde SSER1
(1)
Emergency preparedness (2.3.3, 13.3)
(2)
Ultimate heat sink (2. 4. 4.2,
- 9. 2. 5)
(3)
Dynamic effects of pipe rupture (3.6.1, 3.6.2)
(4)
Cable tray design (3.7. 1)
(5)
Seismic and LOCA loads (4.2. 1)
(6)
PSI/ISI program (3. 9. 6, 3. 9. 7,
- 5. 2. 4, 6. 3. 3, 6. 6)
(7)
Environmental qualification (3. 10)
(8)
Containment pressure analysis (6.2. 1.5)
(9)
Hydrogen in RDT room (6.2.5)
(10)
LOCA 'doses (6.4, 6.5.2, 15.4.8, III.D.3.4)
(ll) Control system failures {7.7.2)
(12) Alternate shutdown (7.4.2, 9.5.1.6)
(13) Under voltage protection (8.4.7)
(14) Diesel generator tank corrosion (9.5.4.2)
(15) Impurity levels (17.5)
(16)
ACRS comments (18)
(17) Financial qualifications (20)
- l. 10 Confirmator Issues Section 1.10 of the SER contained a list of issues that had been essentially res'olved to the staff's satisfaction, but for which certain confirmatory infor-mation was to be provided by the applicant.
Subsequent to the issuance of the SER, the applicant provided the required confirmatory information for a number of those issues.
These are listed below, along with the section of this supplement wherein the issue is resolved.
(1)
Foundation Stability (2.5.4.3)
(2)
Compliance with codes and code cases (5.2. l. 1)
{3)
Automatic loading of non-lE loads (7.3.3)
(4)
Spent fuel pool (9. 1.2,
- 9. 1.4)
(5)
Training (13.2, I.A.l.l, II.B.4)
At this time, information is still pending on a number of confirmatory issues.
These are listed below along with the section of thi s supplement and/or the SER wherein each issue is discussed.
The resolution of these remaining issues will be addressed in a subsequent supplement to the SER.
(1)
Exclusion area control (2. 1.2)
(2)
Maximum earthquake (2.5.2.3)
(3)
Missile protection (3. 5. 2, 3. 5. 3)
(4)
Seismic system anlaysis (3.7.2)
(5)
Dynamic testing and analysis of components
{3.9.2, 3.9.3. 1)
{6)
Pump and valve operability (3.9.3.2, II.E.4.2)
(7)
RCS monitoring (4.2.5, 4.4. 1)
(8)
Pressure-temperature curves (5.3.2q (9)
Valve modifications (5. 4. 3, 6. 2. 4, 7. 3. 2. 2)
(10) Modifications to test programs {5.4.3, 6.2.6, 6.3.1, 8.2.2, 14)
{11) Override, interlocks, and alarms (5.4.3, 7.3.2.2, 7.6.2, 7.6.3, 8.3.2. 1)
(12) TMI instrumentation (6.2.1.1, II.D.3, II.F.1, II.F.2)
(13) Setpoint values (6.2.4, 7.2.4)
(14)
CPC softwar e (7. 2. 3,
- 7. 2. 4)
(15) Instrumentation and control procedures and testing (7.3.2.1, 7.4.2, 7.4.3,
- 7. 7. 3)
Palo Verde SSERl 1-2
(16) Confirmatory site visits (7.1.3.5, 8.1, 9.5.1)
(17) Containment Electrical Penetrations (8.4.3)
(18)
Load Sequencer (8.4.6)
(19) Fire protection modifications (9.5.1)
(20) Training (13.1.2.3)
(21) Procedures (13.5, 15.3.9, I.A.1.2, I.A.1.3, I.C.1, I.C.3, I.C.4, I.C.5, I. C. 6, II.B. 1, II.F. 2; II.K. 1, II.K. 3, Item A-44 of Appendix C to SER).
(22) Control room design review (I.D. 1)
(23) Relief and safety valve tests (II.D.1)
(24) Primary coolant outside containment (III.D.l.1)
- l. 11 License Conditions Section l. 11 of the SER lists several issues for which a condition will be included in the operating license to ensure that NRC requirements are met during plant operations.
In addition to the twelve issues already listed in the SER, this supplement has identified another issues for which a condition will be included in the operating license.
The issue, with an appropriate reference to the section of this supplement, is listed below.
(13) Loading of non-lE loads (7.3.3)
Palo Verde SSERl 1-3
i'
2 SITE CHARACTERISTICS
- 2. 5 Geol o and Seismolo 2.5.4 Stabilit of Subsurface Materials and Foundations
- 2. 5. 4. 3 Foundation Stabilit (2)
Heave/Settlement:
The SER discussed the difference between the predicted and measured values of heave and settlement beneath the powerblock structures. 'he SER also presented the applicant's assessment that the differences were due to the differences between the actual and projected construction schedules.
The sta'ff concurred with the applicant's assess-ment provided that the applicant submitted additional information to verify the conclusion.
By letter dated December 4, 1981, the applicant submitted the necessary information to verify that the building completion dates assumed in the settlement analyses were several months earlier than the actual completion dates.
Therefore, this matter is closed.
The SER also stated that adjustments should be incorporated into the applicant's settlement monitoring program and the SER presented the ele-ments of an acceptable program.
By letter dated December 4, 1981, the applicant has agreed to modify the settlement program in accordance with the recommendations presented in the SER.
Therefore, this matter is now resolved.
(5)
Lateral Loads:
The SER stated that the applicant had committed to provide addit)onal information to confirm that the lateral soil loads used in the design included surcharge effects and seismic-induced earth pressures.
By letter dated December 4, 1981, the applicant provided information relating to the lateral soil loads, the slope preparations and the surcharge effects from the adjacent structures.
The staff has reviewed the information sub-mitted and found that the lateral soil loads used in the design appropriately include surcharge effects from adjacent structures and seismic-induced earth pressures and, therefore, are acceptable.
Therefore, this matter is now closed.
Palo Verde SSERl 2-1
3 DESIGN CRITERIA - STRUCTURE, COMPONENTS, EQUIPMENT AND SYSTEMS 3.9 Mechanical S stems and Com onents 3.9.2 D namic Testin and Anal sis of S stems Com onents and E ui ment In Section 3.9.2 of the SER, the staff, identified an open issue regarding the plant-specific results for the loss-of-coolant-accident (LOCA) asymmetric loads evaluation.
Subsequently, the applicant responded to this concern by letter, dated December 4, 1981.
The staff has evaluated the applicant's response using the guidelines of NUREG-0609, "Asymmetric Blowdown Loads on PWR Primary Systems".
The review included:
1.
the structural analysis of the reactor coolant system (RCS),
2.
the evaluation of the supports for the reactor vessel, reactor coolant
- system, and attached ECCS piping, 3.
the structural analysis and evaluation of the reactor internals and
- CEDMs, and 4.
the analysis and evaluation of ECCS piping attached to the RCS.
In the asymmetric loads evaluation, the applicant first determined the design basis pipe break in the primary reactor coolant piping system.
For the Palo Verde plant, the design basis pipe break in the reactor coolant system is a 350 in~ cold leg break for all components except the fuel assemblies.
The fuel assemblies were designed for a 100 in~ hot leg break.
The methodologies used by the applicant to determine the limiting pipe break areas, and to calculate the resulting blowdown forces and mass-energy
The staff evaluation, as presented in the CESSAR
- SER, concluded that these methodologies are acceptable.
The computer code CEFLASH-4B was used to determine the transient pressures, flow rates, and densities throughout the primary system following the postulated pipe break in the reactor coolant system.
CEFLASH-4B is an updated version of CEFLASH-4A which is an approved and acceptable program as stated in NUREG-0609.
In,its structural evaluation of the primary system, the applicant evaluated the following components for the asymmetric loads:
l.
2.
3.
4.
5.
6.
7.
8.
reactor vessel, steam generator, reactor coolant
- pump, pressurizer, reactor coolant piping, component supports for all major components, reactor internals, and ECCS piping attached to the RCS.
Palo Verde SSERl
Stresses in the above components were combined and evaluated in accordance with th t ff'eptance criteria.
The above components were de g
withstand the loads associated with the LOCA design bases pipe bre k
esa sac as in combination wi e
oa th th 1 ads associated with the safe shutdown earthquake (SSE).
uidelines The LOCA and SSE loads were combined in accordance with the guide ines presented in NUREG-0484 (Rev. l), "Methodology for Combining Dynamic Responses."
I Th ECCS 'ttached to the reactor coolant system has been evaluated by the applicant.
The applicant has evaluated the structural integri yt and func-tional capa i i y o e
b 1't f th ECCS piping to withstand the combined effects of a rienced b
The primary system pipe break loads are experience y
attached ECCS piping via the dynamic motion of the primary sy p
'n s stem.
The rimary s stem i in motions due to pipe break loadings were calculated by using the methodologies described in the h
The primary motions were inputted
. the ECCS i in and as a displacement time-history into a decoupled model of. the piping evaluated by using dynamic analyses methods.
The ECCS piping attached to the broken primary pipe was evaluated to Service D 1'
'usin the rules of Appendix F of the ASME Code Section III.
The ECCS piping attached to the unbroken loops of the primary-y to assure that the functional capability of the ECCS piping is maintained.
e staff has reviewed the applicant's procedure for the evaluation of the ECCS-piping and finds it to be acceptable.
Based on the review o e app i f th 1 cant's procedures used in evaluating the LOCA m etric loads the staff findings are as follows:
The applicant has m
s met the 1 Desi n Criteria 2 and 4 with respect to the relevant requirements of Genera g
desi n of systems and components important to safety to wi s an e
nd the a
ro riate combinations of the effects of normal and ff ts of the safe shutdown earthquake postulated accident conditions with the e
ec s
o e
(SSE) by performing a dynamic system analysis which provides an accepta e
for confirmin the structural design adequacy of the reactor internals nbro e
th t d th combined dynamic loads of postulated
)
d the SSE.
The analysis provides adequate nbroken i in loops to wi s an e
loss of coolant accidents (LOCA) an e
nents of the that the combined stresses and strains in the componen s
o e
assurance a
e c
s will not exceed the allowable reactor coolant system and reactor internals wi d th t the design stress and strain limits for the materials of construction, an a
t of the reactor resulting deflections or displacements at any structural elemen o
e ls wi 11 not distort the reactor internals geometry to the extent that Th ethods used for components analysis have core cooling may be impaired.
e me o
s us been found to be compatible with those used for the systems ana ysis.
e s of corn onent and system analyses are, therefore, proposed combinations o
p al inte rit of the reactor internals acceptable.
The assurance of structura] ln egrl y 0 t
d under LOCA conditions for the most adverse postulated loading even provi es added confidence that the design will withstand a spectrum of lesser pipe breaks and seismic loading events.
Palo Verde SSER1 3-2
5 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS 5.2 Inte rit of Reactor Coolant Presure Boundar 5.2. 1 Com liance with Codes and Code Cases 5.2. 1. 1 Com liance with 10 CFR 50.55a Section 5.2. l. 1 of the SER included a footnote which stated that staff acceptance of the classification in Table 5.2-1 of the PVNGS FSAR was contingent upon the applicant identifying the code edition and addenda used in the construction of the RCPB components within the CESSAR scope of supply.
Subsequently, the applicant submitted a revised Table 5.2-1 which appropriately identified the code edition and addenda used in the construction of those RCPB components in conformance with 10 CFR 50.55a.
Therefore, this matter is now resolved.
5.2.4 Reactor Coolant Pressure Boundar Inservice Ins ection and Testin The SER stated that after the applicant submits its Preservice Inspection (PSI)
- Program, the staff will perform a technical evaluation and report the conclu-sions in a supplement to the SER.
By letter dated November 13, 1981, the applicant submitted its PSI Program for Unit 1.
The staff evaluation of this program is presented below.
5.2.4. 1 Evaluation of Com liance for Palo Verde Unit 1 With 10 CFR 50.55a(
)
The PSI Program for Palo Verde Unit 1 was submitted by the applicant in a letter dated November 13, 1981.
The PSI Program is based on the 1974 Edition of Section XI of the ASME Code including Addenda through Summer 1975 and Appendix=III of Section XI, Winter 1975 Addenda, for ASME Code Class 1 and 2
piping and components.
The preservice examination of the principal components of the reactor coolant pressure
- boundary, such as the reactor vessel, steam generators,-pressurizer and reactor coolant
- pumps, were performed in the fabrication shop.
