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SAFEIT EVALUATION'REPOR BY'THE OFFICE'0F'NUCLENt REACTOR REGULATION i i
0;-S: NUCLEAR REGUIA'ICFY COMISSION IN THE MATI'EP'OF PACIFIC' GAS AND ELECTRIC COMPANY q
. l DIABID CANYON ' NUCLEAR' POWER STATION ; ' UNITS '1
- AND- 2 DOCKEI N05;-50-275*AND*50-323 0-l f
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9707290312 070721 PDR FOIA CO,NNOR 87-214 PDR
s TABLE OF CCNTENTS PAGE
.l.0 IN1PODUCTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1- 1 3.0 DESIGN CRITERIA - STRUCIURES, CCtiFONERfS, EQUIPMENT, AND SYSTEMS........................................................ 3-1 3.6 Protection Against the Dynamic Effects Associated With the Postulated Rupture of Piping. . . . . . . . . . . . . . . . . . . . . . . . . . 3-1 4.0 RFACICR . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 . ........... 4-1 4.4 Thermal and Hydraulic Design. . . . . . . . . . . . . . . . . . . . . . . . . ... . . . 4-1 5.0 REACIOR C00IAND SYSTEM. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-1 5.2 Integrity of Reactor Coolant Pressure Bounda y Components. 5-1 5.2.2 Overpressurization Protection....................... 5-1 5.2.8 Inservice Inspection Program....................... 5-6 l
- 6. 0 ENGINEERED SAFE 1"I FEATURES . . . . . . . . . . . '. . . . . . . . . . . . . . . . . . . . . . . . . . 6-1 6.2 Containment Systems . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-1 6.2.1 Containment Functional Design...................... 6-1 6.3 Emergency Core Cooling System. . . . . . . . . . . . . . . . . . . . . . . .. . . . . . 6-1 l 7.0 INSTRUMENTATION AND CONTR0LS................................... 7-1 1
l l 7.2 Reactor Trip Systs....................................... 7-1 l
7.2.5 Anticipated Transients Without Scram. . . . . . . . . . . . . . . 7-1 l
8.0 ELECTRIC POWER................................................. 8-1 8.1 Gene r al . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 -1 i
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TABLE OF COtEDES l PAGE J
I 9.0 AUXILIARY SYSTEMS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ~ . 9-1 !
1 9.5 Air-Conditioning, Heating, Cooling, and Ventilation Systems. . . 9-1 )
l 9.5.5 Diesel Generator Compartments.......................... 9-1 I i
1 11.0 RADIOACTIVE WASTE MANAGEMEtE......................'................. 11-1 ;
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11.1 Sumary Descr iption . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11-1 i l
13.0 COtOUCT OF OPERATIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13-1 13.6 IM ustr ial Secur ity . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13-1 15.0 ACCIDENT ANALYSIS.................................................. 15-1 15.1 Genera 3...................................................... 15-1 l 15.2 Design Basis Accident Assumptions. . . . . . . . . . . . . . . . . . . . . . . . . . . . 15-1 15.2.2 Fuel Handl ing Accident. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15-1 18.0 REVIEW BY THE ADVISORY COMMITTEE ON REACIOR SAFEGUARDS ( ACRS) . . . . . . 18-1 21.0 FINANCIAL PROTECTION AND INDEMNITY REQUIREMENTS. . . . . . . . . . . . . . . . . . . . 21-1 ,
l 21.1 Preoper ational Storage of Nuclear Fuel . . .. . . . . . . . . . . . . . . . . . . . 21-1 l 21.2 Operating License............................................ 21-1 21.3 Concl us ion . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21-2
22.0 CONCLUSION
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i APPENDICES ;
PAGE I APPENDIX A - CONTINUATION OF THE CHRONOIDGY OF THE RADIOIDGICAL SAFE 7fY REVIEW........................................ A-1 APPENDIX B - ADVISORY COf91I'ITEE ON REACTOR SAFEGUARDS GENERIC ITEMS................................................ B-1 T._A.BLES
,1 TABLE 11.1 - CAIfUIATED RELEASES OF RADIOACTIVE MATERIALS IN l LIQCID EFFIDENTS FROM DIABID CANYON, UNITS 1 AND 2. . 11-7 l l
TABLE 11.2 - CALCULATED RELEASES OF RADIOACTIVE MATERIALS IN GASEOUS EFFLUENTS FROM DIABID CANYCN, UNITS 1 AND 2. 11-8 TABLE 11.3 - COMPAPlSON OF DIABIO CANYON, UNITS 1 AND 2, WITH APPENDIX I TO 10 CFR PART 50. . . . . . . . . . . . . . . . . . . . . . . . 11-9 TABLE 15,1 - POTENTIAL OFFSITE DOSES DUE '10 DESIGN BASIS l
ACCIDENTS........................................... 15-4 i l
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1.0 INTRODUCTION
The Camission's Safety Evaluation Report in the matter of Pacific Gas and Electric Company's application for operating licenses for the Diablo ' Canyon Nuclear Power Station, Units 1 and 2, was issued on October 16, 1974. In the Safety Evaluation Report it was stated that supplemental reports muld be issued to update the Safety Evaluation i
Report in those areas where the staff's evaluation had not been cce-pleted. Supplement Nos. 1, 2, 3, 4, and 5 to the Safety Evaluation Report, issued on January 31, 1975, May 9, 1975, September 18, 1975, May 11, 1976, ard September 10, 1976, respectively, documented the resolution of certain outstanding items, and stam.arized the status of i
the remaining outstanding items.
The purpose of this supplement is to further update the Safety Evaluation Report by providing our evaluation of certain matters that were not resolved when Supplement No. 5 was issued and our evaluation of new issues that have arisen since Supplement No. 5 was issued. Each of the following sections of this supplement is ntnbered the same as the corresponding section of the Safety Evaluation Report that is being updated. A summary of the remaining outstanding issues, which will be addressed in future supplements to the Safety Evaluation Report, is presented in Section 22.0 of this supplement.
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- 1 Appendix A to this supplement is a continuation of the chronology of i the Nuclear Regulatory Comission staff's principal actions with respect to radiological matters related to the processing of the application. Appendix B notes the status of itecs in the Advisory Committee on Reactor Safeguards Report on Generic Items in relation to Diablo Canyon, Units 1 and 2.
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I 3.0 DESIGN CRITERIA - STRUCTURFE , COMPONENTS, EQUIPMEN1' AND SYSTEMS 3.6 Protection Against the Dynamic Effects Associated With the Postulated Rupture of Piping We stated in the Safety Evaluation Report that we would provide our l
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evaluation of postulated pipe rupture outside containment for Diablo Canyon, Units 1 & 2 in a future report.
l In response to our requests, the applicant has provided additional infor-mation in Amendment No. 44 to the Final Safety Analysis Report and in a letter dated July 6,1977. We have determined from our review of this I information that the design of Unit No.1 is acceptable. However, we have not completed our review of Unit 2. Our evaluation is provided below.
The Unit 1 design accommodates the effects of postulated pipe breaks and cracks in high energy fluid piping systems outside containment with respect to pipe whip, jet impingement and resulting reaction forces, and environ-mental effects. The means used to protect safety-related systems and components include physical separation, enclosure in suitably designed structures or compartments, physical pipe enclosures, pipe whip restraints, and equipnent shields. The applicant analyred high energy piping systems for the effects of pipe whip, jet impingement, and environmental effects !
on safety-related systems and structures.
For moderate energy systems, protection from the jet and environmental effects due to critical cracks was incorporated into the design.
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Unit 1 has the ability to sustain a high energy pipe failure coincident with a single active failure and retain the capability of safe cold shutdown. For postulated pipe failures, the resulting environmental effects will not preclude the habitability of the control l room, the accessibility of other areas that have to be manned during _
and following an accident, or the loss of function of electric power supplies, controls and instrumentation needed to complete a safety action.-
Based on our review, we find that the applicant has adequately designed and protected areas and systems required for safe plant shutdown following postulated events, including the combination of pipe failure and single active failure. The design of Unit No.1 meets the criteria set forth' in our letter dated December 12,1972, " General Information Required for Consideration of the Effects of a Piping System Break outside Containment," regarding the protection of safety-related systems and components from a' postulated high energy line break, and the Branch Technical Position APCSB 3-1 regarding the protection of safety-related systems and components from a postulated moderate energy line failure.
Therefore, we have concluded that the Unit No.1 design for the protection of safety-related equipnent from the effects of postulated piping failures outside containment is acceptable.
There will be some design differences between Unit 1 and Unit 2 in the methods provided for protecting against postulated pipe breaks.
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We have requested the applicant to provide us with additional infor- l mation describing where Unit 2 will be different and, where there are differences, describing the protection provided in Unit 2. Our l
evaluation for Unit 2 will be presented in a future supplement to the Safety Evaluation Report.
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- s 4.0 REACIOR I 4.4 Thermal and Hydraulic Desigm Introduction In Supplement No. 4 to the Safety Evaluation Report we indicated that we had reviewed the Diablo Canyon core design calculations with respect to margins from departure from nucleate boiling (DNB). Two areas of our evaluation were not completed. However, we had examined the potential penalties associa-ted with the two open areas and compared the potential penalties with the i margins available. On this basis we had established, as an interim position, acceptable technical specification limits to be ecoloyed pending completion of our evaluations.
Since supplement No. 4 was issued, we have completed our evaluation in one of the two areas that was open. However, in another area, that was closed in Supplement No. 4, we have received additional information that may change our conclusion. Accordingly, we have considered the potential penalties associated with the areas that are now open and cacpared them with the margins that are available. We have determined, as an interim position, acceptable technical specification limits to be employed pending further experimental work.
Our evaluation is provided below.
Effect of Nonuniform Heating on DtB In Section 4.4 of Supplements 2, 3, and 4 we stated that we were reviewing the results of DNB tests involving nonuniform heating. These results were reported in WCAP-8536 (Proprietary) and WCAP-8537 (Nonproprietary),
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" Critical Heat Flux Testing of 17 x 17 Fuel Assembly Geometries with 22 Inch Grid Spacing." We also indicated that, unless our evaluation of this matter was completed before the final technical specifications for Diablo Canyon were ready, we would require that the mininzn allowable departure i
from nucleate boiling ratio (DNBR) be increased by 5 percent above that re- I quired to satisfy the 95/95 criterion to account for the incomplete data base.