Certain welds in these principal components and welds in the primary piping system will be inspected during the field preservice examination.
The staff has reviewed the selection of the primary pressure boundary welds subject to examination and finds the examination sample to be. acceptable.
The specific areas where the Code requirements cannot be met will be identified after the examinations are performed.
The staff evaluation of the preservice examination program will be presented in a subsequent supplement to the SER after the applicant identifies all plant-specific areas where the Code requirements cannot be met and provides a supporting technical justification.
Palo Verde SSER1 5-1
5.2.4.2 Evaluation of Com liance for Palo Verde Units 2 and 3 Mith 10 CFR 50.55a A PSI Program for Units 2 and 3 has not been submitted.
The regulations permit the applicant to meet the requirements set forth in subsequent editions of Section XI which are incorporated by reference in Section 50.55a(b),
subject to the limitations and modifications listed therein.
The applicant is considering updating the PSI Program to meet the requirements of later editions of Section XI.
After the applicant has made a decision about updating the examination requirements of Units 2 and 3, the staff will perform a technical evaluation and report its conclusions in a subsequent supplement to the SER.
The staff has identified a technical issue concerning the ultrasonic testing calibration blocks used for preservice examination of the reactor vessels.
The Unit 1 calibration standards were used for the examination of all three vessels.
Since the vessel calibration blocks were not from actual component dropouts or prologations from those vessels, the staff will require a technical justification that demonstrates that the inservice examination results will be representative or conservative.
The staff will evaluate the preservice examination program and plant-unique requests for relief after all the requests are identified by the applicant.
Palo Verde SSERl 5-2
6 ENGINEERED SAFETY FEATURES 6 '
Inservice Ins ection of Class 2 and 3
Com onents The SER stated that the Preservice Inspection (PSI) Program for Class 2 and 3
components was in preparation.
By letter dated November 13, 1981, the appli-cant submitted its PSI Program for Unit 1.
The staff evaluation of this program is presented below.
6.6. 1 Evaluation of Com liance for Palo Verde Unit 1 With 10 CFR 50.55a(
)
The PSI Program for Palo Verde Unit 1 was submitted by the applicant in a letter dated November 13, 1981.
The PSI Program is based on the 1974 Edition of Section XI of the ASME Code including Addenda through Summer 1975 and Appendix III of Section XI, Winter 1975 Addenda, for ASh1E Code Class 2
components, except for some piping welds in the
- RHR, ECCS, and containment spray systems.
For the aforementioned welds, the applicant is considering updating the program to the examination method required in later Editions of Section XI, as permitted by 10 CFR 50.55a(g).
The examination requirements for ASNE Code Class 3 components consist of visual inspections during hydrostatic tests.
The staff has reviewed the selection of the welds subject to examination and finds the sample to be acceptable.
The specific areas where the Code requirements cannot be met'will be identified after the examinations are performed.
The staff evaluation of the preservice examination program will be presented in a subsequent supplement to the SER after the applicant identifies all plant-specific areas where the Code requirements cannot be met and provides a supporting technical justification.
6.6.2 Evaluation of Com liance for Palo Verde Units 2 and 3 With 10 CFR 50.55a A PSI Program for Units 2 and 3 has not been submitted.
The regulations permit the applicant to meet the requirements set forth in subsequent editions of Section XI which are incorporated by reference in Section 50.55a(b),
subject to the limitations and modifications listed therein.
The applicant is considering updating the PSI Program to meet the requirements of later editions of Section XI.
After the applicant has made a decision about updating the examination requirements of Units 2 and 3, the staff will evaluate the PSI Program and plant-uaique requests for relief from impractical examination requirements and report its conclusions in a subsequent supplement to the SER.
Palo Verde SSER1
)
I "J
I
7 INSTRUMENTATION AND CONTROLS 7.3 En ineered-Safet Features Actuation S stem 7.3.3 Conformance With IE Bulletin 80-06 (b)
The SER stated that the staff found unacceptable the automatic reloading of non-1E loads on the 1E buses.
Subsequently, by letter dated October 23,
- 1981, the applicant committed to modify the design, such that non-lE equipment will not be automatically reloaded'n the 1E buses.
Reinstatement of these non-1E loads will require deliberate operator action.
This design change will result in conformance with the recom-mendations of IE Bulletin 80-06 and, therefore, is acceptable.
License Condition The staff will condition the PVNGS license to require that the design modifica-tions described
- above, which the applicant has committed to perform, be imple-mented prior to fuel load.
7.6 Interlock S stems Im ortant to Safet 7.6.2 Low Tem erature Over ressure Protection The SER stated that the staff will confirm that alarms are provided which indicate to the operator that a low temperature overpressure event is in progress or which indicate that the shutdown cooling relief valves should be armed by manual opening of the SCS isolation valves.
By letter dated October 23,
- 1981, the applicant has committed to add the alarms described above.
The staff finds this commitment to be acceptable.
These alarms should be added prior to performing those preoperational functional tests during which the primary system is susceptible to becoming water solid.
Until such alarms are available, the applicant shall use appropriate administrative procedures during those tests to provide to the operator the indications stated above.
Palo Verde SSERl 7-1
I P
9 AUXILIARYSYSTEMS 9.1 Fuel Stora e Facilit 9.1.2 S ent Fuel Stora e
Part II The SER stated that the applicant had committed to.pruvide information on the compatibility of the nuclear (neutron) poison tubes with the storage pool environment and that the staff would confirm the materials/fluids compatibility in the spent fuel pool.
By letters dated October 22, 1981 and November 9, 1981, the applicant subsequently provided the required information.
The staff evaluation of the information is presented below.
The neutron poison tubes consist of sheets of Boraflex which are sandwiched between 304 stainless steel walls.
The walls are vented to permit the escape of radiolytically produced gases from the Boraflex.
Boraflex consists of boron carbide power in a rubber-like silicone polymeric matrix.
Boraflex is composed of non-conductive materials and, therefore, will not develop a galvanic potential in contact with the metal components.
As a
- result, the corrosion that will occur in the spent fuel storage pool environment should be of little significance during the 40-year life of the plant.
Boraflex has undergone extensive testing to study the effects of gamma irradiation in various environments and the effects of water at elevated temperatures.
The results indicate that Boraflex will not undergo significant degradation during the expected service life of 40 years (Anderson, J. S., August 1978; Anderson, J. S., July 1979; Anderson, J.
S ~, August 1979).
Therefore, the staff concludes that the materials/fluids compatibility of neutron poison tubes with the storage pool environment is adequate for the expected service life of 40 years and is acceptable.
This matter is now closed.
- 9. 1.4 Fuel Handlin S stem The SER indicated that the applicant's initial response to NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants," only addressed heavy load lifts required prior to initial fuel load.
The applicant subsequently revised the response to NUREG-0612 to address safe handling of all heavy loads in safety related plant
- areas, including those required when performing plant maintenance.
The review and evaluation of this information will be performed at a future date in connection with the long term review of NUREG-0612.
Therefore, the staff concludes that the applicant's commitments concerning NUREG-0612 are acceptable.
Palo Verde SSERl 9-1
9.6 References Brand Industries Incor orated Re ort on the Effects of High Temperature Borated Mater xposure on n Absorbin Materials
" Brand Industries, Inc., Report 748-21-1, August 1978.
d t Study of Boraflex Neutron Shielding Materials,"
- Anderson, J. S., "Irra la son Brand Industries, Inc., Report 748-10-1, Ju y J. S., "Boraflex Neutron ie in Sh'd' Material-Product Performance Data,"
R t 748-30-1, August 1979.
Brand Industries, Inc.,
epor Palo Verde SSERl 9-2
13 CONDUCT OF OPERATIONS 12.2 T~i 13.2. 1 Trainin Pro ram for Licensed Plant Staff The SER stated that the staff had identified four discrepancies in the appli-cant's training program for licensed plant staff.
Subsequently, by letters dated October 28, 1981 and January 13, 1982, the applicant submitted additional information to address the deficiencies listed in the SER.
The deficiencies and the applicant's responses are as follows:
Defici enc (1)
The staff found that Figure 13.2. 1 of the PVNGS FSAR (training program schedule),
in relation to the fuel loading date, and Figure 13.2.2, showing personnel by position titles that are scheduled for various training courses, to be obsolete due to the time lapse from the original submittal.
Also, the material presented did not reflect the licensing requirement changes that have taken place since the TMI-2 incident and, therefore, should be modified to include post-TMI requirements.
~Res ense Figure 13.2. 1 and Figure 13.2.2 have now been revised and updated.
A new figure, combining the training program and personnel receiving the training by position titles, is presented in the revised Figure 13.2. 1, dated January 13, 1982.
Also, the licensing requirement
- changes, including post-TMI requirements, will be incorporated in a revision to Section 13.2. 1 of the FSAR.
The applicant's training program has been modified to meet the additional requirements contained in Enclosure 1, Item A. 2. C, of. H.
R. Denton' March 28, 1980 letter, which states that the training programs shall be modified; as necessary, to provide:
1.
Training in heat transfer, fluid flow, and thermodynamics.
2.
Training in the use of installed plant systems to control or mitigate an accident in which the core is severely damaged.
3.
Increased emphasis on reactor and plant transients.
The training time allowed for each of the six phases in the training program for candidates for an operator license or senior operator license, as described in Section 13.2. 1. 1. 1, was adequate.
However, there was insufficient informa-tion provided, as required by the Standard Review Plan, on the organization teaching the courses or supervising instruction, and the position titles for which the courses are given.
Palo Verde SSERl 13-1
~Res onse The Palo Verde Nuclear Training Manager, under the direction of the Manager of Nuclear Operations, has overall responsibility for the conduct and administra-tion of the training program for staff personnel.
The organization teaching the courses or supervising instruction is as follows:
1.
Advanced Engineering Training for the initial group of senior operator candidates has been taught on-site by the Center for Nuclear Studies of Memphis State University.
2.
Research or Training Reactor training is provided by several vendors and universities.
3.
Nuclear Steam Supply System (NSSS) 'training has been conducted on-site by Combustion Engineering, Inc.
The applicant' training department or a
consultant will conduct subsequent equivalent courses.
4.
Balance of, plant systems training has been conducted on-site by General Electric Company and the applicant's training department staff..
5.
Simulator training will be conducted by the Training Department staff and PVNGS will take full advantage of the on-site plant-specific simulator.
6.
On-the-job training involves participation in startup testing, procedure preparation, and qualification on plant systems under the direction of the Operations Superintendent.
Revised FSAR Figure 13.2. 1 gives the position titles of personnel to whom courses are given.
Deficienc (3}
Similar comments as item 2 above apply to the modifications of the program to compensate for different experience levels of training as described in the PVNGS FSAR Section 13.2. l. 1.2.
~Res onse Modifications to the training program to compensate for different experience levels of training will be determined by the PVNGS Manager of Nuclear Operations, who has overall responsibility for the conduct and administration of the training programs for all staff personnel.
Deficienc (4
There was insufficient information provided on the operator requalification and replacement training program as required by the Standard Review Plan.
Also missing was the additional requirement imposed by the Harold R.
Denton letter of March 28, 1980.
Palo Verde SSER1 13-2
~Ree once The requalification program will include lectures, on-the-job training, and evaluations on a regular and continuing basis.
The plant specific PVNGS
'imulator will be utilized to fulfillcertain requirements of the requalifi-cation program.
Plant staff personnel who maintain a current Senior Operator or Operator license will participate in the requalification program.
The requalification program will commence within three months after receipt of a plant operating license and be conducted on a two-year, repeating cycle.
A lecture series, consisting of at least six lectures, will be presented annually, 'consistent with plant schedule.
Subjects and depth of coverage will be determined by evaluation of the annual evaluation examination results, oral interviews and practical demonstrations which indicate general weaknesses in operator knowledge.
No more than 50K of the lecture series shall consist of videotapes and films.
All licensed operators will review the content of off-normal, emergency.,
and security procedures on a regularly scheduled basis.
The review will consist of any of the following: self-study, lectures conducted by shift supervisors, or simulated walk through.
All off-normal, emergency, and security procedures will be reviewed in each requalification period.
Changes to the plant design, procedures, technical specifications and limiting conditions of operation will be promulgated to licensed operators.
P The performance standards applied to the annual requalification examination shall be used in evaluating the results of the oral and written examinations.
If the performance standards are not met, the licensed individual shall complete an accelerated requalification program prior to resuming licensed duties.
A permanent record shall be maintained for each operator containing verification of each program completion and the overall grade scores for the two-year program.
This permanent record file shall be maintained for the life of the facility and conforms with the requirements of Appendix A to 10 CFR Part 55.
Conclusion On the basis of its review, the staff has concluded that the training programs and schedules for all staff members meet the provisions of Standard Review Plan 13;2. 1, Reactor Operator Training, (NUREG-0800),
10 CFR Part 55, Regulatory Guide 1.8, ANSI N18. 1, 1971, and H.
R.
Denton's March 28, 1980 letter and are acceptable for the preoperational test program, for operator 1icensing exami-nations and fuel loading.
Therefore, this matter is now resolved.
13.3 Emer enc Plannin The SER stated that the applicant had not yet responded to the staff review comments on the PVNGS Emergency
- Plan, which were forwarded to the applicant by letter dated October 14, 1981.
Palo Verde SSER1 13-3
Subsequently, the staff conducted a site visit on November 10, 1981 and met with the applicant on November ll, 1981 at which time the staff review comments were discussed.
As a result, by letter dated November 30, 1981, the applicant submitted acceptable commitments that addressed the deficiencies identified in the staff review comments.
A detailed evaluation of the applicant's Emergency Plan is presented in Appendix C to this supplement.
As stated in the conclusions to Appendix C, the staff's evaluation of the overall state of emergency preparedness for the Palo Verde site will be presented in a subsequent supplement to the SER upon completion of the six items noted in Appendix C.
13.6 Ph sical Securit Section 13.6 of the SER provided 'the status of the staff's review of the appli-cant's 'Physical Security Plan and stated that the applicant had not yet responded to comments on the Guard Training and gualification Plan and the Safeguard Contingency Plan.
Subsequently, the applicant filed acceptable responses to those comments and the staff completed its review.
Therefore, the discussion in Section 13.6 of the SER should be replaced with the following discussion.
The applicant has submitted security plans entitled "Palo Verde Nuclear Generating Station Security Plan,"
and "Palo Verde Nuclear Generating Station Training and gualification Plan," for protection against radiological sabotage.
The contingency plan is incorporated as Chapter 8 in the Palo Verde Nuclear Generating Station Security Plan.
The plans were reviewed in accordance with Section 13.6 "Physical Security" of the July 1981 edition of the "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants."
(SRP, NUREG-0800)
As a result of the staff's evaluation, certain portions of these plans were identified as requiring additional information and upgrading to satisfy the requirements of Section 73.55 and Appendices B and C of 10 CFR Part 73.
The applicant filed revisions to these plans which satisfied these requirements.
The revised plans are considered to comply with the Commission's regulations contained in 10 CFR Parts 50 and 73 and, therefore, are acceptable.
An ongoing review of the progress of the implementation of these plans will be performed by the staff to assure conformance with the performance requirements of 10 CFR Part 73.
The identification of vital areas and measures used to control access to these
- areas, as descirbed in the plan, may be subject to amendments in the future.
The applicant's security plans are being protected from unauthorized disclosure in accordance with Section 73.21 of 10 CFR Part 73.
Palo Verde SSERl 13-4
18 REPORT OF THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS The Advisory Committee on Reactor Safeguards reviewed the application for licenses to operate the Palo Verde Nuclear Generating Station, Units 1, 2 and 3
during its 260th meeting on December 10-12, l981.
A copy of the Committee's
- report, dated December 15, 1981, is included as Appendix B to this supplement.
In its report, the Committee made a number of comments which included requests and recommendations.
The staff's response to those items is presented below.
A licant's Or anization for Technical Su ort of the 0 eratin Plant The Committee believes that the applicant should promptly analyze the skill requirements needed to support operations and make certain that the necessary capabilities will be available when needed.
The Committee requested an update on the organizational arrangement in about one year.
The staff will continue to follow the development of the applicant's organiza-tional plan for technical support of the operating plant and will provide the Committee with an update on the organizational arrangement in about one year.
Effective Use of Simulators for Trainin The Committee recommended that Arizona Public Service Company examine industry-sponsored programs concerning effective use of simulators for training and make certain that its approach takes account of current understanding of simulator training limitations.
As stated in Section 13.2. 1 of the SER and this supplement, the staff has completed its review of the applicant's training program for licensed plant staff and finds it to be acceptable.
As noted in the SER, the program inte-grates simulator training with five other phases of training.
In addition, candidates for licenses on PVNGS Units 2 and 3 will have an opportunity to obtain experience an operating PVNGS units as part of their training program.
Both the staff and the applicant are aware of industry sponsored programs concerning effective use of simulators.
Where simulator training limitations are noted, they will be appropriately taken into account in the training program.
Reliabi lit of Shutdown Heat Removal S stem The Committee stated that:
"In the Palo Verde design the primary system does not include capa-bility for rapid, direct depressurization when the plant has been shut down.
This places extra importance on the reliability of the auxiliary feedwater system and makes it necessary that the NRC Staff and the Applicant assure the availability and dependability of 'this system for a wide variety of transients.
It also places extra requirements on the continued integrity of the two steam generators as the only method of heat removal immediately after shutdown.
The Palo Verde SSERl 18-1
ACRS recommends that the NRC Staff and the Arizona Public Service Company give additional attention to the matter of shutdown heat removal for Palo Verde and develop a detailed evaluation and justifi-cation for the position judged to be acceptable.
The Committee wishes to be kept informed."
The staff is currently evaluating the above Committee comments and recommendation and will provide a response in a subsequent supplement to the SER.
Studies on S stems Interactions and S stems Reliabilit The Committee stated that Arizona Public Service Company should expand its studies on systems. interactions and systems reliability.
Item A-17 in Appendix C to the SER discusses the ongoing staff efforts to reach a generic resolution to the issue of systems interactions in nuclear power plants.
It is expected that the development of systematic ways to identify, rank and evaluate systems interactions will go further to reduce the likelihood of inter-system failures resulting in the loss of plant safety functions and,
- hence, improve systems reliability.
After resolution of this generic issue, the staff will determine where additional studies by Arizona Public Service Company are required.
Com letion of Construction and Preo erational Testin The Committee requested that, prior to fuel loading on Unit 1, a report be provided to the Committee describing significant construction deficiencies and their disposition, the effectiveness of the quality assurance
- program, and the results of the preoperational test program.
'In addition, the Committee stated that a review of the startup experience on Unit 1 should be made prior to fuel loading on Unit 2 and that the Committee be kept informed.
Prior to fuel loading on Unit 1, the staff will provide a report to the Committee describing significant construction deficiencies and their disposition, the effectiveness of the quality assurance
- program, and the results of the preopera-tional test program.
Prior to fuel loading on Unit 2, the staff will review the startup experience on Unit 1 and inform the Committee of the results of its review.
Instrumentation to Follow the Course of An Accident The Committee's report included additional comments by ACRS member M. Bender and by ACRS members H.
W.
Lewis and N.
S. Plesset which express concerns about the capability of proposed instrumentation systems to provide unambiguous infor-mation for detection of inadequate core cooling and about the adequacy of the staff acceptance criteria for this instrumentation.
Mr. Bender also expressed concerns regarding the general NRC requirements for instrumentation to follow the course of an accident as outlined in Regulatory Guide 1.97.
Regulatory Guide 1.97 was developed over a period of several years (beginning before the TMI-2 event) and made extensive use of work performed by the IEEE and ANS to develop IEEE Std. 497-1977, "Trial-Use Standard Criteria for Post Accident Monitoring 'Instrumentation for Nuclear Power Generating Stations,"
and Pal o Ver de SSERl 18-2
ANSI/ANS-45-1980, "Criteria for Accident Monitoring Functions in Light-Water-Cooled Reactors."
Throughout the period of development of Regulatory Guide 1.97, comments from industry and the ACRS were received and considered.
Atomic Industrial Forum input was also used in development of the list of parameters in the final regulatory guide.
The staff believes that the specific criteria proposed by Mr. Bender in the ACRS letter of December 15,
- 1981, were included in developing the guidance delineated in Regulatory Guide 1.97.
The staff agrees that only a few additions to the pre-TMI-accident instrumentation should be necessary.
Furthermore, since regulatory guides are not regulations, individual licensees may take exceptions to the guidance in Regulatory Guide 1.97 to reflect plant specific design features.
The effect of implementing Regulatory Guide 1.97 will not be to provide indica-tions for a large number of additional plant parameters, since most of the parameters are currently indicated in plant control rooms.
It should also be noted that not all parameters listed in the regulatory guide are to be contin-uously displayed.
Some must only be available should the need arise after an accident has occurred.
The primary effect of implementing the regulatory guide will be to insure that (1) the equipment used for indication of specific para-meters will survive the accident environment which could exist when the indication is required, and (2) sufficient redundancy of indications is provided to insure that required information will be available following single equipment failures (such as loss of a single power source).
The staff goal is that the Regulatory Guide 1.97 instrument displays, the Safety Parameter Display System, the control room design review, and the development of symptom oriented emergency operating procedures be integrated with respect to the overall enhancement of operator ability to comprehend plant conditions.
The specific concerns by Mr. Bender, directed at proposed instrumentation to detect inadequate core cooling, questioned the usefulness of proposed devices to provide information needed for the unambiguous interpretation of changing coolant inventory in the reactor core.
Mr. Lewis and Mr. Plesset cited their earlier comments on this matter which were included in the Committee's report, dated November 17,
- 1981, on St.
Lucie Unit 2.
They also emphasized their belief that proposed instrumentation does not provide an unambiguous measure of liquid level in the reactor vessel.
The staff review criteria for inadequate core cooling instrumentation are defined in NUREG-0737, Item II. F. 2, and are more extensive
- than, but inclusive of, the Regulatory Guide 1. 97 requirements.
The level monitoring systems are being evaluated as a component of an overall instrumentation system required to detect the approach to, existence of, and recovery from conditions of inadequate core cooling.
These level monitoring systems are intended to function under a variety of non-mechanistic accident conditions which are not necessarily anticipated but most certainly include dynamic effects associated with all small break LOCAs and TMI-2 type events.
The staff criteria specify that the integrated systems must provide an unambiguous, easy-to-interpret indication of inadequate core cooling (ICC).
The role of the level monitoring instrumentation is to provide information on system behavior during the interval between loss of subcooling margin and the occurrence of superheat indicative of partial core uncovery.
It also provides information to verify other indicators of coolant system and core cooling conditions during all phases of small break LOCA transients and during unpredictable scenarios involving multiple failures as occurred in the TMI-2 Palo Verde SSERl 18-3
accident.
The total information is useful as input for operator actions in response to the transient and for diagnosis of the event as input to short and long range operating decisions.
A response to the Committee comments on St.
Lucie 2 was provided in Supplement No.
1 to the St.
Lucie Unit 2 Safety Evaluation Report, NUREG-0843, dated
- December, 1981.
Since that earlier response, the staff has developed a plan to better articulate the design objectives of the ICC system and to better under-stand the industry approach to inadequate core cooling measurements.
This will increase assurance that some of the proposed devices provide an acceptable means to achieve the design objectives which appear to be shared by the Committee and the staff.
This plan is being initiated with an NRC/Industry meeting in February, at which the level measurement suppliers will present an assessment of their proposed instrumentation systems in relation to analyzed accident scenarios.
Utility representatives will be invited to participate.
These meetings are specifically designed to include the issues highlighted by the Committee.
Subsequent to the meeting with licensee/vendors, the staff will discuss its proposals with the NRC Committee to Review Generic Requirements (CRGR).
The CRGR will provide guidance to develop an NRC position with regard to implementation of ICC/water level instrumentation.
The staff expects that an agenda will then be established for detailed industry and staff presentations to the Committee.
Palo Verde SSER1 18-4
20 FINANCIAL QUALIFICATIONS In Section 20 of the SER, the. staff completed its evaluation of the financial qualifications of each applicant to operate PVNGS 1-3, shut down, if necessary, and maintain the facilities in a safe condition.
This evaluation was performed for the Arizona Public Service Company, Salt River Project Agricultural Improve-ment and Power District, Southern California Edison
- Company, Public Service Company of New Mexico, El Paso Electric Company, Los Angeles District of Mater and
- Power, and South California Public Power Authority.