We have now completed our evaluation of these topical reports. Our approval was documerited in a letter to Westinghouse dated February 11, 1977. Accord-ingly, the 5 percent penalty that was associated with this open item is no longer appropriate. l Effect of Bowed Rod on DNB In Section 4.4 of Supplement No. 4 we stated that we had completed our review of WCAP-8176 (Proprietary) and WCAP-8323 (Nonproprietary), "Effect of Bowed Rod on ENB." We had found these reports to provide an acceptable data base for predicting the effects of rod bowing on DNB heat flux.
However, in a letter dated August 13, 1976, Westinghouse reported data that indicated these methods for predicting the effect of fuel rod bowing on DNB may not contain adequate margins when unheated rods, such as instrument tubes, l
are present. Further experimental verification of these data is still in progress. As an interim position, we will require appropriate limitations in the technical specifications pending completion of the experimental work and our evaluation of it.
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i Ancunt of Rod Bowing We stated in ection 4.4 of Supplement No. 4 that we had not completed our evaluation of the rod bowing model presented in Westinghouse Topical Reports WCAP-8691 (Proprietary) ard WCAP-8692 (Nonproprietary), " Fuel Rod Bowing," ,
l dated December 1975. However, based on our evaluation to date, usiry conserv-ative calculations, we had estimated the penalties on [NBR relative to the 95/95 criterion that would be necessary to account for the data, including l
an allowance for uncertainty in extrapolation of data from 15 x 15 fuel assemblies to predict the amount of bowing in 17 x 17 fuel assemblies.
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Since we have not yet completed our evaluation of these topical reports, our position remains uncharged. l Evaluation As an interim measure, to account for the items discussed above, we will include a burnup dependent penalty factor to be applied to the reactor opera-ting limits in Section 3.2.3 of the Technical Specifications to reflect the I reduced DNB corditions caused by increasing fuel rod bowirg. The enthalpy hot channel factor, a parameter which varies inversely with EIG, is used to account for this penalty. This penalty is derived by extrapolating available test data and provides a conservative safety margin which we con-sider acceptable for the Diablo Canyon plant.
Conclusion We have concluded that, as an interim position, the Diablo Canyon cores can be operated safety utilizing appropriate technical specification limits dis-cussed above to account for uncertainties with regard to rod bowing and ,DNB.
We consider this matter resolved.
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5.0 REACIOR COOLANT SYSTEM I
5.2 Integrity of Reactor Coolant Pressure Boundary Components $
1 5.2.2 Overpressurization Protection i
Introduction !
In the Safety Evaluation Report we concitx3ed that the overpressure protection for the reactor coolant system, as provided by the three spring loaded pressurizer safety valves, was acceptable. (In addition, Diablo Canyon employs three power operated pressurizer relief valves.
However, they were not relied upon and were not discussed in the Safety l Evaluation Report.)
Since then, we have reviewed the possibility of an overpressure transient at temperatures below operating temperatures during startup or shutdown when the allowable pressure is lower than the pressurizer relief valve setpoints. The applicant has proposed interim measures to minimize the possibility of such overpressure transients below normal operating temperatures. We have reviewed the interim measures and found them acceptable for use on Unit 1 during the first fuel cycle. Our evaluation is provided below.
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l Background l Incidents known as pressure transients (events that have exceeded the technical specification temperature-pressure limits of the reactor vessel) were discussed in a technical report issued in November 1976, "NURDG-0138, " staff Discussion of Fifteen Technical Issues listed in 5-1 L_-_________-______--___________-________-_-_______-_
Attachment to November 3,1976 Memorandum from Director, NRR to NRR Staff." The report concluded in part that pressure transients are of concern during startup and shutdown because, at these relatively low temperatures, the vessel has less toughness than at operating temperatures and irradiation increases the temperature at which steel attains maximum toughness. The majority of these incidents occurred during cold shutdown while the system was in a solid-water condition (no bubble in the pressurizer).
The allowable pressure limits at lower temperatures are determined in accordance with Appendix G to 10 CFR Part 50 and are included in the technical specifications for operating licenses. The limits change 1 during the life of the plant as the pressure vessel becomes irradiated.
Because it would be impractical to change these ' limits continuously, they are calculated for an extended period of time. Thus, the limits in effect at a given time may be based on properties expected in the vessel five or more years in the future, makirg them conservative during the early portion of this period. The report concluded that large safety margins exist for unirradiated reactor vessels and new plants can be permitted to be licensed un6er existing safety criteria. Nevertheless, we have concluded that administrative procedures and overpressure protection devices should be upgraded to reduce the likelihood of future overpressure transients. l 5-2 i
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Evaluation The applicant has described several modifications to his administrative i procedures, design, and operator trainirs to minimize the likeli. hood of I
overpressure transients. These measures were documented in Amendment 48 to the Final Safety Analysis Report and in a letter dated July 5,1977. )
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(1) Operator Training - Operators have been briefed on the types of l events that could cause overpressurization and changes have been i
made in the operatin, procedures to minimize the probability of such events.
I (2) Besidual Heat Removal Relief Valves - The residual heat removal system will not be isolated from the reactor coolant system when the plant is in a solid-water condition, maxing the residual heat renoval system relief valves available for pressure relief. These valves are ser to relieve at pressures lower than the Appendix G pressure limits for the Unit 1,oressure vessel.
(3) Steam Bubble - A steam bubble will be for aed in the pressurizer l
when the reactor coolant temperature reaches 160 degrees Fahrenheit during plant heatup. During the cooldown the bubble will be collapsed when the reactor coolant temperature reaches 160 degrees Fahrenheit.
This will minimize the amount of time in a water-solid condition.
(4) Operating Procedures - As many operations as possible will be performed while the plant is not in a solid-water condition.
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In addition, instructions have been added to the operating proce-dures which only permit slow changes in the residual heat removal system flow rate.
(5) Letdown Line - The letdown- flow control valve will be placed in manual control in a fully open position when the reactor coolant system is in a water-solid condition. Additionally, an orifice relief valve on this line is available as another pressure relieving mechanism.
(6) Reactor Coolant Pumps - At least one pump will be operating when-the reactor coolant system temperature is above 160 degrees Fahrenheit. If all the pumps should be shut down for any reason, appropriate restrictions will be employed on restarting to avoid possible pressure transients from injecting any accumulation of cold pump seal injection water into a warm reactor coolant system.
l (7) Accumulators - Che accmulator isolation valves will be closed 1 and power will be removed and locked out whenever the primary system is in a solid-water condition.
(8) Alarm - An interim alarm will be installed to annunciate on the ,
main control board whenever the reactor coolant system pressure approaches within 50 pounds per square inch of the allowable pressure as calcuated by a function generator. The pressure cal- l culated by the function generator conservatively bounds the ;
Appendix G Ifmits.
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i (9) Safety Injection Pumps - Safety Injection P.mnps are kept inactive whenever the primary system'is solid. Power is removed by physic-ally racking out the pump motor circuit breakers. When the ptznp
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motors are de-energized, an alarm is actuated in the control room )!
indicating the status.
i We have reviewed these proposed administrative and' design changes and I a find them acceptable as an interim measure to minimize the likelihood I
of an overpressure transient at low temperature.
In addition, we have reviewed the material properties of the Unit I reactor vessel. In light of the margins availaole during the first fuel cycle we have determined that credible overpressure transients would not cause reactor vessel failure. ,
Conclusions Based on our evaluation as described above, we have concluded that the interim measures proposed by the applicant to minimize the possibility of an overpressure transient are acceptable for use on Unit I during the first fuel cycle.
The applicant is a member of a group of utilities that is developing a long-term solution for this issue. The design modifications under consideration would employ the power operated pressurizer relief valves, with lifting setpoints programmed as a function of reactor I coolant temperature, to preclude violating the pressure limits of 5-5 L
Appendix G to 10 CFR Part 50. We will review the proposed long-term solution when the supporting analytical information is availabla.
We will include a condition in any operating license for Unit I requiring that an acceptable long-term overpressure protection system be installed prior to the initiation of the second fuel cycle. f Since the Unit 2 reactor vessel has somewhat less resistance to brittle fracture at low temperature than the Unit l vessel we have not yet evaluated this matter for Unit 2. We will provide our evaluation in a future supplement to the Safety Evaluation Report.
5.2.8 Inservice Insoection Procram In the Safety Evaluation Report we stated that the inservice inspection program would ecmply with: (1) the ASME Boiler and Pressure Vessel Code,Section XI, including Addenda through Winter 1972, for Class I com-ponents and (2) Regulatory Guide 1.51 for inspection of Class 2 systems.
Those provisions were acceptable at that time.
l Since then, the Commission has revised its requirements for inservice inspection. The new requirements for ASME Class 1, 2, and 3 canpon-ents are given in Section 50.55a(g) of 10 CFR Part 50, initially pub-lished February 12, 1976. We will require, in the technical specifica-tions, that the applicant ca: ply with this regulation. I In very general terms, the regulation requires that the inservice inspections conform, to the extent practical within the limitations of design, geometry and materials, to the requirements of a periodically updated edition of ASME Section XI.
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In a letter dated May 23,1977, we provided detailed guidance to the -
applicant on complying with the regulation. We also requested sub-mittaloftheapplicant'sproposedinserviceinspectionplan,a5ong with justifications for any items where the applicant may determine )
l that meeting ASME Section XI requirements would be impractical. '
We requested the submittal by September 30, 1977, which should provide adequate time for oyr evaluation of any requests for relief from ASME Section XI requirements prior to the need to implement the inservice j inspection program at the Diablo Canyon plant. ]
The technical specifications will require conformance to 10 CFR 50.55a(g) and, accordingly, we have concluded that this provides an acceptable basis I
for the inservice inspection and . testing program to comply with' Criterion
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32 of the General Design Criteria.
We consider this matter resolved. t i-1 1
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6.0 ENGINEERED SAFCI"i FF15URE_S 6.2 Containment Systems 6.2.1 Containment Functional Deggn
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In :Netion 6.2.1 of the Safety Evaluation Report we presented our i evaluation of the applicet's calculations of the containment presssure i response to a postulated loss-of-coolant accident and a postulated main steam line break. (6ubcompartment pressure responses were open items in the Safety Evaluation Report. Later, they were resolved in Section 6.2.1 of Supplements 2 and 3.)