Subsequently, by letter dated November 6, 1981, the application was amended to reflect a transfer by the El Paso Electric Company to the M-S-R Public Power Agency of a 3.95/o undivided ownership interest as a tenant in common with the other participants in PVNGS 1-3.*
The staff is currently evaluating the financial qualifications of the M-S-R Public Power Agency to assume its owner-ship interest and will report the results of that evaluation in a subsequent supplement to the SER.
"By separate letter, also dated November 6,
- 1981, the applicant has also filed a request for amendments to the construction permits requesting approval for this transfer of ownership interest.
This request for amendments to the construction permits is also in the review process at this time.
Palo Verde SSERl 20-1
22 TMI-2 REQUIREMENTS 22.2 Evaluation of TMI Re uirements I.A.l. 1 Shift Techincal Advisor In Section
- 22. 2 of the SER, the staff noted that the applicant had committed to provide additional information on the requalification training of Shift Technical Advisors (STAs) and any plans for the eventual phaseout of the STA program.
By letter dated December 3, 1981, the applicant described the proposed requalification training program for STAs, and stated that there was no current plan for phase out of the STA program.
Regarding the STA requalification program, the applicant stated that the program would be conducted on a two-year repeating cycle, commencing within six months of issuance of a plant operating license.
During this cycle, STAs shall attend and satisfactorily complete 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> of formal classroom retr'aining which will address transient and accident analysis and will emphasize the STA role in accident assessment.
In addition, during the requalification cycle, the STAs shall participate in.at least two weeks of simulator training (using the PVNGS plant specific simulator) where they shall act as STA during simulation of significant plant transients.
Based on the above information, and the evaluation presented in the SER, the staff finds the applicant's proposed program for training and requalification of STAs to be acceptable.
Therefore, this matter is now resolved.
I.D. 1 Control Room Desi n Review The SER stated that, with the exception of HED Item 102 and CLD Item 11.032, the staff found acceptable the corrective actions and implementation proposed by the applicant for each of the human engineering findings identified in the applicant's Preliminary Design Assessment report.
The SER also stated that the HFEB review team identified additional human engineering discrepancies in the staff's Control Room Design Review/Audit (CRDR/A) report which would be transmitted to the applicant.
Subsequently, the CRDR/A report was transmitted to the applicant.
By letters dated December 1, 10, and 16, 1981, the applicant provided its proposed disposition of each of the human engineering findings identified in the CRDR/A report.
The corrective actions and implementation schedules proposed by the applicant are acceptable to the HFEB staff.
As stated in the SER, there were a number of systems and items which were unavailable for review during the site visit either in whole or in part.
Using the guidance provided in Section 6 of NUREG-0700, "Guidelines for Control Room Design Reviews," the staff requires that the applicant perform an evaluation of those systems and items not reviewed, which are repeated below in more detail, and that the applicant submit its findings, proposed corrective actions, and schedule for implementing the actions.
The report of these items shall be Palo Verde SSERl 22-1
submitted for staff review and approval no later than 120 days before issuance of the operating license.
l.
A detailed comparison of the simulator with the Unit 1 control room could not be performed to identify all differences that might exist.
- 2. ~12
- Document organization and storage
- Spare parts, operating expendable and tools
- Supervisor access
- Non-essential personnel access 3.
Emer enc E ui ment
- Operator protective equipment
- Fire, radiation, and rescue equipment
- Emergency equipment storage 4.
Environment
- Temperature and humidity
- Ventilation
- Auditory
- Personal storage
- Ambience and confort 5.
The absence of documents made it impossible to evaluate consistency of procedure terminology with labels, displays, abbreviations, or document indexing and cross-referencing.
6.
Due to the existing state of the system, it was not possible to adequately evaluate all of the CRT displays for content and data presentation format.
7.
Lack of actual emergency gear prevented the evaluation of the operation of controls while wearing or using the emergency
- gear, or the availability of face masks with diaphragms capable of transmitting speech.
8.
The actual discernability and reliability of audio signals above ambient noise could not be measured.
Palo Verde SSER1 22-2
9.
The capability of complete internal and external'communications during emergencies (i.e., paging at the remote shutdown panel and/or direct communication with back panels, shift supervisor's office, etc.) could not be evaluated.
10.
Since only Panel B06 had color-shaded background panel sections, it was not possible to evaluate the effectiveness throughout the entire control room of the use of shading colors to identify groups of functionally related controls and displays.
11.
The proposed Plant Protection System logic alarm box on Panel B05 could not be evaluated because it is not yet installed.
12.
The out-of-service and temporary labeling systems had not been developed.
13.
The following instrumentation systems which HFEB typically reviews were not available.
(a)
In-core Thermocouple instrumentation displays, and (b)
Sub-Cooling monitor instrumentation displays.
Actions implemented to correct deficiencies will be audited by the NRC.
The results of this audit of the PVNGS Unit 1 control room will be reported in a subsequent supplement to the SER.
II.B.4 De raded Core - Trainin The following discussion supercedes the discussion presented in the Section 22 of the SER under Item II.B.4 and completes the review of this item.
The applicant has committed, in its letters dated October 28, 1981 and January 13, 1982, to utilize a degraded core training course.
The course covers 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br /> of training to control or mitigate an accident in which the core is severely damaged.
This material will be incorporated in the FSAR by an amendment from the applicant.
This course will be given to all shift technical advisors and operations personnel, from the Plant Manager to and including licensed operators, prior to fuel loading.
Managers and technicians in instrumentation and controls, health physics, and chemistry, will be given training commensurate with their responsibilities during accidents which involve severe core damage.
Based on this commitment, the staff has concluded that the Arizona Public Service Company has met staff requirements for training personnel in the use of installed plant systems to control or mitigate an accident in which the core is severely damaged.
This training program will be in general agreement with the "Training Guidelines for Recognizing and Mitigating the Consequences of Severe Core Damage,"
from the Institute of Nuclear Power Operations, Document Number STG-01, Revision 1, dated January 15, 1981.
The applicant has committed to complete, prior to fuel load, the trai ning of all operating personnel in the use of installed systems to monitor and control accidents in which the core may be severely damaged.
Palo Verde SSER1 22-3
Ti
APPENDIX A CONTINUATION OF CHRONOLOGY OF RADIOLOGICAL REVIEW October 8, 1981 October 14, 1981 October 14, 1981 October 15, 1981 October 15, 1981 October 19-20, 1981 October 20, 1981 October 20-21, 1981 October 21, 1981 October 21, 1981 October 22, 1981 October 22, 1981 October 23, 1981 October 23, 1981 October 23, 1981 Letter to applicant transmitting revision to September 2,
1981 letter concerning MEB draft SER.
Letter to applicant transmitting comments on Emergency Pl an.
Meeting with applicant to discuss its organization.
Letter from applicant forwarding clarification of design of radwaste management systems.
Letter from applicant concerning upcoming meeting to discuss implications of Standard Review Plan 13. l. 1.
Meeting with applicant to discuss sequencer test program.
Letter from applicant transmitting revised FSAR Figure
- 13. 1-7.
Meeting with applicant to discuss subsurface materials and to conduct foundations site visit.
Appeals meeting with applicant to discuss staff position on organization.
Letter from applicant transmitting revised response to staff question 471.17.
Meeting with applicant to discuss results of control room design audit.
Letter from applicant transmitting results from meeting held October 20-21, 1981.
Letter from applicant in response to August 21, 1981 letter.
Letter from applicant transmitting information regarding conflicting information in FSAR Chapters 13 and 17.
Letter from applicant forwarding revised response to question 222.02.
Palo Verde SSER1 A-1
October 27, 1981 Letter from applicant transmitting "Nuclear Operations, Activities."
October 27, 1981 Letter from applicant advising that responses to comments on emergency plan will be provided by February 26, 1982.
October 28, 1981 October 28, 1981 Letter to applicant forwarding the control room design review/audit report.
Letter from applicant transmitting revision to FSAR Section 13.2 and revisions to responses for Items I.A.3. 1 and II.B.4 of NUREG-0737.
October 28, 1981 Letter from applicant transmitting report, "Simulated Undervoltage Test of the IAV54E Relay and Analysis of the Operation of this Relay at the Proper Setting for the Events of the 'Voltage Regulation Study on ANPP.'"
October 28, 1981 October 28, 1981 October 28, 1981 Letter from applicant forwarding revised information on sampling parameters, secondary systems drain sampling, and other FSAR page changes.
Letter from applicant concerning design of diesel engine combustion air intake and exhaust system piping and components.
Letter to applicant forwarding list of questions from combined ACRS CESSAR/Palo Verde Subcommittee.
October 28, 1981 Letter from applicant forwarding revised responses to TMI Action Plan Items II.F. 1 and III.D.1. 1 (NUREG-0737).
November 2, 1981 November 2, 1981 Letter from applicant concerning compliance with guidelines for containment purge valve operability.
Letter from applicant transmitting Amendment 2 to Lessons Learned Implementation Report.
November 6, 1981 November 10, 1981 November 11, 1981 November 13, 1981 Letter from applicant requesting amendment to construction permits to add M -
S -
R Public Power Agency as additional applicant.
Letter to applicant regarding storage of low-level radioactive wastes at power reactor sites (Generic Letter 81-38).
Meeting with applicant to discuss on-site emergency pl an.
Letter from applicant transmitting Volume II of Containment System Independent Design Review.
Palo Verde SSER1 A-2
November 13, 1981 Letter from applicant transmitting "Field Pre-Service Inspection Program for the Arizona Nuclear Power project."
November 16, 1981 November 20, 1981 November 20, 1981 Issuance of Safety Evaluation Report.
Letter from applicant forwarding contingency plan and guard training and qualification plan.
Letter from applicant forwarding information on equi pment qual ificati on.
November 23-24, 1981 ACRS Subcommittee meeting with staff and applicant.
November 30, 1981 November 30, 1982 December 1, 1981 December 1,
1981 Letter to applicant concerning NRC volume reduction policy (Generic Letter 81-39).
Lettr from applicant transmitting responses to draft SER concerni ng emer gency pl arming.
Letter from applicant providing clarification concerning diesel storage tank corrosion protection.
Letter from applicant transmitting resolutions of staff concerns regarding the control room design review audit.
December 1,
1981 Letter from applicant providing justification for exceptions taken to Regulatory Guides 1.38 (Revisions 0
and 2).
December 1,
1981 December 2,
1981 Letter from applicant transmitting summary of November 19, 1981 meeting on preservice examination program.
Letter to applicant requesting verification that pertinent regulations in 10 CFR Parts 20, 50 and 100 are met.
December 3,
1981 Letter from applicant transmitting response to SER Chapter 7 Open Item and Chapter 7 Confirmatory Items 1-5.
December 3,
1981 December 3, 1981 December 4, 1981 Letter from applicant transmitting information concerning plans for Shift Technical Advisor requalification training and a description of long-term Shift Technical Advisor program.
Letter from applicant transmitting responses to open items regarding the sequencer test program.
Letter from applicant forwarding information concerning seismic and loss-of-coolant accident loads.
Palo Verde SSERl A-3
December 4, 1981 December 7,
1981 Letter from applicant forwarding "Foundation Instrument Report," concerning settlement monitoring, lateral pressures and liquification analysis.
Letter from applicant describing functional assignments for technical support center and mini-technical support center.
December 8, 1981 December 9,
1981 Meeting with applicant to hear applicant's appeals to staff positions on (1) tornado missile protection for spray nozzles and diesel generator fuel oil tank corrosion protection and (2) ultimate heat sink capacity.
Meeting with applicant to discuss control room design review.
December 10, 1981 December 10, 1981 December ll, 1981 Letter from applicant forwarding clarifications to resolutions submitted December 1, 1981 on control room design review.
ACRS meeting with staff and applicant.
Letter from applicant transmitting response to
'onfirmatory Item No.
22 in SER concerning spent fuel pool.
December 11, 1981 Letter from applicant forwarding Amendment No.
7 to FSAR.
December 15, 1981 December 16, 1981 Letter from Advisory Committee on Reactor Safeguards.
Letter from applicant providing clarification on human factors concern.
December 16, 1981 Letter to applicant regarding qualification of reactor operators
- license examinations (Generic Letter 81-40).
December 16, 1981 Letter from applicant requesting extension of submittal date for verification that pertinent regulations (10 CFR 20, 50 and 100) are met.
December 22, 1981 Letter from applicant advising that Unit 1 continues at accelerated pace and that the scheduled fuel load date is November 1982.
December 29, 1981 December 30, 1981 Letter from applicant transmitting response to Open Item No. 6, Preservice Examinat'ion Program.
Letter from applicant forwarding responses to questions concerning cable tray design.
Palo Verde SSER1 A-4
Oecember 31, 1981 Letter from applicant forwarding revised response to staff question concerning damping and shear modulus values used for seismic analysis.
January 5,
1982 January 8, 1982 Letter from applicant transmitting response to Open Item No. 8'n ECCS performance analysis.
Letter to applicant transmitting request for additional information.
January 12, 1982 January 19, 1982 January 21, 1982 January 25, 1982 January 25, 1982 Letter to applicant concerning long term operability of deep draft pumps.
Letter to applicant requesting additional information concerning setpoint acceptability.
\\
Letter from applicant transmitting additional information regarding interaction of the Corridor Building with Category I Structures.
Letter from applicant providing comments on Safety Evaluation Report (Sections 5.4.3, 6.2.4, and 9.5. 1.9).
Letter from applicant providing dates for submittal of information on equipment qualification.
Palo Verde SSER1
I
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Vlp
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++*++
APPENDIX B UNITEDSTATES NUCLEAR REGULATORY COMMISSION ADVISORYCOMMITTEEON REACTOR SAFEGUARDS WASHlNGTOM, O. C. 20555 December 15, 1981 Honorable Nunzio J. Pallahino Chairman U. S. Nuclear Regulatory Commission Mashington, DC 20555
SUBJECT:
ACRS REPORT ON THE PALO VERDE NUCLEAR GENERATING STATION mIITS 1, 2, AND 3
Dear Dr. Palladino:
During its 260th meeting, December 10-12, 1981, the Advisory Committee on Reactor Safeguards reviewed the application of the Arizona Public Service Company, the Salt River Project Agricultural Improvement and Power District, the El Paso Electric Company, the Public Service Company of New Mexico, and the Southern California Edison Company (Applicants) for a license to operate the Palo Verde Nuclear Generating Station Units 1, 2, and 3.
The joint applicants have designated the Arizona Public Service Company as the Project Manager and Operating Agent with full authority to construct and operate the power station.
The project was considered at a Subcommittee meeting in Phoenix, Arizona on November 23-24,
- 1981, and members of the Committee toured the facility on November 23, 1981.
In its review the Committee had the benefit of discussions with representatives of the Arizona Public Service
- Company, Combustion Engineering, Inc., Bechtel Power Corporation, the NRC Staff, and members of the public.
The Committee also had the benefit of the documents listed.
The Committee commented on the construction permit application for the Palo Verde Nuclear Generating Station Units 1, 2, and 3 in a report dated November 12, 1975 to the NRC Chairman.
The Palo Verde application is submitted in accordance with the Commission's regulations as described in Appendix 0 to Part 50, "Licensing of Production and Utilization Facilities," and Section 2.110 of Part 2, "Rules of Prac-tice," of Title 10 of the Code of Federal Regulations.
NRC policy stated in the Federal Register (42 FR 34395 and 43 FR 38954) allows for a reference system that involves an entire facility design or major fraction of a design outside the context of a license application.
For this application the reference system is the Combustion Engineering standard nuclear steam supply system known as its Standard Reference System 80.
This design has been reviewed by the ACRS and discussed in its report dated December 15, 1981, "Final Design Approval for Combustion Engineering, Inc. Standard Nuclear Steam Supply System (Standard Reference System 80)".
B-1
Honorable Nunzio J. Palladino 2
December 15, 1981 This power station is located in a sparsely populated section of Maricopa County, Arizona, about 36 miles west of the nearest boundary of Phoenix, Arizona.
The nearest densely populated center is Sun City, Arizona, about 35 miles east-northeast of the site, which had a 1980 population of about 57,800 persons.
.Palo Verde is the first commercial nuclear power station to be operated by Arizona Public Service Company and the first in the state of Arizona.
The Palo Verde Nuclear Generating Station uses three System 80 pressurized water nuclear steam supply systems designed by Combustion Engineering, Inc.
Each of these has a design core power output of 3800 MMt.
The turbine gen-erators are oriented so as to minimize plant damage should turbine failure occur.
The containment is a steel-lined, prestressed concrete cylindrical structure with a hemispherical dome and a design pressure of 60 psig.
The cooling tower makeup is supplied from treated sewage effluent from the city of Phoenix.
The Committee's review included consideration of the management organization and capability, and the operator training program.
The organizational plan for technical support of the operating plant is still being formulated.
The Committee notes that the Arizona Public Service Company management personnel have extensive experience in both commercial and other nuclear plant opera-tion and construction.
The utility anticipates using many of its installa-tion surveillance staff members as part of the technical support team.
The ACRS encourages this organizational arrangement, but believes the Applicant should promptly analyze the skill requirements needed to support operations and make certain that the necessary capabilities will be available when needed.
In order that the Committee be kept informed, we request an update on the organizational arrangement in about one year from this date.
The Committee notes that Arizona Public Service Company has a training simulator in operation at the Palo Verde site.
The Committee's review in-dicated that the training program is being developed and that use of the plant simulator is still in the process of being integrated into the pro-gram.
The Committee recommends that Arizona Public Service Company examine industry-sponsored programs concerning effective use of simulators for training and make certain that its approach takes account of current under-standing of simulator training limitations.
Discussion with the Arizona Public Service Company staff indicated that emergency operating procedures for dealing with off-normal plant behavior are incomplete.
Development of such procedures should be expedited to provide maximum time to make use of them in the operational training pro-gram.
In the Palo Verde design the primary system does not include capability for rapid, direct depressurization when the plant has been shut down.
This places extra importance on the reliability of the auxiliary feedwater 8-2
Honorable Nunzio J. Palladino 3
December 15, 1981 system and makes it necessary that the NRC Staff and the Applicant assure the availability and dependability of this system for a wide variety of transients.
It also places extra requirements on the continued integrity of the two steam generators as the only method of heat removal immediately after shutdown.
The ACRS recommends that the NRC Staff and the Arizona Public Service Company give additional attention to the matter of shutdown heat removal for Palo Verde and develop a detailed evaluation and justifi-cation for the position judged to be acceptable.
The Committee wishes to be kept informed.
Arizona Public Service Company should expand its studies on systems inter-actions and systems reliability.
A number of items have been identified as Outstanding
- Issues, Confirmatory
- Issues, and proposed License Conditions in the NRC Staff's Safety Evaluation Report dated November 1981.
The ACRS is satisfied with the progress on these topics and believes that they should be resolved in a manner satisfactory to the NRC Staff.
Our approval of the operation of this plant is contingent upon the satisfac-tory completion of construction and preoperational testing.
For this reason, we request that, prior to fuel loading on Unit 1, a report be provided to the Committee describing significant construction deficiencies and their disposi-tion, effectiveness of the quality assurance
- program, and results of the preoperational test program.
In addition, a review of the star tup experience on Unit I should be made prior to fuel loading on Unit 2 and the Committee kept informed.
We believe that if due consideration is given to the recommendations
- above, and subject to satisfactory completion of construction, staffing, and pre-operational testing, there is reasonable assurance that Palo Verde Nuclear Generating Station Units 1, 2, and 3 can each be operated at power levels up to the design core power output of 3800 MWt without undue risk to the health and safety of the public.
Additional comments by ACRS member M. Bender and ACRS members H.
W. Lewis and M. S. Plesset are presented below.
Sincerely yours, J.
Carson Mark Chairman Additional Comments by ACRS Member M. Bender The NRC requirements for instrumentation to follow the course of an acci-dent have been generally outlined in Regulatory Guide 1.97.
The ACRS has concentrated most of its attention on instrumentation to detect inadequate B-3
Honorabl.e Nunzio J. Palladino December 15, 1981 core coolinq, sometimes called pressure vessel coolant level measuring instrumentation.
The Regulatory Guide 1.97 requirements and the emphasis on measurement of vessel coolant levels both seem to have confused the real accident diagnosis requirements.
The proposed coolant level indicators could only have value under quiescent conditions.
The proposed devices, differential pressure indicators and
- heated junction thermocouples, require considerable information about hy-draulic conditions, pressure distribution, and density var iations in the pri-mary coolant circuit to be useful for unambiguous interpretation of changing coolant inventory in the reactor core.
A full understanding of mass and energy distribution and related physical behavior of the nuclear system would be needed to make such information diagnostically useful under most accident conditions.
The main value would appear to be for conditions where the system has been depressurized and the coolant state is known, for example, prior to refueling.
Such knowledge does not appear relevant to the circum-stances of primary concern such as accident conditions comparable to the TMI-2 event.
Regulatory Guide 1.97 has a mixture of requirements, some directed to pre-accident symptom identification, some to actual surveillance of rapidly changing transients, and some to surveillance of accident recuperation con-ditions.
Although all of these requirements could be justified under some circumstances, it is likely that, if everything listed in the guide were provided, the operators could be overwhelmed by the informational detail and their diagnostic capability actually impaired.
At a time when unambiguous accident diagnostic information is urgently needed, a maze of indicating and analytical devices that might confuse the operators hardly makes sense.
I propose the following criteria as a basis for determining accident diagnostics adequacy.
1.
Does the operator have a well-defined set of signals to guide his emergency response to important accidents7 2.
Do the emergency procedures enable the operator to avoid misinter-pretation of those signals under circumstances where accident diagnosis is needed in conjunction with emergency actions7 3.
In accident recovery is the sensor capability adequate to enable the operators to establish whether a stable and safe operating condition is being maintained until. the system can be brought to cold shutdown and reliable decay heat removal functions assured?
4.
If fuel failures occur, is there capability to determine whether the failures are of minor or major significance (clad reaction B-4
Honorable Nunzio J. Palladino Oecember 15, 1981 with water and fuel melting); whether bulk quantities of radioac-tive nuclides have been released to the primary coolant circuitry, the containment interior, or are leaking from containment; and whether the containment boundary is jeopardized by overpressure or overtemperature?
Only a few additions to the pre-TMI-accident instrumentation appear necessary to address these considerations.
However, to be certain that necessary in-for'mation is available, the actions required of operators during accidents must be thoroughly examined.
Emergency procedure guidance is now,being developed by the nuclear steam supply equipment vendors.
This guidance must be converted into usable procedures that may be testable on nuclear plant simulators.
Palo Verde and a few other installations have simulators that might be used for this purpose.
Those operating organizations having appro-priate simulat'ion equipment should give priority attention to proving the effectiveness of the diagnostic equipment in conjunction with proposed emergency procedures in order to verify diagnostic adequacy.
No serious effort in this direction appears to have been initiated up to this time.
Additional Comments by ACRS Members H.
W. Lewis and M. S. Plesset We do not wish to belabor the points we made in our addendum to the ACRS letter dated November 17, 1981 on the St. Lucie Plant Unit 2, but they are as relevant here as there.
The Staff continues to accept instruments that do not provide an unambiguous measure of liquid level in the pressure
- vessel, and continues to lack an adequate rationale therefor.
We do not find fault with the Applicants for their efforts to be responsive to the Staff, but are concerned about the proliferation of inadequately considered requirements, of which this is only one example.
To sanctify an ambiguous indication of core water level is to play with fire.
In this particular case (heated thermo-couples in a separator tube), not only dynamic effects, but a pressure vessel full of high-void-fraction water will spoof the instrument, and tend to lull the operator into a false sense of security about the coolant inventory.
In that specific case, the instrument will indicate that the vessel is nearly ful 1.
None of the above is meant to suggest that we oppose the provision of instrumentation to follow the course of. an accident or to detect the onset of inadequate core cooling - unambiguous diagnosis of accident conditions through improved instrumentation and training is a high priority.
Our concern is a piecemeal and incoherent approach to the problem, as exemplified here.
References:
1 1
C p,"
1 1
1 1
1 1
Final Safety Analysis Report," with Amendments 1 through 6.
2.
U.S-Nuclear Regulatory Commission, "Safety Evaluation Report Related to the Operation of Palo Verde Nuclear Generating Station, Units 1, 2, and 3," NUREG-0850, dated November 1981.
B-5
Honorable Nunzio J. Palladino December 15, 1981 3.
Combustion Engineering, Inc., "System 80 CESSAR FSAR," with Amendments 1 through 5.
4.
U.S. Nuclear Regulatory Commission, "Safety Evaluation Report Re-lated to the Final Design of the Standard Nuclear Steam Supply Reference System CESSAR System 80," NUREG-0852, dated November 1981.