1 Our recent evaluations of a postulated main stean line creak inside containment have indicatud a potential concern in two areas: (1) The peak calculated containment pressure and temperature, and (2) The enviremental qualification of safety-related equipent located inside ,
I containment that must function. 1 In a Ic*.ter to the applicant dated June 1,1977 ve requested additional information on these subjects. The applicant currently expects to respond to our request on about September 15, 1977. We will review the applicant's subnittal and provide our evalugicn in a future supplement
'to the Safety Evaluation Report.
i Emergency Core Cooling System (ECCS) !
6.3 We stated in Suppleinent No. 5 to the Safety Evaluation Report that the ECCS performance evaluation was acceptable, subject to satisfactory resolution ll 6-1 i
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of two. items concerning fuel rod bowing and the temperature of the reactor coolant in the upper reactor head. These matters are now resolved.
Our evaluation of rod bowing and departure from nucleate boiling is presented in Section 4.4 of this supolement. Our evaluation of ECCS performance is precented below.
Emergency Core Cooling System Performance In the analyses for Diablo Canyon, Units 1 and 2, an initial upper head temperature corresponding to the core inlet temperature had been used.
These analyses had been accepted as described in Supplement No.
4 and Supplement No. 5.
In August 1976, Westinghouse reported that the fluid temperature in the upper head region may be higher. Recent data from operating facilities have indicated that t'he effective upper head temperature is between the cold leg and hot leg temperature. kkstinghouse has performed sensitivity studies that show the calculated peak clad temperature increases for higher I
upper head temperatures. These were reported in a letter to the staff dated August 13, 1976. As a result, we have requested that the large breaks (which are most limiting) be conservatively reanalyzed with an upper head temperature corresponding to the bot leg temperature.
In Amendment '47 to the Final Safety Analysis Report the applicant submitted loss-of-coolant analyses for a large pipe break assuming the hot leg temperature would exist in the upper reactor hea$. The higher upper head temperature was used in a generic study provided in the Westinghouse Topical 6-2 k
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Report, WCAP-8865, " Westinghouse ECCS-Four Icop Plant (17 x 17) Sensitivity Studies with Upper Head Fluid Temperature at T(bot)," dated October,1976.
This generic study indicated that the double-ended cold leg guillotine break was still the most limiting break for Westinghouse four-loop plants.
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l The applicant has referenced the, loss-of-coolant analyses performed for I Salem Nuclear Generating Station, Unit No.1 (Docket No. 50-272), to establish that the worst break size for Diablo Canyon, Units 1 and 2, remains the double-ended guillotine break "ith a discharge coefficient of 0.6. The Salem Unit 1 analyses are for a design similar to Diablo Canyon, Unit 1. We have reviewed them and found them to be acceptable and, therefore, they are acceptable for our review of Diablo Canyon, Unit t
. I and 2. These calculations satisfy the requirements of 10 CFR 50.46 to analyze a spectrum of different breaks to determine the worst case.
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The applicant has established that the double-ended cold leg guillotine break loss-of-coolant accident analysis using a discharge coefficient of 0.6 presented for Unit No. 2 provides boundirg results for both Unit I and Unit 2. The analyses were perforrbed wi2 the ~0ctober,1975 version of the Westinghouse Evaluation Model. This was docunented in Westinghouse !
1 Topical Reports WCAP 8522 (Proprietary) and WCAP 8523 (Nonproprietary),
" Westinghouse ECCS Evaluation Model, October 1975 Version," November 1975.
We have reviewed and approved this model. Our approval was docunented in a letter to Westinghouse dated May 13, 1976.
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The analyses, performed with a total peaking factor of 2.32 at a power level of 102 percent of 3411 megawatts. thermal, identified the worst break as a double-ended cold leg guillotine with a discharge coefficient i
of 0.6. The calculated peak clad temperature is 2130 degrees Farenheit, which is below the acceptable limit (2200 degrees Farenheit) specified ]
1 in 10 CFR Part 50.46. In addition, the calculated maximum local- I metal / water reaction of 6,76 parcent and a total core-wide metal / water reaction of less than 0.3 percent are well below the allowable limits of 17 percent and 1 percent, respectively, as delineated in 10 CFR 50.46.
The applicant's calculation of the containment pressure transient was based on the Westinghouse Evaluation Model and revised mass ard energy 1
release data. All plant dependent input parameters remained identical to those assumed for the emergency core cooling system evaluation previously performed and reported in Ameninent 35 to the Final Safety Analysis Report.
i Since the reanalysis of the containc.ent pressure transient was done using the Westinghouse Evaluation Model and the same plant dependent input parameters, we reaffirm our conclusions as stated in Supplement No. 4 i
stated that the calculated containment pressures are in accor, dance with j i
Appendix K.to 10 CFR Part 50 and, therefore, are acceptable.
The single failure evaluation of the emergency core cooling system has been previously reviewed and found acceptable, as stated in Supplement No. 5.
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The evaluations of long term boric acid buildup anl subnerged valves have been previously reviewed and found acceptable as stated in Supplement No.
- 4. A 5 percent spike penalty was discussed in Section 6.3 of Supplement No. 4 in connection with rod bow. This is no lorger appropriate for the ECCS* performance analysis in light of our current evaluation of rod bow,
' which is pre'se,nted in Section 4.4. of this Supplenant.
The applicant has stated that the reactor core ard internals have been designed so that the reactor can be safely shut down and the essential heat transfer geometry of the core preserved following a postulated loss-of-coolarit accident. The clad temperature transient is terminated at a time when the core geometry is still amenable to cooling and the 1
cladding oxidation limits of 17 percent are not exceeded during or after !
quenching. ,
Based on our review as described in the safety traluation Report and
- i Supplements Nos. 4 ard 5 and in this supplement, we have concluded that (1) the IDCA analyses that were performed are whally in accordance with the requirements of Appendix K to 10 CFR Part 50, (2) the ECCS coolirg performance conforms to the peak clad temperature and maximum ozidation !
and hydrogen generation criteria of 10 CFR 50.46, (3) ECCS cooling performance will be adequate despite any postulated failure of a single active emponent, (4) adequate systems are available to provide long-term core cooling to the reactor vessel and, therefore, (5) the emergency core cooling system design is acceptable. i 6-5
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7.0 INSTRUMENTATION AND, CorfrROLS 7.2 Reactor Trip Systm 7.2.5 Anticipated Transients Without Scram In the Scfety Evaluation Report and in Supplement No. 4 we stated that we had not completed our evaluation of the information which had been subnittw3 by the applicant concerning anticipated transients without scram. The current status of this matter is described below.
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Our position with respect to anticipated transients without scram is provided in the technical report, " Anticipated Transients Without l Scram for Water-Cooled Power Reactors," WASH-1270, dated September 1973.
Unit I was classified by the staff as a "I.C." facility as defined in WASH-1270; for this Unit, WASH-1270 indicat.es an analysis describing and evaluating the consequences of a postulated anticipated transient without scram would be acceptable. Unit 2 was classified as a 'I.B."
l l facility; for this Unit, WASH-1270 indicates a program to incorporate any design changes necessary to assure that the consequences of anticipated transients without scram would be accepable. The applicant submitted the infonution described by WASH-1270 in a letter dated October 1, 197'4. In that letter the applicant referred to two Westinghcase topical reports, WCAP-8330, " Westinghouse ATWS Analysis,"
dated August 1974 and WCAP-7706, "An Evaluatica of Solid State Logic Reactor Ptotection in Anticipated Transients," dated July 1971. The applicant stated that these reports contain the necessary analyses 7-1
.. 1 l
and that they are applicable to the Diablo Canyon Nuclear Power Plant.
In eidition, the applicant stated that Unit 1 as well as Unit 2 would 1 be considered as a category "I.B." facility.
We evaluated these Westinghouse topical reports and presented our conclusions in a report, " Status Report on Anticipated Transients Without Scram for Westinghouse Reactors," dated December 9,1975. ..
This report described a ntrnber of outstanding issues. We have discussed the status report with the Advisory Conmittee on Reactor Safeguards.
Since then Westinghouse has submitted additional information. We are i
evaluating this information and we plan to issue staff positions in )
1 l
September 1977. l If our review shows that design modifications are necessary to mitigate the consequences of anticipated transients without scram we will require that they be implemented at Diablo Canyon, Units 1 and 2.
The effect of this matter on the review and licensing process at this time does not change the conclusion as stated in WASH-1270 that
! lunitations on operation on this account are not necessary or appropriate. This conclusion is based on our determination that the likelihood of an anticipated transient without scram event is very low' considering the number of plants now in operational status, 'or expected to come 6to operation before our requirements can be fully implemented.
7-2
8.0 ELECTRIC KMER 8.1 General Effects of Degradation of Offsite Power Voltaae We were informed on July 20, 1976 by the Northeast Nuclear Energy '
Company that following a trip on July 5,1976 at the Millstone Nuclear i Station, Unit 2 (Docket No. 50-336), equipent failures occurred during 1
a degraded grid voltage corviition.
1 On July 27, 1976, we notified all licensees with operating reactor j facilities of the events which had occurred at the Millstone, Unit No. 2 facility, and requested the licensees to investigate the potential for equipent failures for degraded voltage conditions. ,
As a result of our initial investigation and evaluation of the events j occurring at Millstone, Unit 2, we considered it necessary to require all plants presently in review for an operating license to conduct a thorough evaluation of the problem and to sutnit formal reports.
In a letter dated June 6,1977, we requested that the applicant l evaluate the Diablo Canyon, Units 1 and 2 design for the Class IE electrical distribution system to determine whether the operability of safety related equipent, including associated control circuitry and instrumentation, can be adversely affected by short term or long term degradation in the offsite power system.
l 8-1
The applicant currently expects to sutnit a response by about October 15, 1977. We will evaluate the applicant's response and will report the resolution of this matter in a future supplement to the Safety Evaluation Report.