B-6
APPENDIX C EMERGENCY PREPAREDNESS EVALUATION REPORT BY THE DIVISION OF, EMERGENCY PREPAREDNESS OFFICE OF INSPECTION AND ENfORCEMENT U. S.
NUCLEAR REGULATORY COMMISSION IN THE MATTER OF PALO VERDE NUCLEAR GENERATING STATION DOCKET, NOS.
50-528,
- 529, 530 DECEMBER 1981
INTRODUCTION The Arizona Public Service Company (APS) filed with the Nuclear Regulatory Commission a revision to,the Palo Verde Emergency
- Plan, dated April, 1981.
In addition the Arizona Public Service Company filed the following supporting information:
Date
~Sub ect 09-03-81 A brief description of the prompt notification system 08-03-81 08-28-81 Lesson learned implementation (emergency centers and meteorological instrumentation requirements)
Evacuation time estimates The plan and above listed submittals were reviewed against the requirements in Sections
- 50. 33 and 50. 47 of 10 CFR Part 50, the requirements in Appendix E to 10 CfR Part 50, and the guidance evaluation criteria in NUREG-0654/fEMA-REP-l, Revision 1, entitled "Criteria for Preparation and Evaluation of Radiological Emergency
Response
Plans and Preparedness in Support of Nuclear Power Plants,"
November 1980.
The results of the staff review were forwarded to the applicant by letter dated October 14, 1981, which identified a number of deficiencies.
The staff conducted a site visit on November 10, 1981 and held a meeting with the applicant on November 11, 1981 at which the preliminary review results were discussed.
As a result, the applicant filed on November 30, 1981, acceptable commitments that addressed the deficiencies identified during the preliminary review.
This evaluation report follows the format of Part II of NUREG-0654,in that for each of the planning standards is listed (1) a summary of the applicable por-tions of the revised Emer g'ency Plan, as supplemented by the above listed sub-mittals and (2) a list of the commitments made by the applicant in its November 30, 1981 letter to resolve the deficiencies noted by the staff's review.
The final section of this report provides the staff's conclusions.
EVALUATION A.
ASSIGNMENT OF RESPONSIBILITY (ORGANIZATIONAL CONTROL)
The Federal, State and local organizations that are intended to be part of the overall response organization for the Emergency Planning Zones are identified.
The State of Arizona, through the Division of Emergency Services (ADES),
has responsibility for control of offsite actions during a radiological emergency.
Arizona law makes the Arizona StateEmergency Plan binding on other Arizona governmental agencies and, therefore, it is cited in lieu of separate letters of agreement.
The concept of operations for discharge of offsite responsibilities together with a discussion of the responsibilities assigned to various State and local governmental agencies will be provided in Annex L of the Arizona Emergency Plan.
In
- addition, a figure that specifies the principal operations and technical responsibilities for various offsite emergency functions is provided.
The onsite concept of operations is described and block diagrams showing the interfaces between and among the principal response centers are pro-vided.
However, the interfaces and relationship of the applicant's emer-gency response with offsite agencies is not clearly described and is not shown on the block diagrams.
In particular, the interfaces in the follow-ing areas are not sufficiently described:
protective action decision making, offsite monitoring and coordination of results, processing of plant evacuees, and plant access control.
The Emergency Coordinator is identified as the individual who shall be in charge of the initial onsite emergency response.
For an Alert or higher level of emergency, the TSC Director or Emergency Operations Director will assume overall control when the EOF is activated.
Primary and alternate lead personnel will be designated to assure continuous 24-hour operation.
The Emergency Operations Director is respon-sible for ensuring continuity of resources.
However, the individual with responsibility for 24-hour manpower planning and resources for this effort is not identified.
The applicant's November 30, l981 submittal committed to provide the following in a revised version of the plan to correct noted deficiencies:
2.
3.
a copy of the Arizona State Emergency Plan.
A final Plan that has been endorsed by all the State and local governmental agencies with primary roles must be provided before a favorable finding can be made; a discussion of how the exact boundaries of the plume EPZ were established; a discussion of the onsite/offsite concept of operation to demonstrate the interfaces and coordination in the following areas:
protective action decision making, offsite monitoring and result coordination, C-2
control and processing of plant evacuees, and plant access control if plant evacuation is ordered; 4.
identification of each step in the protective action decision making and implementation process (e. g., siren activation/EBS message) to demonstrate 24-hour capabilities; 5.
identification of who is responsible for 24-hour APS manpower planning and the resources to be used for this effort; and 6.
current letters of agreement will be provide prior to fuel loading of Unit 1.
(Agreements with all local support agencies to include ambulance, and hospital support or any other group not addressed by the State plan must be provided before a favorable finding can be made.
All agreements should be reviewed and certified current, and dated within one year of anticipated license issuance).
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B.
ONSITE EMERGENCY ORGANIZATION The Duty Officer is designated as the Emergency Coordinator, and is on-shift at all times and possesses the required authority and responsibility.
A line of succession for the Emergency Coordinator is identified.
The functional responsibilities of the Emergency Coordinator are-established.
The Emergency Coordinator may not delegate the responsi-bility for the decision to notify offsite officials or for recommending protective actions.
'The interfaces among the onsite emergency functions and emergency response centers are illustrated in block diagrams.
However, the interfaces with offsite support and corporate level support is not shown and the role of the Satellite TSC is not discussed.
Corporate support will be coordinated through the Corporate Emergency Center (CEC).
The plan specified the corporate management responsible for:
techni cal suppor t, logistics support, media interface, and overall corporate management.
However, the corporate level point of contact with offsite agencies is not clearly designated.
The applicant's November 30, 1981 submittal committed to provide the following in a revised version of the plan to correct noted deficiencies:
1.
the relationship of the normal shift organization to the emergency functions identified in NUREG-0654, Table B-1; 2.
the criteria for transfer of the responsibilities of the Emergency Coordinator to the Emergency Operations Director; 3.
the normal and emergency position titles of those personnel, to the working
- level, who will perform the functions and tasks identified in NUREG-0654, Table B-j. to include who will act as communicator during the backshift.
In
- addition, a description of the methods and compensatory measures used to meet the 30-minute and 60-minute staffing requirements will be provided (A
description of how the augmentation requirements will be demonstrated along with how this capability will be maintained must also be provided);
4.
block diagrams showing the interfaces with corporate, local support (fire, medical, etc.),
vendor and other offsite support agencies and a discussion of how technical support will be provided on a timely basis and interfaced with onsite efforts; C-4
5.
specification of the contractor support used to augment the licensee emergency response; letters of agreement with all local, vendor, or contractor (e. g.,
- NSSS, INPO) support that provide the information set forth in NUREG-0654, Criterion 8-9.
(This must demonstrate how offsite technical support will be coordinated);
7.
sufficient detail to assure that events can be classified and offsite officials notified and provided with a protective action recommendation within 15 minutes of the fALs being exceeded (The, plan did not specify who had the authority and responsibility to classify events.
In addition,
=-the fact that the Duty Officer may not be at the affected unit may delay this process);
8.
a discussion of the role, manning and interface with other centers of the Satellite TSC;. and 9.
a listing of the assessment,
- sampling, survey and other HP/Chemistry Technician tasks that may be required during the initial phase of an accident along with a discussion of the conditions that will be considered (e.g.,
rapid offsite survey if instruments are inoperable) when determining what actions to take.
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C.
EHERGENCY RESPONSE SUPPORT AND RESOURCES Request for Federal assistance will be channeled through the Arizona Division of Emergency Services (ADES) or the Arizona Radiation Regulatory Agency (ARRA).
The exception to this is. direct contact between the plant and NRC.
The plan summarizes the Federal resources available and the responsibilities of the NRC and FEHA in accordance with the National Radiological Emergency Preparedness/Response Plan for Commercial Nuclear Power Plant Accidents (Haster Plan).
However, the actual method of requesting Federal Assistance is not specified or supported by a letter of agreement with the Federal point of contact (DOE).
A licensee representative will be part of the staffing of the State EOC.
The applicant's November 30, 1981 submittal committed to provide the following in a revised version of the plan to correct noted deficiencies:
1.
the expected response times of various types of Federal assistance; 2.
a summary of the possible resources that could be used to support a Federal response in accordance with NUREG-0654, Criterion C. l.c; 3.
the specific Federal points of contact for requesting various types of assistance.
(This will be supported with letters of agreement prior to Unit l fuel loading);
and 4.
a description of the onsite radiological laboratories, their capabilities, availability during accident conditions and a description of offsite laboratory support.
D.
EMERGENCY CLASSIFICATION SYSTEM The plan has established four classes of emergencies compatible with those in Appendix j. to NUREG-0654.
The plan has Emergency Action Levels (EALs) for most of the specific initiating conditions listed in Appendix 1 to NUREG-0654.
However, in most cases, the EALs are'ot specific instrument readings or other explic-itly observable/measurable indicators.
~Exam le 1
The EAL for "loss of 2 of 3 fission product barriers with potential loss of the 3rd" is "as evaluated - based upon consideration of coolant activity, coolant inventory and makeup rate, and containment pressure".
Discussion Specific indicators based on containment pressure, temperature, radiation
- levels, core temperature or water levels must be established for Control Room instrumentation to indicate loss of the first two fission product barriers.
The EAL for potential loss of the containment must include specific observable levels for all failure modes.
Mode EAL TYAe overpressure steam explosion.
failure to isolate
. containment pressure level very high radiation levels control room indicator lights or out-o'-plant radiation levels The applicant's November 30, 1981 submittal committed to provide the following in a revised plan to correct noted deficiencies:
l.
emergency Action Levels (EALs) for each initiating condition specified in Appendix j. to NUREG-0654 (The EALs will be observables I'e. g., instrument
- readings, equipment status indications, alarm annunciators]
which are both necessary and sufficient to explicitly character'ize each initiating condi-tion to the extent possible.
These must include the instruments used to identify inadequate core cooling [water level, subcooling, exit core temperature]);
and 2.
a discussion of how the EALs will be integrated into Control Room operator response to demonstrate that'a system has been established that will allow rapid and accurate classification of events (This must include the job aids to be used and how the system wi 11 be tested).
C-7
E.
NOTIFICATION METHODS AND PROCEDURES Procedures for notification of response organizations, including the means for notification, are established to include provisions for verification by an authentication code.
The contents of the initial emergency notification messages have been established and contain the required information.
However, provisions for protective actions are tied only to radioactive releases.
In a letter dated September 3, 1981, the applicant described briefly the public warning system.
It will contain a network of 36 sirens located to provide a minimum of 60 dB(c) sound level to the populated area within the exposure pathway plume EPZ.
The applicant will attempt to have the system available for the first joint exercise tentatively scheduled for September 1982.
However, neither the plan nor the letter describe the "administra-tive" means for warning system activation and for providing a public mes-sage over the Emergency Broadcast System (EBS).
The plan should include a
complete description of the protective action decision making process to demonstrate the capability to make protective action decisions within 15 minutes of receipt of a protective action recommendation from the plant on a 24-hour-a-day basis.
In addition, the plan does not demonstrate that the warning system signal and public information messages can be provided within 15 minutes of making of the protective action decision by offsite officials.
F The applicant's November 30, 1981 submittal committed to provide the following in a revised version of the plan to correct noted deficiencies:
a complete description of the administrative and physical means for prompt alerting and notification of the public within the plume exposure pathway EPZ (Sufficient detail will be provided to allow for evaluation against the criteria set forth in Appendix 3 to NUREG-0654.
This should include sufficient detail to demonstrate that protective action decisions and public notification can be made in accordance with time standards speci-fied in 10 CFR Part 50, Appendix E, IV.D.3);
copies of the prearranged messages to be used to inform the public of their expected response that cover a range of protective measures as set forth in NUREG-0654, Criterion E.7; 3.
a revision of the initial notification messages to provide for recommending protective. measures based on plant status (in addition to recommendations based on projected doses) and using a key-hole approach; and APS will provide a letter. to the NRC committing to having the physical and administrative means for notification in place before Unit 1 fuel loading and will also provide a schedule for installation and testing.
C-8
F.
EMERGENCY COMMUNICATIONS Communications with contiguous State/local governments within the EPZs are provided by a dedicated telephone system backed up by the National Warning System.
The system is further supplemented by a radio/microwave system to the State.
However, i't is not clearly demonstrated that all those involved in protective action decision making and implementation are provided with a backup system that will operate if the telephone system and normal power are lost.
I The provisions for communications with the NRC, emergency centers and monitoring teams are described but not in sufficient detai 1 to demonstrate compliance with the requirements of NUREG-0696.
Communication between medical transportation and the site and State EOC is provided.
Periodic tests of the communication systems will be conducted The applicant's November 30, 1981 submittal committed to provide the following in a revised plan to correct noted deficiencies:
1.
the organizational titles and alternates for both ends of the communication links which would be involved in initiating emergency response
- actions, and assurance that such stations will be manned 24-hours per day; 2.
a diverse means of communication which will operate with loss of normal power between the site and the offsite response agencies responsible for emergency response actions (since those individuals responsible for protec-tive action decision making and implementation are not identified, it is not clear that adequate backup communications have been provided);
and 3.
a further description of onsite communications in and between the emergency
(This should include a
description of the emergency communications system used to coordinate repair/corrective actions on a continuous basis [e.g.
an in-plant radio system] to demonstrate compliance with the requirements of NUREG-0696).
C-9
G.
PUBLIC EDUCATION AND INFORMATION Coordinated annual dissemination of an information brochure to the public regarding how they will be notified, and what their actions should be in any emergency, contacts for further information, general information on radiation and means of advising officials of. special needs is provided.
An Emergency News Center (ENC) is provided at the Palo Verde Inn.
Provisions for information flow to the Emergency News Center from the EOF is provided.
A backup ENC, which may be used to communicate information to the media, is located above the EOF.
Procedures in use at the ENC provide for coordination with governmental entities in dealing with rumors.
Programs will be conducted to acquaint news media representatives with the emergency
- plans, information concerning radiation, and points of contact for release of public information in an emergency.
The applicant's November 30, 1981 submittal committed to provide the following information in a revised plan to correct noted deficiencies:
1.
identification of what advance arrangements have or will be made for assuring that the Palo Verde Inn will be available and has the capability for use as the Emergency News Center; 2.
a further description of the emergency information brochure which will be provided for both the resident and transient population around the site to demonstrate that the information specified in NUREG-0654, Criterion G. 1 will be provided (in addition, a discussion of the relationship between accident conditions and why various protective actions may be recommended should be included in the brochure);
3.
a description of the provisions to supply the public information in a form that wi 11 be available when needed; 4.
a description of the provision to reach transients to include motels and the construction workers at the other units.
(This shall include posting of information); and 5.
identity of the licensee spokesman at the ENC and how news media releases will be coordinated with State officials.
C-10
H.
EMERGENCY FACILITIES AND E UIPMENT A central Technical Support Center (TSC) located below grade that has adequate shielding and ventilation to ensure habitability during Design Basis Accidents is provided.
This location is not within 2 minutes of the Control Room of each unit.
The TSC will have a Safety. Display System readout for each of the three units as well as the ability to call up plant parameters as specified in Regulatory Guide 1.97.
Each unit has a
Satellite TSC area immediately adjacent to its Control Room with capability for plant parameter display.
The TSC has its own computer tied to each unit that can access plant parameters, meteorological data and effluent monitoring system.
The TSC has communications with CR,
The TSC is the focal point for onsite emergency response and for assisting the Control Room during an emergency.
The TSC description in the plan and the additional information provided in the August 3, 1981 applicant sub-mittal on NUREG-0737 items did not provide sufficient detail to demonstrate compliance with NUREG-0696.
The Control Room has direct access to the Satellite TSC which includes a
terminal for the TSC computer to expedite initial offsite dose projections and to allow initial direction of emergency response prior to activation of the full emergency organization.
An Emergency Operation Facility (EOF) has been established below grade with adequate shielding and ventilation to ensure habitability during Design Basis Accidents.
The functions performed at the EOF are:
making protective action recommendations, liaison with offsite agencies, overall management of site response, and receipt and analysis of field monitoring data.
The EOF has communications with the Control
- Room, TSC,
- ENC, State
The EOF. has access to the Regulatory Guide 1.97 parameters via a CRT terminal and to an automated real-time offsite dose projection system.
The EOF description in the plan and the additional information provided in the August 3, 1981 applicant submittal on NUREG-0737 items did not provide sufficient detail to demonstrate com-pliance with NUREG-0696.
An Operational Support Center (OSC) is located in the lunch room of each unit's auxiliary building.
Emergency supplies are stored in kits adjacent to the OSC.
The OSC has communicators with the TSC and the Control Room.
The location of the alternate OSC was not specified.
For alerts and more severe classifications, onsite and offsite emergency centers wi 11 be manned and activated.
The emergency organization will be fully manned within 90 minutes.
The plan describes monitoring systems for geophysical, radiological and fire monitoring.
However, since the EALs are not complete, the process monitors used to classify an event cannot be evaluated.
The plan must
specify the range of the monitors used as EAL indicators to allow for a determination if they are compatible with the specific EALs.
The licensee has made provisions to acquire data from an TLD monitoring system.
In addition, portable field equipment for field assessment is provided.
The plan and applicant's submittal of August 3, 1981 on NUREG-0737 items describe provisions for a meteorological system designed to meet the requirements of NUREG-0654 that consisted of 2 towers that provide wind direction, windspeed, dewpoint, precipitation, temperature, and tempera-ture differential with height.
These parameters will read out in the CR,
The meteorological system described in the plan and in the applicant's August 3, 1981 NUREG-0737 submittal is currently undergoing review for compliance with NUREG-0654, Appendix 2.
quarterly inspections of the operational readiness of emergency equipment will be conducted.
Sufficient reserves of instruments/equipment will be maintained to replace those undergoing calibration or repair.
Calibration will be conducted in accordance with technical specifications.
A general description of the emergency supplies and equipment is provided.
However, it is not clear where the supplies are located, quantities pro-
- vided, and how the supplies are divided into kits.
The applicant's'ovember 30, 1981 submittal committed to provide the following in a revised plan to correct noted deficiencies:
sufficient detail to demonstrate that the EOF and TSC are compatible with the guidance of NUREG-0696 and a description of the concept of operation of the Satellite TSC and how it will be integrated into,the site response (This should include a description of the contingency arrangements to be made if coordination of offsite monitoring and coordination with offsite officials must be transferred from the EOF due to radiological conditions);
activation of the TSC within 30 minutes, the expected time and criteria for flow of responsibility to the EOF and description of who will report to these facilities (this must demonstrate the flow of responsibility/
functions is in accordance with NUREG-0696);
3.
a complete description of the meteorological measurement system for evaluation against the criteria set forth in Appendix 2 of NUREG-0654, and a schedule for meeting 'the milestones specified in Annex 1 to that Appendix; identification of the onsite radiological and process monitoring systems in accordance with criteria H.5.a and b of NUREG-0654.
(The monitors identified must include those used for obtaining Emergency Action Levels for the appropriate initiating conditions listed in Appendix 1 to NUREG-0654.
This must also include instrument identification, readout
- location, and range);
sufficient information to establish that the offsite dosimetry system will meet the requirements of the NRC Radiological Assessment Branch Technical
Position for the Environmental Radiological Monitoring Program to include the location of the TLDs stations.
6.
a description of the kits to be provided at each emergency center and their use to demonstrate that sufficient supplies and equipment have been provided to support the'mergency response actions to be performed -from those locations; and 7.
a description of the provisions for an alternate OSC.
C-13
I.
ACCIDENT ASSESSMENT The plan established methods and techniques used to determine the magnitude of the release of radioactive materials.
A Meteorological data will be accessed in the
Meteorological data will be provided to State officials by voice link.
If the effluent monitors are off scale or inoperable, estimates of the release rate is obtained from prepared tables based on 10 CFR 100 accident source terms.
These estimates will be confirmed by field monitoring teams The plan states that monitoring equipment has the capability to detect, under field conditions, radioiodine concentrations as low as 10-7 mCi/cc;
- however, the system is not described or the location of this equipment specified.
The plan describes a means for relating iodine and noble gas concentration to dose rate and comparing these to the EPA PAGs.
The applicant's November 30, 1981 submittal committed to provide the following in a revised plan to correct noted deficiencies:
specific EALs and a description of the assessment systems to demonstrate compliance with NUREG-0654, Criterion I. 1; 2.
3.
sufficient detail to enable evaluation against NUREG-0737, items II.B.3, II.F. 1 and III.D.3.3 as they relate to emergency response to include how these results will be used in dose projections, determination of core condition, and corrective action determination to include how these tasks will be performed and integrated into the total response (This should include response times, criteria for performing, training, and drills);
a method for relating source term to offsite doses based on containment leakage (This should include plots which show the containment radiation monitor reading vs.
time following an accident for release of gap activity and release of fuel inventory.
These should be used as part of the EALs);
5.
6.
the basis for the models used to relate radioiodine and noble gas concentration to dose and to relate monitor reading to release rate (e.g.,
how was change in mix as a function of times considered)
(In addition a
description of the transport model used in dose projection should be provided);
provisions for determining containment radiation levels if the containment monitor is inoperable; a method which may be used to relate effluent monitor readings to offsite doses various conditions (This should include the job aids
- used, such as tables and/or monograms);
a description of the provision for field monitoring to provide the information required by NUREG-0654, Criteria 1.7 and 1.8; a description of the means used to confirm the actual release mix if the effluent monitors are offscale or inoperable to include provision for rapid onsite and offsite field surveys (The criteria for dispatch of teams should be discussed);
and the location of the job aids used for the backup means of dose projection.
C-15
J.
PROTECTIVE
RESPONSE
0
. The means used to warn or advise onsite personnel have been established using an installed evacuation alarm and public address system.
Radiological monitoring will be provided for evacuating onsite personnel at the Security Building.
In the event onsite radiation levels make a
rapid site evacuation advisable provisions have been made for offsite monitoring/decontamination 2.5 miles from the site.
However, there are no provisions for rapid site evacuation based on plant status nor provisions for determining the habitability of the offsite monitoring location.
There are no provisions for decontamination of plant evacuees offsite.
If conditions are such that the TSC Director deems it appropriate, nonessential personnel will be evacuated.
The plan does not state that these personnel will be evacuated for Site and General Emergencies.
The plan has specified that accountability will be accomplished within 30 minutes for the normal staff.
However, accountability for construction workers at the other units is not discussed.
Respiratory protection, protective equipment and radioprotective drugs will be available on entry to the station and at a central location.
In addition, protective. equipment is located throughout the site.
The plan has established provisions for recommending protective action based on projected doses.
However, the criteria for recommending protec-tive actions based on plant (core/containment) conditions are not provided.
In addition no provisions are included for consideration of offsite condi-tions such as evacuation times or availability of shelter.
Also, the recommendations provided are only for the downwind direction.
The plan did not contain evacuation time estimates.
However, the licensee submitted on August 28, 1981 the good weather evacuation time analysis from the Draft Arizona State Emergency Plan.
This study is being evaluated for compliance with criteria in Appendix 4 of NUREG-0654.
The plan contained maps showing the population distribution around the site.
The applicant's November 30, 1981 submittal committed to provide the following in a revised plan to correct noted deficiencies:
2.
provision for notification and accountability of the construction workers at the units under construction and a discussion of notification of any population in the other owner controlled areas; provisions for site evacuation signs and site evacuation maps; 3.
maps showing the evacuation routes to be used by persons on-site to include the construction workers at the units under construction;
4.
a discussion of the provisions for distribution of KI to all onsite emergency workers, the criteria for its use and who has the authority to issue it; 6.
a mechanism for recommending protective actions based on plant (core/
containment conditions) that.takes into consideration offsite conditions that may influence the choice of the most effective protective action (e. g., evacuation time of the construction workers) and that Uses of the key-hole approach (i.e., not limiting protective actions to only the downwind direction) (This should include recommending evacuation for 2 miles around the site and 5 miles downwind under core melt conditions and plume EPZ sheltering for all General Emergencies.
This mechanism should be reviewed by local officials);
evacuation time estimates that are integrated into the protective action decision making mechanism (These estimates should include times under rain conditions and a discussion of the observed daily departure times of the construction workers.