Subject to resolution of this matter, we reaffirm our conclusions as stated in the Safety Evaluation Report that the electrical power system )
for Diablo-Canyon, Units 1 and 2 is acceptable..
l i
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8-2
S 9.0 AUXILIARY SYST'z:MS 9.5 Air Conditioning; Heating and Ventilation Systems 9.5.5 Diesel Generator Compartments In" the Safety Evaluation Report we presented our evaluation of the diesel generator compartment ventilation systsc.
In addition to the ventil' ti6n a features described in the Safety Evalu-ation Report, the diesel generators draw some cooling air from the main turbine bay section of the turbine building. This air is essential for cooling the gener9 tors.
The applicant has informally ccrmitted to install a normally closed fire door to isolate the diesel generator ccr:cartments from the main section of the turbine building. An alternate source of cooling air will be provided, via ductwrk, from the area above the diesel gene-l rator empartments. Since this modification will eliminate the possi-j bility of a fire in the main section of the turbine building causing overheating of the diesel generators, we find the applicant's cormitment acceptable.
i We will review the details of this modification when they are sutnitted 1
and provide our evaluation in a future supplement to the Safety Eval-uation Report.
9-1 1
11.0 RADIOACTIVE WASTE ENAGEMNT 11.1 Sumary Description Background - Appendix I to 10 CFR Part 50 In the Safety Evaluation Report we presented our evaluation of the radioactive waste management systems. We found them to be acceptable j and capable of maintaining the amounts of radioactivity in effluent streams as low as practicable.
In Supplement No. 4 to the Safety Evaluation Report we indicated that the Comission had adopted a new regulation, Appendix I to 10 CFR Part 50. Appendix I provided numerical guidelines as to what l 1
constituted maintaining the amount of radioactivity in effluent streams as low as reasonably achievable and required a reevaluation of the 1
liquid and gaseous radioactive waste management systems. l 1
l The applicant subnitted the information necessary for our evaluation )
i in Amendment 44 to the Final Safety Analysis Report, dated July 29, j 1976 and in a letter dated July 30, 1976. The applicant elected to l detionstrate compliance with an option allowed by a September 4,1975 l amendment to AppeMix I. This option allows certain applicants to dettonstrata compliance with the Annex to Appendix I, " Concluding Statement of Position of the Regulatory Staff (Docket-RM-50-2)," in lieu of performing the detailed cost-benefit sttx3y that would l otherwise be required by Section II.D of AppeMix I.
We have ccrt:pleted our evaluation which is 6escribed below.
11-1
1 l
i 4
Evaluation - Appendix I to 10 CFR Part 50 We have evaluated the radioactive waste manage:nnt systems proposed for Diablo Canyon, Units 1 and 2, to reduce the quantities of radio-active materials released to the envirornent in liquid and gaseous effluents. The radioactive waste management systems were previously described in Section 3.4 of the Final Envirorr. ental Statement (FES) i i
dated May 1973, and in Section 11 of our Safety Evaluation Report l dated October 16, 1974. Based on more recent operating data applicable to the Diablo Canyon Station, and on ch&nges in our calculational model, we have generated new liquid and gaseous source terms to determine conformance with Appendix I. These values are different from those ;
given in Tables 3.6 end 3.7 of the rinal Envirornental Statement.
The new source terms, shown in Tables 11 1 and 11.2, were calculated using the models and methodology described in NCREGa0017, " Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents i from Pressurized Water Reactors (IMR-GALE Code)," April 1976. These ll source terms were used to calculate the doses described below. Our evaluction of the dispersion of radionuclides in and the disposition of radionuclides fecm the atmosphere were bawd on analyses using the methodology provided in Regulatory Guide 1.111, " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," March 1976. The mathm atical models used to perform the dose calculations are contained in Regulatory 11-2 1
l l
i Guide 1.109, " Calculation of Annual Average Ibses to Man from Routine Releases of Reactor Effluents for the Purpose of Implementim Apperdix I,"
March 1976.
In'cluded in our analysis are dose evaluations of three effluent )
categories: (1) pathways associated with liquid effluent releases to the Pacific Ocean, (2) noble gases released to the atmosphere, and (3) pathways associated with radioiodines, particulate, caroon-14 and tritium released to the atmosphere.
The dese evaluation of pathways associated with liquid effluents was )
i based on the maximum exposed individual. The dietary and living (
1 habits for an adult irdividual included the consumption of 21 kilograms !
l per year of fish harvested in the immediate vicinity of the cooling l water discharge into Diablo Cove, and recreational use of its shore-line for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> per year. There are no drinking water sources j l
receinng Diablo Canyon Station liquid effluents. The maxinran dose l comitment resulting from extr>sure to water fran Diablo Cove was estimated to ba 0.024 mrem per year (total body) aM 0.077 mrem per l year (thyroid) for an aoult.
The dose evaluation of noble gases released to the atmosphere included a calculation of beta and gamma air doses at the site boundary and total body and skin doses at the residence having the highest dcso.
The maximum air doses at the site bouMary were found at 0.5 mile north-northwest relative to Diablo Canyon, Unit No.s 1 and 2. The 11-3
_ ______________m_ . _ _ _ _ _ _ _ _ _ .
location of maximum total body and skin doses were determined to be at a residence located 1.5 miles north-northwest of the station.
'Ite dose evaluation of pathways associated with radioicdine, particulate, Carbon-14 and tritium released to the atnesphere was also based on the maximum exposed individual. This individual is a child whose diet included the consumption of 41 kilograms per year J
of beef from an animal grazing year-round located 0.5 mile north- l i
northwest of Diablo Canyon, Unit Nos. I and 2.
As shown in Table 11.1, the expected quantity of radioactive materials released in liquid effluents from Diablo Canyon, Units 1 and 2, will be less than 5 Curiec pei year per reactor (0.34 Curies per year per reactor), excluding tritium and dissolved gases, in conformance with the amendment to Section II.D of Appendix I. The liquid effluents ,
i released frca Diablo Canyon, Units 1 and 2, till not result in an l i
annual dose or dose cemitment to the total body or to any organ of an )
i indivirlual, in an unrestricted area from all pathways of exposure, in i excess of 5 mrem as shown in Table 11.3.
Based on our evaluation of the gaseous radwaste management systems, f the total quantity of radioactive materials released in gaseous i
effluents will not result in an annual gamma air dose in excess of i I
10 meads and a beta air dose in excess of 20 mrads at every location which could be occupied by individuals near ground level, and at or beyond the site boundary as shown in Table 11.3. As shown in Table 11.2, 11-4
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i.
the annual total quantity of Iodine-131 released in gaseous effluents will be less than 1 Curie per reactor (0.076 Curie per year per reactor) in conformance with the September 4, 1975 amendment to Appendix I.
The annual total quantity of radiciodine and radioactive particulate released in gaseous effluents from Diablo Canyon, Units 1 and 2, will not result' in an annual dose or dose ecmitment to any organ of an individual in an unrestricted area from all pathways of exposure in excess of 15 mrem as shown in Table 11.3.
Based on our evaluation, the radwaste treatment systems proposed for 1
Diablo Canyon, Units 1 and 2, are capable of maintaining releases cf 1
l radioactive materials in liquid and gaseous effluents during normal l
l operation at doses which will not exceed the design objectives of ;
\
1 Sections II.A, B and C of Appendix I of 10 CFR Part 50.
Our evaluation also shows that the applicant's proposed design of Diablo Canyon, Units 1 and 2, satisfies the design objectives set l forth in the option provided by the Comission's September 4,1975 amendment to Appendix I and, therefore, satisfiesSection II.D of Appendix I of 10 CFR Part 50.
We conclude that the liquid and gaseous radwaste treatment systems will reduce radioactive materials in effluents to "as low as is reasonably achievable levels" in accordance with 10 CFR Part 50.34a and, therefore, are acceptable.
11-5
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D
-s Based on our evalua' tion as set forth in our Safety Evaluation Report 1
and our evaluations as stated above in this Supplement, we find the {
I liquid, gaseous and solid radwaste systems and associated process and effluent radiological monitoring and sampling systems for,Diablo -
Canyon, Units 1 and 2, to be acceptable.
We consider this matte'r resolved.
1 I
l 11-6
~. _ . _ . _ _ . . .
l TABLE 11.1 CAIfUIATED RELEASES OF RADI0 ACTIVE-MATERIALS IN LIQUID' EFFLUENTS FROM DIABID CANYONi UNITS-l' AND 2 CURIES PER' YEAR PER~REACIOR Nuclide Ci/yr/ reactor N_aclide Ci/yr/ reactor Corrosion & Activation Prodacts Fission Products--(Continued)
Chrcnitn-51 3.3(-4) Tellurium-127m 8 (-5)
Manganese-54 1.1(-3) Tellurium-127 8 (-5)
Iron-55 5 (-4) Tellurim-129m 2.7(-5)
Iror.-59 2.2(-4) Telluritm-129 1.7(-4)
Cobalt-58 8 (-3) Iodine-130 4 (-5)
Cobalt-60 9.3(-3) Iodine-131 7.5(-2)
Zirconium-95 1.4(-3) Tellurim-132 2.8(-2)
Niobium-95 2 (-3) Iodine-132 1.7(-3)
Iodine-133 1.1(-2) l Cesium-134 8.4(-2)
Iodine-135 1.9(-3)
Fission Products Cesium-136 9.6(-3) 7.6(-2) l Cesium-137 Bramine-83 1 (-4) Barium-137m 4.9('-2)
Rubiditn-86 9 (-5) Baritn-140 2 (-5)
Strontium-89 8 (-$) Lanthanium-140 2 (-5)
Yttrium-91 2 (-5) Ceritn-141 1 (-5)
Zirconium-95 1 (-5) Cerium-144 5.2(-3)
Niobita-95 2 (-5) All Others 7 (-5)
Molybdenu::.-99 7.5(-4)
Technetitn-99m 8.2(-4) Total 3.4(-1)
Ruthenita-103 1.5(-4) (except Tritium)
Ruthenitm-106 2.4(-3)
Silver-110m 4.4(-4) Tr ititu 710
- Exponential notation: 1(-4) = 1 x 10-4
- Nuclides whose release rates are less than 10 Ci/yr/ reactor are not listed individually, but are incitx3ed in the category "All Others" 11-7
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{
1 TAB 2 11.2 CAIfULATED RELEASES OF RADIOACTIVE MATERIALS IN GASEOUS EFF W ENTS FROM i DIAB W CANYON, UNITS 1 AND 2 {
CURIC.S PER YEAR PER REACTOR Reactor Auxiliary Turbine Air Decay ]
Radionuclides
- Building Building Building Ejector Tanks Total j i
Krypton-83m a- a a a a a l 2 2 a 1 a 5 j Krypton-85m a a a 250 250 l Krypton-85 6 Krypton-87 a 1 a a a 1 l Krypton-88 2 4 a 3 a 9 i a a a a a a 1 Krypton-89 10 a a a 22 32 Xenon-131m Xenon-133m 20 2 a 1 a 23 Xenon-133 1900 110 a 68 170 2300 a a a a a a Xenon-135m 10 6 a 4 a 20 Xenon-135 a a a a a a i Xenon-137 a a a a a a Xenon-138 Iodine-131 2.6(-3) 4.4(-2) 1(-3) 2.8(-2) a 7.6(-2) l Iodine-133 2.8(-3) 6.2(-2) 1.2(-3) 3.9(-2) a 1(-1) l Manganese-54 2.4(-2) 1.8(-4) c. c 4.5(-5) 2.3(-4)
Iron-59 8.3(-7) 6(-5)- c c 1.5(-5) 7.6(-5)
Cobalt-58 8.3(-6) 6(-4) e c 1.5(-4) 7.6(-4) !
Cobalt-60 3.8(-6) 2.7(-4) c c 7(-5) 3.4(-4)
Strontitn-89 1.9(-7) 1.3(-5) c c 3.3(-6) 1.6(-5)
Strontium-90 3.3(-8) 2.4(-6) c c 6(-7) 3(-6)
Cesium-134 2.4(-6) 1.8(-4) c c 4.5(-5) 2.3(-4)
Cesium-137 4.2(-6) 3(-4) e c 7.5(-5) 3.8(-4)
Tritium-3 710 e c c c 710 a a a 7 8 Carbon-14 1 Argon-41 25 e c e c 25
, less than 1.0 curie per year per reactor for noble gases and carbon-14, less than 10 curie per year par reactor for iodine exponential notation: 1.4(-2) = 1.4 x 10-4 less than 1 percent of total for this nuclide radionuclides not listed are released in quantities less than those specified in notes a and c from all sources 11-8
. - + - .
j '..
l TABLE 11.3 COMPARISON-OF'DIABID CANYON; -UNITS 1 AND 2l'WITH APPENDIX I TO 10 CFR PARF 50, SECTIONS II.A, II.B-AND II.C-(MAY 5;-1975(a)-AND 5ECTION II.D, ANNEX-(SEPTEMBER 4,1975)(b) '
Appendix I(a) Annex (b) Design Calculated Doses a Criterion Design Objectives (c)- Objectives Units 1 or ' 2 -
I Liquid Efficents Dose to total 3 mre#yr/ unit 5 mrevyr/ site 0.012 mreW yr/ unit body from all ;
pathways l Dose to any 10 mre#yr/ unit 5 mreW yr/ site 0/038 mreWyr/ unit )
organ from all pathways Noble Gas' Effluents (d)
Gansna dose in air 10 mrad /yr/ unit 10 mrad /yr/ site 0.22 itrad/yr/ unit i
Beta dose in air 20 mrad /yr/ unit 20 mrad /yr/ site 0.51 mrad /yr/ unit Dor.e to total 5 mreWyr/ unit 5 mreW yr/ site 0.016 mreW yr/ unit body of an individual Dose to skin of 15 mreWyr/ unit 15 mreWyr/ site 0.043 mreW yr/ unit an individual Radioiodines'and Other Radionuclides Released to the Atmosphere (e)
Dose to any 15 mreW yr/ unit 15 mreWyr/ site , 1.04 mreW yr/ unit organ from all
' Pathways ,
(a) Federal Register, V. 50, p.19442, May 5,1975.
(b) Fedetal Register, V. 40, p. 40816, September 4,1975.
j (c) Design Objectives given on a site basis. Therefore, these design objectives apply to 2 units at the site.
(d) Limited to noble gases only. .
(e) Carbon-14 and Trititsu have been added to this category.
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.J 13.0 CCFryKT % P g II y 1 13.6 Industrial Security In the Safety Evaluation Report we stated that the industrial security program was acceptable. Since then', the Ccanission has revised the requirements for industrial security programs at commercial nuclear
/
power plants. The new requirements are given in Section 73.55 of 10 CFR Part 73, published in the Federal Register on February 24, 1977.
As required by the regulation, the applicant s:frnitted an amended physical security plan on May 25, 1977 as Revision 2 to the plan.
1 Revisions 3, 4 and 5 were subnitted on June 3,1977, June 15,1977 and l June 29, 1977, respectively.
l l
The regulation requires that a revised security plan, that complies with the requirements of 10 CFR 73.55, except for certain items of construction and installation of equipment that is not already in l place, must be implemented by May 25, 1977 or on the date of receipt of an operating license, whichever is later. ':he applicant has comnitted to implementing these portions of the security plan prior to the date of fuel loading.
A security plan that complies with all the requirements of the new regulation, incl'uding construction and installation items, must be implemented by August 24, 1978 or on the date of receipt of an 1
13-1 I
~~ ' ' - '
O.
operating license, whichever is later. These construction and instal-lation items will be implemented by the applicant prior to August 24, 1978, to upgrade the physical security measures for the plant a site.
This on-going upgrading of physical security is consistent with the graded implementation permitted by the regulation and is, therefore, g acceptable.
We have evaluated the amended security plan and' a security plan review team has visited the plant site as part of this overall evaluation.
As a result of our evaluation, certain areas have been identified where additional information is required before the amended j i
security plan can be found in conformance with the regulation. The 2
applicant has made commitments to modify the snended security plan such that the level of protection will be consistent with the performance requirements of Section (a) of S73.55.
1 i
We will review the details of implementing these commitments and provide our evaluation in a future supplement to the Safety Evaluation Report. 1 I
1 1
13-2 l
l
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15.0 ACCIDDTP ANALYSES l
15.1 General We have presented our evaluation of the accident analyses for Diablo l \
Canyon in the safety Evaluation Report and in Supplements 2 and 3 to the i Safety Evaluation Report. Since then, we have requested detailed eval-uations of fuel handling accidents inside the containment bailding for all plants under review for operating licenses. Our letter of March 11,'1977 forwarded,the request for Diablo Canyon. The applicant provided this evaluation in Amendment 49 to the Final Safety Analysis Report. We have completed our evaluation, which is presented below, i
The applicant has described the plant systems used to mitigate the consequences of a fuel handling accident inside containment. Five 110,000 cubic foot per minute fan coolers are radially arranged around the core cavity, with the inlets located about 30 feet above the surface of the refueling cavity pool and about 20 feet from the edge of the pool.
During refueling, one or more of these fan coolers will be operating.
These five inlets join into a common annular ring. The ring is purged by the containment purge exhaust fan during refueling operations at a rate of 55,000 cubic feet per minute through two 48-inch diameter butterfly valves, in series, and released to the atmosphere through the plant vent. A radiation monitor is located at the plant vent downstream of the isolation valves. Upon detection of a high radiation signal, the butterfly purge valves are automatically closed, isolating the containment.
l 15-1
l Due to the location of the radiation monitor downstream of the isolation I valves, a portion of any radioactivity would be released to the atmosphere before the isolation valves can close.
l The applicant has performed an analysis of a fuel handling accident inside 1
i containment.' In this analysis, he assumd that any activity released to l l i the containment above the refueling cavity pool would be well mixed in a j volume about 40 feet high and extending around the pool cavity and transfer canal. Due to the location of these inlets relative to the refueling 1
cavity pool and transfer canal, we concur with the applicant that substan-tial mixing of any released activity would occur in the containant prior to its release from the plant vent. There would be additional dilution of contaminated air since a recirculation system exhausts between 10 l l l percent and 50 percent of the air flow and returns the remainder to the l I
containment atmosphere. The applicant has conservatively omitted this
)
effect in his analysis. This would further reduce the dosec shown in {
l Table 15.1.
The applicant's proposed technical specification limit on closure time for the purge isolation valves is 10 seconds. An a$ditional 9.0 seconds of l transit time and instrument response time would elapse between the time the leading edge of any activity passed the isolation valves and the time the valves would begin to close. Therefore, the valves would close, isolating the containment, within about 19 seconds after initially detecting activity leaving the p1 tnt vent.
15-2 wo J
We have evaluated the applicant's analysis and performed an independent assessment of a postulated fuel handling acci6ent inside containment.
We have assumed a valve closure time of 10 secords and a containment isolation time of 19 seconds. Our assumptions are described below and the calculated doses are shown in Table 15.1. We have concluded that, with the present plant systems, the radiological consequences in the event of a fuel handling accident inside containment would be well within the guideline values of 10 CFR Part 100, and therefore, acceptably low.
15.2 Design Basis Accident Assumptions 15.2.2 Fuel Handling Accident The assumptions used in our evaluation to calculate the offsite doses l from a refueling accident inside of containment are:
(1) Rupture of all fuel rods in one assembly.
(2) All gap activity in rods, assumed to be 10 percent of the noble gases and 10 percent of the iodine.
(3) Peaking factor of 1.65.
(4) The accident occurs 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after shutixun.
(5) 99 percent of the iodine is retained by the pool water.
(6) Mixing in 33,600 cubic feet of containment air (0.0126 of the total containment volume).
(7) Exhaust rate of 229 cubic feet per second of contaminated air to the envirornent, contained in 917 cubic feet per second of con-tainment purge flow.
(8) Isolation valve closure time of 10 seconds.
(9) Containment isolation time of 19 secords.
(10) 0-2 hour relative concentration values (X/Q) determined from onsite meteorological program.
15-3 e , . -w .w .., - > - _e ..m_ w. e e.- , o
_________.__________m___._m__
8
. _ _ _ .. . . _ . . . _ . . _ . . . . . . . _ . . _ . . _ _ . . _ . . . _ . _ _ . _ ~ _ . i TABLE 15.1 PCTTENTIAL OFFSITE DOSES DUE 'IU DESIGN BASIS ACCIDENTS l
l l
l - Two Hour Course of Accident l Exclusion Boundary Iow Population Zone 1 (800 meters) (9600 meters) j Thyroid 'Whole Bo3y
Fuel Handling (In Containment) 20 less than 1 1 less than 1 1
i l
15-4 1
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i 18.0 REVIEW BY THE ADVISORY CO W R TEE ON REACIOR SAFEGUARDS-(ACRS)
The Advisory Committee on Reactor Safeguards completed a partial review of the Diablo Canyon Operating license application in June 1975, and the Commit' tee's report on this partial review was attached as Appendix B to Supplement No. 3 to the Safety Evaluation Report. In its report, the Committee stated that it would complete its review of the seismic design bases, adequacy of the seismic design, protection against tsunamis and certain other matters after the NRC staff's review of seismic-related topics was completed.
1 The Comittee's report also contained other comments and recommendations. )
The actions we have taken or plan to take in response to these coments and recommendations are described in the following paragraphs.
l
)
- 1. The Comittee stated that the results of tests and analyses for 17 x 17 fuel-assemblies should be evaluated fully by the NRC staff, and resolved to its satisfaction, prior to the full core use of 17 x 17 fuel to produce power. These incitr3ed: Fuel assembly flow tests, DNB tests for non-uniform heat flux, and the effect of fuel rod bowing on INB after the first fuel cycle.
Our evaluation of the fuel assembly flow tests is completed as discussed in Section 4.2.1 of Supplement No. 3 to the Safety Evaluation Report. Our evaluation of DNB and fuel rod bowing is empleted as discussed in Section 4.4 of this supplement.
18-1 w ___ _ _ _ _ - _ _ _ _ _ _ _ _ _ - _ _ _ - _ _ _ _ - - - _ - _ ._ _ . _ - _ - _ _ _ _ ___ ___
I
- 2. The applicant's discussions with the Committee incitr$ed a presen-tation by Westinghouse concerning an augmented startup prograe proposed for implementation in some of the first Westinghouse reactors with full core employment of the 17 x 17 fuel assembly.
The Cmmittee recommended that the NRC staff evaluate the results of the augmented startup prograin as well as overall operating ex-perience with large, high power-density reactors, prior to sustained operation at full power.
Augmented startup program tests have been performed on two of the first plants using 17 x 17 fuel assemblies, Trojan and Beaver Valley. The licensees have sutnitted reporL on these tests. Further results will be published by Westinghouse in the near future. Preliminary indications are that the core peaking factor varied surprisingly little from the steady state value during load following using constant axial offset control. This is also true for results of tests in 15x15 cores.
l As a part of the generic evaluation of 17 x 17 fuel, we insisted on develognent of the portion of the tests which are designed to show a ccxnparison between calculated and measured power distribution. This position was stated in Section 18.2.3 of Supplement No. I to the Safety Evaluation Report for the Beaver Valley Power Station, Unit No.1 (Docket No. 50-334). It is important to note, however, that while we required these tests, we view them as confirmatory in nature. Thus, we are not 18-2
______.m__.
R making our evaluation and acceptance of the results of the test program a condition on the further operation of the plants utilizing 17 x 17 fuel. This is consistent with our practice of reviewing startup test reports in parallel with the operation of the plant.
We feel this arrangement has proven satisfactory.
- 3. The Comittee stated that the NRC staff should review the effective-ness of the proposed method cf constant axial offset control in pro-tectirig against adverse consequences of postulated reactor transients and accidents. The Committee wished to be kept informed.
We considered constant axial offset control in connection with our generic review of 17 x 17 fuel as presented in the Westinghouse Topical Report WCAP-8185, " Reference Core Report 17 x 17," July 26, 1974. We have concluded that this method of control can be an effective method of controlling a reactor so that the maximum peaking factor which occurs in the core will never exceed the maximum allowable.
Accordingly, we have found the control method acce?able for use at Diablo Canyon and other reactors now operating or soon to be operated.
Further, we are performing independent analyses. Our consultants at Brookhaven National Laboratory have completed an' independent evaluation 7 of the beginning-of-life analysis of constant axial offset control (A. Burlik l i
1 et al, " Power Peaking During Ioad Following Using Westinghouse Constant Axial Offset Power Distribution Control," SNL-NUREG 22477, January 1977). This evaluation confirmed the Westinghouse results. End of life analyses will be performed in the future.
1 18-3 ,,
1
.. _ , _ _ _ . . ._ _ _ . _ . . _ _ _ ~ . _ . . _ _ . . _ _ . _
8
- 4. The Comnittee stated that the results of prototype 17 x 17 fuel i J
assembly irradiations should be followed closely. l Two prototype 17 217 fuel assemblies are installed in each of the l
Surry reactors. Inspection after two fuel cycles in Unit I and one fuel cycle in Unit 2 has revealed no amxnalies. We will continue to follow the results of these prototype irra3iations.-
The 17 x 17 fuel assemblies to be used in full core implementation 5.
~
of the 17 x 17 fuel design have eight spacer grids, as contrasted with the two prototype 17 x 17 assemblies in each of the Surry reactors (discussed above) which have seven spacer grids. Ib11owing each cycle of operation, the 17 x 17 fuel assemblies with eight spacer grids (in the Diablo Canyon reactors and other reactors) will be examined for fuel rod integrity, fuel rod and assembly dimension l and alignment and surface deposits. In view of the fact that the 17 x 17 fuel array is a new design and that no prototype irradiations are planned for 17 x 17. fuel containing eight spacer grids, the Committee stated that the results of surveillance programs for this type of fuel should be followed closely. The Committee wished to be kept informed.
The fuel surveillance program for 17 x 17 fuel assemblies with eight spacer grids is described in further detail in Section 4.2.1 of Supplement No. 2 to the Safety Evaluation Report. We will follow this program closely and, when results become available, we will keep the Comnittee informed.
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- 6. The Comittee wished to be kept informed concerning the repalts of varicus ongoing 17 x 17 fuel test and analytical programs, and my design changes which might be proposed in the futute.
Our actions concerning several specific 17 x 17 fuel test and analytical programs are described above. We vill keep the Comittee infonned concerning the results of the various 17 x 17 fuel test and analytical programs and any design changes which might be pro-posed in the future.
- 7. The Committee wished to be kept informed concerning the reevaluation of emergency core cooling system performance and operating 1dmits l and procedures for power distribution monitoring.
(
Our evaluation of this matter is completed as discussed in Sections 4.4 and 6.3 of this Supplement.
- 8. The Committee stated that the NRC staff review of Anticipated Tran-sients Without Scram should be ccnnpleted and the matter should be resolved in a manner satisfactory to the NRC staff ard to the Committee. i The status of this matter is described in Section 7.2.5 of this Supplement. We wil'2 continue our review and will require that any changes indicated to be needed by the results of approved analyses to be incorporated into the Diablo canyon plant.
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- 9. The Committee stated that the matter concernin; protection against tornade missiles should be resolved in a manner satisfactory to the ;
I NRC staff.
Our evaluation of this matter is continuing. We will report the final resolution in a future supplement to the Safety Evaluation i Report. )
- 10. The Committee stated that the matter concerning environmental qualification of Class I instrumentation and electrical equipnent should be resolved in a manner satisfactory to the NRC staff knd i tne Cocmittee.
Our evaluation of this matter is continuing. We will report the l final resolution in a future supplement to the Safety Evaluation Report.
- 11. The Committee stated that generic problems, relating to large water reactors, should be dealt with appropriately by the NRC staff and the l applicant ac suitable approaches are developed.
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l The status of each of these items is discussed in Appendix 3 to this supplement.
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21.0 FINANCIAL PROTECTION AND INDEPHITY REQUIREMENTS 22.1 Preoperational Storace of Nuclear ' Fuel In the Safety Evaluation Report we described financial protection and i
indemnification requirements that the applicant would be required to meet in connection with preoperational storage of nuclear fuel. Since then, the applicant has met these requirements to the extent necessary to obtain fuel storage licenses pursuant to 10 CFR Part 70 and has received such fuel storage licenses for Unit I and Unit 2. The applicant will continue to be required to maintain the necessary financial protection and to pay annual fees in connection with the indemnification agreements as described in the Safety Evaluation Report..
21.2 Operatino License In the Safety Evaluation Report we described the financial protection and indemnity requirements for the Diablo Canyon Nuclear Power Station, Units 1 and 2.
By publication in the Federal Register Voltrne 42, Number 74, April 18, 1977, 10 CFR Part 140, " Financial Protection Requirements and Indemnity ,
Agreements," was amended to increase the amount of primary financial protection from private sources required for facilities having a rated capacity of 100 electrical megawatts or more to S140 million, effective May 1,1977. ($110 million was the figure discussed in the Safety Evalua-tion Report.) The requirements remain the same as discussed in the Safety Evaluation Report except for the different amount of required protection from private sources.
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l 21.3 Conclusion On the basis of the above considerations and those identified in the Eafety Evaluation Report, we reaffirm our conclusion that the presently applicable requirements of 10 CFR Part 140 have been satisfied and that prior to issuance of any operating license, the applicant will be required to comply with the previsions of 10 CFR Part 140 ' applicable to operating licenses, including those as to goof of financial 1
protection in the requisite amount and as to the execution of an appropriate indemnity agreement with the Ccanission. <
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.c... - .__. . ._ - _.. _ _ _ _ :.._._... 1 22,0 CONCLUSIONS In Section 22 of Supplement No. S to the Safety Evaluation Report, we stated that several items were still outstanding, and that faverable resolution of these items would be required before operating licenses for Diablo Canyon Units l and 2 C0uld be issued. Resolutions for a ntrter of those items have been presentad in this supplement. Items which currently remain -
outstarx31ng are sunesized below.
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- 1. An evaluation of the plant's capability to withstand an earthquake of magnitude 7.5 on the HosGri fault (Section 3.7 of Supplement No. 4 and
)
Supplement No. 5).
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- 2. An evaluation of the enviroignental and seismic qualification of Category l I electrical, instrumentation and control equipnent (Sections 3.10 and 7.8 of the Safety Evaluaticn Report).
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- 3. An evaluatior3 of the effects of postulated pipe breaks outside contain- -
ment for Unit 2. However, the evaluation has been complet2d for Unit 1 (Section 3.6 of this Supplement) . l 1
- 4. An evaluation of the means of protecting the reactor coolant system j from overpressurization trangients at low temperature for Unit 1 in the I long term (after the first fuel cycle) and for Unit' 2._ However, the i i
evaluation has been completed for the short term provisions for Unit 1 ;
i (Section 5.2.2 of this Supplement). I 22-1
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- 5. An evaluation of the details for implementing industrial security provisions ~ (Section 13.6 of this Supplement) .
- 6. An evaluation of the vulnerability of the electric power systems and equipnent to a degraded grid voltage condition (Section 8.0 of this l Supplement).
- 7. An evaluation of a postulated main steam line break inside containmen;t -l (Section 6.2.1 of this Supplement).
- 8. An evaluation of the details of modifications to the diesel generator compartment ventilation system (Section 9.5.5 or this supplement).
- 9. An evaluation of the plant's tornado missile protection (Section 3.5 of Supplement No. 3 to the Safety Evaluation Report).
Subject to favorable resolution of the outstanding matters described above, the conclusions as stated in Section 22 of the Safety Evaluation Report remain unchanged.
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APPENDIX A j CONTINUATION'OF THE CHRONOLOGY-OF THE RADIOILGICAL SAFEIT REVIEW l
l September 30, 1976 Letter from applicant providing information about schedule for A'IWS analysis requested on July 2,1977 l l
Septemtier 30, 1976 Letter to applicant requesting reevaluation of fire protection capabilities October 11, 1976 ACRS Subemmittee meeting in Ios Angeles, California to discuss seismic design i October 27, 1976 Meeting with applicant to discuss qualification of I Class IE electrical equipment October 27, 1976 Letter from applicant providing schedule for sutmitting fire protection analysis requested on September 30, 1977 l
November 13, 1976 ACRS full comittee meeting on Diablo Canyon in Washington, D. C. to discuss seismic design December 3, 1976 Letter to applicant requesting additinal information l
about containment structural integrity test December 7, 1976 Public Hearings on Envirorzental review in San Luis to Obispo, California (incitding routine release of December 17, 1976 radioactivity - Appendix I to 10 CFR 50)
December 17, 1976 Letter to applicant providing guidance on fire protection reevaluation l
December 20, 1976 Letter from Executive Director of ACRS providing coments on seismic design basis December 28, 1976 Meeting with applicant to discuss seismic design and reevaluation December 29, 1976 January 5,1977 Meeting with applicant to discuss seismic design reevaluation and ACRS cuments.
January 7, 1977 Sutnittal of Amendment 46 including (1) changes to quality assurance program, (2) addition of autmatic circulating water ptnp trip and (3) miscellaneous changes l
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January 10, 1977 Letter to applicant requesting information aboat reactor coolant system overpressurization protection January 19, 1977 Letter from applicant subnitting report entitled, "Near-Field Ground Motion Simulation for a Vertical Fault with Dip-Slip" February 4, 1977 Meeting with applicant to discuss seismic design -
. criteria for reevaluation of structures February 10, 1977 Letter from applicant about the schedule for sutraitting information on reactor coolant system overpressurization protection requested on January 10, 1977i February 18, 1977 Letter to House Subcommittee on Oversight and Investigations about Diabic Canyon seismic design licensing situation February 25, 1977 Letter to applicant providity guidance on complyirg with new security regulation 10 CFR 73.55 l
March 11, 1977 Letter to applicant requesting information about pcstulated fuel handling accident inside containment March 16,1977 Letter frem applicant providing additional information about containment structural integrity test requested on December 3, 1976 l
March 18,1977 Letter to California Energy Commission about the i possibility of an interim operating license March 31,1977 Letter to House Subecmmittee on Energy and the Environment about tle Diablo Canyon seismic design licensing situation March 31, 1977 Subnittal of Amendment 47 including (1) information about the qualification of Class IE electrical equipnent-including a report on subnerged equignent, (2) a reanalysis of ECCS performance with upper head temperature at T(hot) and (3) miscellaneous changes April 7,1977 Letter to applicant requesting information about i instrunent trip setpoint values April 11, 1977 Letter from PG&E expressing intention to apply for an interim operating license April 20,1977 Meeting with intervenor's consultant to discuss Diablo Canyon quality assurance program A-2
l April 21, 1977 Subnittal of Revision 1 to physical security plan April 28, 1977 Letter from applicant providing schedule for subnitting information about instrtunent trip setpoint values requested on April 7,1977 April 29, 1977 Meeting with applicant to discuss seismic design reevaluation-systems needed for safe shutdown April 29, 1977 Subnittal c' Amendment 48 including (1) minor revisions to the electric ;ower systems, (2) minor revisions to the meteorology program, (3) information about reactor coolant system overpressurization I
protection and (4) miscellaneous changes l May 3,'1977 Meeting with applicant to discuss seismic design
! reevaluation-interim operating license May 4, 1977 Letter to applicant providing guidance on complying with new security regulation 10 CFR 73.55 May 12, 1977 Prehearirg Conference in Ics Angeles, California to discuss safety t:ontentiions and hearing schedule May'23, 1977 Letter to applicant requesting information about inservice inspection progran under 10 CFR 50.55a(g)
May 25, 1977 Subnittal of Revision 2 to the security p1'an l May 27, 1977 St mittal of Amendment 49 incitriing (1) minor i revisions to tornado protection analysis, (2)
I minor revisions to information about qualification of Class IE electrical equipnent, (3) "tinor
, additions to description of ECCS performance l evaluation, (4) a detailed analysis of a postulated fuel handling accident incide containmerit and (5) miscellaneous chages May 27, 1977 Letter to applicaat providing staff ccmr,ents on Revision 1 to the security plan June 1,'1977 Letter to applicant requesting information about postulated steam line break inside containment 1
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June 2,1977 Meeting with applicant' to discuss seismic reeval-uation and interim licensing June 3, 1977 Submittal of Revision 3 to the security plan June 3,1977 Letter to House Subecmittee on Oversight and Investigations answering additional questions on Diablo Canyon seismic design licensing situation June 5, 1977 Submittal of Amendment 50 including (1) partial results of seisnic reevaluation of the plant, (2) probability studies on earthquakes and (3) other seismic studies and responses to ACRS coments and questions l June 6, 1977 Letter to applicant requesting information about degraded grid voltage June 7,1977 Letter to applicant requesting information about postualated pipe breaks outside contairr.ent l
l June 7, 1977 ,
Meeting with applicant to discuss the staff review l of the security plan June 10, 1977 Letter from applicant providing information'about ECCS" performance evaluation !
June 15, 1977 Submittal of Revision 4 to the security plan June 16,1977 Letter from applicant providing schedule for sub-mitting information about steam )ine break inside ,
containment requested on June 1,1977 l June 16, 1977 Letter from applicant revising schedule for submitting information about instrument trip setpoint values requested April 7, 1977 June 21, 1977 ACRS Subcommittee meeting in Ics to Angeles, California to to discuss seismic design June 23,1977 .
June 29, 1977 Submittal of Revision 5 to the security plan June 30,1977 Congressional hearings on Diablo Canyon seianic design licensing situation-House Subcommittee on Energy and the Envirorsnent A-4 Lm-__
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l July 1,1977 Letter from applicant providing schedule for subnitting information about degraded grid voltage requested on June 6,1977 July 5,1977 Letter from applicant submitting 'informatio about protecting the reactor coolant system from over-pressurization transients July E, 1977 Letter from applicant subnitting information about postulated pipe breaks outside of containment July 6, 1977 Letter from applicant subnitting information about the tornado wind design of steel siding for the cable spreading rooms and switchgear rooms in the turbine building 1 July 8, 1977 Letter from applicant submitting information about the analysis of a fuel harxiling accident inside containment d
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APPENDIX B ADVISORY CCMiITTEE ON'REACIOR' SAFEGUARDS GENERIC ITEMS The Advisory Committee on Reactor Safeguards (Comittee) periodically issues a report listing various generic matters applicable to large
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light-water reactors. The most recent such report was issued on February 24,.1977. In addit. ion, the NRC staff periodically reports on the status of its efforts to resolve these generic items. The !
latest staff status report was issued on January 31,1977.
These are items which the Committee and the NRC staff, while finding present plant designs acceptable, believe have potential for adding to overall safety margins and so should be considered for application to the extent reasonable and practicable as solutions are found, recognizing that such solutions may occur after completion of a specific plant. This is consistent with our continuing efforts to reduce the already small saf4ty risk from nuclear power plants.
The current status of each of these generic items as it relates to Diablo Canyon is indicated below. The ntsnbering corresponds to that in the February 24, 1977 report of the Committee.
Group II
- Resolution Pending (1) Turbine-Missiles This item is under generic review. The staff believes that the ,
risk from turbine missiles at Diablo Canyon is small. As part of the generic review we are evaluating the effectiveness of B-1
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measures such as frequent valve testing and turbine rotor material inservice inspection in reducing the risk '(Staff status report l I
dated January 31, 1977.
(2), Effective Operation of Containment Sprays on IDCA This item is resolved for Diablo Canyon by use of sodium hydroxide additive to the contairinent sprays (SER Section 6.2.3).
The generic review continues, including reviews of alternate l additives that may be preferable (Staff status report dated January 31, 1977).
(3) Possible Failure of Pressure Vessel Post-IDCA by Thermal Shock j l
This item is resolved for Diablo Canyon by employment of an acceptable vessel design (SER Section 5.3 and Begulatory Guide 1.2) .
The research program i's continuing and, as indicated in our status report dated January 31, 1977, the work done to date suggests that flawed, irradiated reactor vessels subjected to thermal shock from IDCA-ECCS water will not fail catastrophically. Further, as indicated in Regulatory Guide 1.2 and" Section 5.3 of the SER, if it should be cencluded that the margin of safety against reactor vessel brittle failure due to DCCS operation at any time during versel life is unacceptable, an engineering solution, such as annealing, could be applied.
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i (4) Instruments t_o Detect (Severe) Puel Failures l
This item is resolved for Diablo Canyon by employment of a l I
failed fuel detection system (FSAR, Section 7.7.1) . ]
l, The generic review /research program is cor.tinuing as discussed in our status report dated January 31, 1977. The Comittee's 1 1
- status report, dated February 4,1977, irdicates that this item is resolved for limited fuel failures but that more
- ' work is needed for the severe failure case to establish ,
l instrumentation criteria. l 4
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(5) Monitoring for Excessive Vibration or Loose Parta Inside the Pressure Vessel l
This item is resolved for Diablo Canyon by employment of a loose parts monitor (SER-Section 5.4) .
The generic review is continuing and will establish criteria for such systems as irdicated in our status report dated January 31, 1977.
(6) Comon Mode Failures
'Ihis item is under generic review as indicated in our status report dated January 31, 1977.
(7) Behavior of Reactor Fuel Under Abnormal Conditions A generic research program is underway as fIdiCated in our status report dated January 31, 1977.
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l (8) _BWR Recirculation Pump Overspeed Daring IDCA l This item is not applicable to Diablo Canyon, which is a I
pressurized water reactor facility.
(9) The' Advisability of Seismic' Scram As indicated in our letter of May 19, 1977 to the Comittee,
- I the staff considers the generic studies to be completed and j does not plan to require the installation of seismic scram i devices on ecmercial power reactors.
However, the advisability of such a device for Diablo Canyon is still being discussed by IG&E, the NRC staff and the Comittee, t
(10) Energency Core Cooling-System Capability-for Future Plants This item does not apply to Diablo Canyon, which is an existing plant.
Groop-IIA-Resolution Pending ' Items-Since December 18,1972 (1) Control Rod Drop Accident-(BWR's)
This item does not. apply to Diablo Canyon which is a pressurized water reactor facility, (2) Ice' Condenser Containments This item does not apply to Diablo Canyon which does not employ an ice cordenser containment.
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e (3) Rupture of High Pressure Lines Outside Containment We expect this item to be resolved for Diablo Canyon by employment of our current acceptance criteria in our review, which is not yet completed. (Section 3.6 of this supplement) .
The staff considers the generic item to be acceptably resolved as indicated in our status report dated January 31, 1977.
(4) PWR Pump Overspeed During a IDCA l
This item is under generic review as indicated in our status report dated January 31, 1977 and in Section 5.2.6 of the SER.
(5) Isolation of Low Pressure from High Pressure System This item is resolved for Diablo Canyon by employment of acceptable design measures. The Residual Heat Removal System overpressurization interlocks are discussed in Section 7.6 of the Safety Evaluation Report. Ieak testing of the emergency core cooling system check valves is specified in the ASME Boiler ard Pressure vessel Code,Section XI, and thus will be a requirement of the Technical Specifications (Section 5.2.8 of this Supplement).
The NRC staff considers the generic item acceptably resolved. However, 1
there is a continuing effort incitxling development of an ANSI Standard on the subject (Staff status report dated January 31, 1977).
(6) Steam Generator Tube Leakage 1
This item is resolved for Diablo Canyon by employment of acceptable j measures. These include the use of volatile secordary chemistry B-5
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control and titanitn condenser tubes to prevent steam generator tube degradation. They also include inspections according to.
Regulatory Guide 1.83 and monitoring of steam generator tube l 1eakage and secondary system radioactivity to detect any possible degradation in 'a timely manner.
I The generic review is conti'n uing as discucsed in our status report dated' January 31, 1977.
(7) ACRS/NRC Periodic 10-Year Review of all Power Reactors This item is urder generic review as indicated in our
]
status report dated January 31, 1977.
Group I'!B Resolution Pending - Items Added Since ' February 13, 1974 (1) Computer Reactor Protection System i
This item does not apply to Diablo Canyon which does not l employ this type of reactor protection system.
1 I
(2) Qualification of New Fuel Geometries This item is resolved for Diablo Canyon by employment of an acceptable fuel design and surveillance program. The evaluation, testing and surveillance for the 17 x 17 Diablo Canyon fuel design.are summarized in items 1 through 7 of Section 18.0 of this supplement in response to specific Connittee connents in these areas.
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The generic review is continuing as discussed in our status report to the Comittee dated January 31,1977. (Also see 1 1
item II.7 above concerning research into fuel behavior under l l
abnormal conditions). l l
(3) Behavior of BWR Mark III Containments This. item is not applicable to Diablo Canyon which is a 1
pressurized water reactor facility. l i
'(4) Stress Corrosion Cracking'in BWR Pioing This item is not applicable to Ciablo Canyon which is a pressurized water reactor facility. l Group IIC Resolution Pending - Items Added Since March 12, 1975, (1) Lockina Out of ECCS Power Operated Valves This item is resolved for Diablo Canyon by employment of acceptable measures which emply with Branch Technical Position l
l EICSB 18 from Appendix 7A to the Standard Review Plan (Section l 6.3 of Supplements 4 and 5 to the SER).
l l
The generic review is continuing as indicated in our status report dated January 31, 1977.
(2) Design Features to Control Sabotage We expect item to be resolved for Diablo Canyon by employing our current acceptance criteria in our review, which is not yet completed (Section 13.0 of this supplement).
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The generic review is continuing as indicated in our status report dated January 31, 1977.
(3) Decontamination and Decommissioning of Reactors This item is under generic review as indicated in our status report to the Committee dated January 13, 1977.
(4) Vessel Support Structures This item is under generic review as discussed in our status report dated January 31, 1977.
We have concluded that licensing plants for operation is acceptable while our review continues (Section 5.2.1 of Supplement No. 4) .
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In addition, we have requested that the applicant complete the appro-priate analyses for Diab'.o Canyon as soon as possible and we will l report on the progress of these analyses in future supplements to the Safety Evaluation Report.
(5) Water Hammer This item is resolved for Diablo Canyon by employment of acceptable measures to prevent feedwater system water hamer. Diablo Canyon employs a ccccination of (1) modifications to the feedwater system and steam generators, (2) procedural limitations and (3) testing to ensure the design and procedures will be effective in eliminating feedwater system water hamer (SER Supplement 4, Section 10.4) .
The generic studies are continuing as indicated in our status report to the Comittee dated January 31, 1977.
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e (6) Maintenance and Insoection of Plants This item pertains to design improvements for future plants. Accord-ingly, it does not apply to Diablo Canyon which is an existing plant.
(7) Behavior of BWR Mark I Containments
. This item does not apply to Diablo Canyon which is a pressurized water reactor facility. l Grouc IID Resolution Pending - Items Added Since April 16, 1976 (1) Safety Related Interfaces Between Reactor Island and Balance-of-Plant This item pertains to standard plants. Accordingly, it does not apply to Diablo Canyon which is not a standard plant.
(2) Assurance of Continuous Long-Term Capability of Hermetic Seals on Instrumentation and Electrical Equirxnent This is a new item that was not addressed in our status report dated January 31, 1977. The generic matter will be discussed in a future staff status report.
Our evaluation of the qualification of Class IE electrical, instru-mentation and control equipnent for Diablo Canyon is not yet completed (SER Sections 3.10 and 7.8). However, the proper design and qualification testing L2 such equipnent is included in our evaluations and will be considered. The Quality Assurance program for Diablo Canyon provides for the use of approved procedures and checks in performing maintenance on such equipnent. We have l approved the quality assurance program (Section 17.0 Safety l
Evaluation Report Supplement No. 3).
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t_________ _______________________m----____-..___-. . _ _ _ _ _ - . _ . _ _ _ _ . ._____-.-_____.--._.____m
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JUN 2 9 sn Docket Nos. 50-27 M i and 50-323 ,
k APPLICANT: Pacific Gas & Electric Company (PG&E) !
I FACILITY: Diablo Canyon Nuclear Power Station, Units 1 and 2 (Diablo Canyon)
SUM 4ARY OF MEETING HELD ON JUNE 2, 1977. TO DISCUSS DIABLO CANYON SEISMIC DESIGN l
We met with the applicant on Juhe 2,1977, to discuss the seismic design l of Diablo Canyon. A list of attendees is provided in the enclosure. i BACKGROUND Diablo Canyon had originally been designed to withstand an earthquake with a horizontal ground acceleration of 0.4g, based on the geological investigations that had been conducted in connection with the construction pemit review. During the operating license review, which was in pro-Jress, we had requested that PG8E reevaluate the plant's seismic capabilities to detemine what modifications might be necessary to ensure that the plant ; l could withstand a more severe earthquake with a horizontal ground acceleration -
of 0.75g, based on newer geological information, PGAE was performing such a reanalysis. ;
PG&E had also expressed its intention.to apply for an interim operating license for Diablo Canyon to allow plant operation while the seismic !
reevaluation wasyttiircompleted. The material needed to justify such {
an interim licehse application had previously been outlined by the NRC staff as follows: .
i i
/ f (1) A demonstration of the need to consider such an action, and i
' I :
(2) Information and analyses to demonstrate that the requisite level of !
safety would be assured during the period of the interim license. t This information should include all available results of the seismic reassessment program, supported by: !
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F Pacific Gas & Electric Company -2 JUN 2 8 WU (a) a realistic assessment of the probability of large earthquakes in the site environs and the probability of the plant to with-stand such earthquakes without failures of structures and equip-ment sufficient to lead to unacceptable radiological consequences to the public; (b) a comitment to make any changes to the design determined to be necessary on the basis of the continuing seismic reassessment program; and (c) an avaTction of the practicality of making the need changes '!
to a p'. ant which has been in operation during the term of the interim license.
INTERIM LICENSE APPLICATION - RELATIONSHIP TO REGULATIONS At a previous meeting on May 3, 1977, we had discussed the question of whether or not the interim license application would constitute a request for an exemption to the NRC Regulations, in particular to appendix A to 10 CFR Part 100 and to Criterion 2 of the General Design Criteria (Appendix A to 10 CFR Part 50).
At this meeting, June 2,1977, we informed PG&E of our opinion on this subject as follows:
(1) Regardless of whether or not the interim license was to be considered ,
an exemption from, en exemption to or a waivet of the regulations, the information needed to support the application, outlined above, would be substantially the same. The fundamental criterion in any approach would be that the plant must be shown to have an acceptable level of safety before an operating license would beitssued.
(2) We had considered three possible ways of stating the interim license application:
(a) the first possib /1tuy would be a p tition for waiver of a part'cular rule under f0 CFR Part $n758 , exeihton of.
The sole basis 'that would be allowed here was that due to special circumstances application of the particular rule would not serve-the purpose for which the rule was intended.
In this case;"the regulations explicitly spelled out the subsequent procedures to be followed by the Licensing