In addition how the special traffic control measures used during the daily departure of the construction workers will be imple-mented during an emergency should be included);
7.
readable maps that supply the information specified in NUREG-0654, Criteria 10.a., b., c.;
8.
an estimate of the protection afforded by local structures to include the plant construction sites that are integrated into the protective action recommendation mechanism; 9.
the criteria for immediate evacuation of the site to an offsite monitoring point that includes plant (core/containment) conditions, how evacuees will be directed to the offsite monitoring point, a discussion of the traffic control impact (construction worker evacuation) of monitoring plant person-
- nel, and the provisions for evacuation even further from the site if conditions warrant; and 10.
a further description of the onsite protective equipment and supplies to demonstrate sufficient respiratory protection and instruments to support the required emergency function such as repair/corrective
- actions, monitor-ing, movement between centers and reentry.
C-17
K.
RADIOLOGICAL EXPOSURE CONTROL Exposure guidelines consistent with EPA standards have been established for:
emergency duties not related to protecting equipment, personnel or the public; preventing extensive equipment
- damage, further escape of effluents or controlling fires; and livesaving missions, e.g.,
search and rescue, or to prevent.conditions that would injure persons.
The Supervisory Radiation Physicist may authorize emergency workers to receive doses in excess of 10 CFR 20 occupational limits.
However, it is not clear he/she is onsite 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />s-a-day.
The Radiation Exposure and Maintenance (RE&M) System is accessible in the emergency centers and provides 24-hour-a-day capability to determine the doses received by emergency workers.
Action levels for determining the need for decontamination are specified.
Liquid waste from decontamination will be collected and handled in accord-ance with the Station Manual.
Provisions for onsite contamination control are provided.
The applicant's November 30, 1981 submittal committed to provide the following in a revised plan to correct noted deficiencies:
a description of how the onsite emergency radiation protection program will differ from the normal program (e.g.,
use of high range instruments, automatic use of respiratory protection during entries, emergency
- RWPs, etc.) (In addition, the provisions to be used to reduce the risk associated with emergency operations
[e. g.
as briefings and practice runs at other units if time permits] should be discussed along with how the emergency radiation protection procedures will be implemented.
The identity of who on the backshift authorizes exceeding 10 CFR'0 limits should be specified);
2.
capability for decontaminating relocated onsite personnel in accordance with Criter ion K. 7 of NUREG-0654; 3.
a description of the "job" TLD system and its utilization during emergencies including location of readers, provisions for backup capabilities, and recordkeeping if the RE&M system is lost; and 4.
a discussion of onsite contamination control, measures relating to drinking water and food supplies.
C-18
L.
MEDICAL AND PUBLIC HEALTH SUPPORT Arrangements have been made for a local hospital and medical transportation having the appropriate capabilities; however, letters of agreement are not provided.
Onsite first aid capability is provided.
The applicant's November 30, 1981 submittal committed to provide the following in a revised plan to correct noted deficiencies:
1.
arrangements for a backup hospital with the capability for long-term treatment of contaminated individuals; and 2.
letters of agreement in accordance with NUREG-0654, Criterion B.9.
C-19
N.
RECOVERY AND RE-ENTRY PLANNING AND POST-ACCIDENT. OPERATIONS General plans and procedures for recovery are developed The general structure, functions and membership of the facility recovery organization are provided.
The applicant's November 30, 1981 submittal committed to provide the following in a revised plan to correct noted deficiencies:
1.
provisions to re-enter following a site evacuation including where personnel will obtain necessary instruments and protective equipment; 2.
the criteria for downgrading the level of emergency and a description of the decision making process; 3.
provision to inform all parties that the "Recovery" phase has been entered; and 4.
provisions for estimating total population dose.
C-20
N.
EXERCISES AND DRILLS An annual joint exercise involving the State, local and station personnel will be conducted.
A critique of the annual exercise by Federal, State and local observers is provided.
The scenarios of the exercises are varied from,year to year such that all major elements of the Emergency
Response
Plan and its procedures are evaluated every five years.
Additionally, exercises may be conducted under various weather conditions, at differing times of day, and some are unannounced.
Required fire drills, communication drills, medical emergency drills, radiological monitoring health physics drills are provided.
Exercise scenarios include the required information.
A critique by government observers, resulting in a formal evaluation, is provided.
Observer comments are discussed in a post-exercise critique, with deficiencies being identified and proper corrective action determined by responsible plant staff.
The Emergency Planning Coordinator is responsible for implementation of corrective action identified as part of the annual exercises'."
The applicant's November 30, 1981 submittal committed to provide the following in a revised plan to correct noted deficiencies:
l.
monthly testing of the communications with the NRC in accordance with paragraph E.9.d of Appendix E to 10 CFR Part 50; 2.
a discussion of the mix of structured and less structured aspects of the program which will allow free play in decision 'making during exercises; and 3.
provisions for licensee observers, critiques following exercises and evaluation of their comments.
C-21
0.
RADIOLOGICAL EMERGENCY RESPONSE TRAINING Site-specific training is offered to offsite emergency organizations par ticul arly the Sheri ff' depar tment, and the Arizona Radi ati on Regul atory Agency.
Onsite emergency groups receive annual specialized training.
A specialized training program including initial and periodic retraining sessions has been established for the following organizational categories:
a) directors/coordinators; b) personnel responsible for accident assess-ment; c) fire and hazards control teams; d) repair and damage control; e) first aid and rescue; f) medical support personnel; g) licensee head-quarters support personnel; and h) personnel responsible for communications.
The applicant's November 30, 1981,submittal committed to provide the following in a revised plan to correct noted deficiencies:
l.
a description of how those officials responsible for offsite protective action decision making will receive training on protective measures and their relationship to plan conditions; confirmation that all personnel who perform emergency functions, duties or tasks will receive annual training; 3.
4.
a descrip ion o ow t
- f. h each person who is assigned a non-trivial emergency te task that is not part of their normal duties will annually-demonstra e
their ability to perform these tasks during a practical exercise; and a description of how the training program will be documented (e.g.,
lesson
- plans, performance objectives, etc.).
C-22
P.
RESPONSIBILITY FOR THE PLANNING EFFORT:
DEVELOPMENT PERIODIC REVIEW AND DISTRIBUTION OF EMERGENCY PLANS The Corporate Health Physicist and Emergency Planner is assigned as the Corporate Emergency Planning Coordinator;
- however, the onsite individual with this responsibility has not been identified.
The Vice President of Electric Operations has overall responsibility for emergency response planning.
Periodic revision of the plan, as
- needed, including changes identified by drills and exercises, is provided.
The revision and updating of plans is in accordance with the Station Manual which provides for physically making of changes to the text and distributing changes to all plan holders.
A listing of.supporting plans is given.
Appropriate independent annual audit of the emergency preparedness program are provided.
Telephone numbers and responsible contacts in the emergency procedures will be updated quarterly.
The applicant's November 30, 1981 submittal committed to provide the following in a revised plan to correct noted deficiencies:
1.
a listing of the procedures required to implement the plan in accordance with NUREG-0654, Criterion P.7; 2.
provisions for training the individuals responsible for the emergency planning effort; and 3.
identification of an onsite individual with responsibility for emergency planning.
C-23
CONCLUSIONS In accordance with Section 50.47(a)(l) of 10 CFR Part 50, no operating license for a nuclear power reactor wi 11 be issued unless a finding is made by NRC that the state of onsite and offsite emergency preparedness provides reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency.
Since the staff is required by Section 50.47(a)(2) of 10 CFR Part 50 to base its finding on a review of. FENA's findings and deter-minations as to whether State and local emergency plans are adequate and capable of being implemented, the following items are required in order to make a favor-able finding.
(1)
Correction of the deficiencies identified in the foregoing evaluation to an extent sufficient that the 16 standards in Section 50.47(b) of 10 CFR Part 50 are met.
(2) 'ubmission of the State and local emergency response plans in accordance with Section 50.33 of 10 CFR Part 50.
The submission should include a
definitive description of the Emergency Planning Zones (EPZs) including the factors identified in the aforementioned section of the regulations which were used in determining the"exact size and configuration of the EPZs.
(3)
Submission and satisfactory evaluation of the emergency plan implementing procedures in accordance with the requirements of Section V of Appendix E
to 10 CFR Part 50.
(4)
Acceptable review of the FEMA findings and determinations as to whether the State and local emergency response plans are adequate.
(5)
Acceptable findings from an onsite appraisal to establish that the applicant's plan is capable of being implemented.
(6)
Acceptable 'completion of the joint exercise held in accordance with the requirements of Section IV.F.1.b of Appendix E to 10 CFR Part 50 to demonstrate the dynamic capability to implement the onsite and offsite emergency response plans.
Upon completion of the above items, the staff's evaluation of the overall state of emergency preparedness for the Palo Verde site will be presented in a supplement to this report.
C-24
APPENDIX 0 PRINCIPAL CONTRIBUTORS R. Kirkwood D. Terao J.
Chen M.
Hum E.
Rossi J.
Rosenthal S. Kirslis J.
Mermiel V. DeLi so A. Ramey-Smith C. Gaskin T.
McKenna L. Phillips G. Zwetzig Mechanical engineering Mechanical engineering Geotechnical engineering Materials engineering Instrumentation and control Instrumentation and control Chemical engineering Auxiliary systems Operator licensing Human factors issues Security Emergency planning Core performance Licensee qualifications Palo Verde SSERl D-1
APPENDIX E
ERRATA TO SAFETY EVALUATION REPORT Section 1.10 Pa e 1-9 Item (ll) - Change 5.3.3 to 5.4.3 within parentheses.
Item (13) - Change 5.3.3 to 5.4.3 within parentheses.
Item (14) - Delete III.D.3. 3 within parentheses.
Section
- 6. 2. 4 Pa e 6-12 Last Para ra h
Delete item (5) in its entirety and renumber item (6) as item (5).
Section 9.5.3 Pa e 9-43 Item c)
The time period shown in the second paragraph should read 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> instead of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> (two changes required).
Section 9.5.8 Pa es 9-53 and 9-54 Insert last paragraph on page 9-53 at the end of last paragraph on page 9-52.
Palo Verde SSERl E-1
NRC FORM 335 17.771 U.S. NUCLEAR REGULATORY COMMISSION BIBLIOGRAPHICDATA SHEET
- 1. REPORT NUMBER (Asstuned btr DDCJ NUREG-0857 Su lement No. 1 4, TITLE AND SUBTITLE (Add Volume No.,ifapprop'riateJ Safety Evaluation Report Related to the Operation of Palo Verde Nuclear Generating Station, Units 1, 2 and 3
- 7. AUTHOR(s)
- 2. (Leave blank J
- 3. RECIPIENT'S ACCESSION No.
- 5. DATE REPORT COMPLETED
- 9. PERFORMING ORGANIZATION NAME AND MAILINGADDRESS (Include Zip Code/
U.S ~ Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Washington, D.CD 20555 MONTH Februa DATE REPORT ISSUED MONTH Feb
- 6. (Leave blankJ B. (Leave blankl YEAR 1982 YEAR
- 12. SPONSORING ORGANIZATION NAME AND MAILINGADDRESS (Include Zip CodeJ Same as 9. above
- 10. PROJECT/TASK/WORK UNITNo.
- 11. CONTRACT No.
13, TYPE OF REPORT PE RIOO COVE RE 0 (Inclusive datesJ 15 SUPPLEMENTARY NOTES
- 14. (Leave blankJ Pertains to Docket Nos.
STN 0- 28 STN 0- 2 8c STN 0-0
- 16. ABSTRACT (200 words or lessl Supplement No. 1 to the Safety Evaluation Report for the application filed by Arizona Public Service Company, et al, for licenses to operate the Palo Verde Nuclear Generating Station, Units 1, 2 and 3 (Docket Nos.
STN 50-528/529/530)7 located in Maricopa County, Arizona has been prepared by the Office of Nuclear Reactor Regulation of the Nuclear Regulatory Commission.
The purpose of this supplement is to update the Safety Evaluation by providing (1) the evaluation of additional information submitted by the applicants since the Safety Evaluation Report was issued, (2) the evaluation of the matters the staff had under review when the Safety Evaluation Report was issued, and (3) the response to comments made by the Advisory Committee on Reactor Safeguards.
17, KEY WORDS AND DOCUMENT ANALYSIS 17a. OESCRIPTORS 17b. IOENTIFIERSIOPEN ENDED TERMS IB. AVAILABILITYSTATEMENT Unlimited N4C FORM 335 17 77)
- 19. SECURITY CLASS (This reportl Unclassified
- 20. SECURITY CLASS LThis pageJ
'ncfass3.Pea
- 21. No. OF PAGES 22, PRICE S
1: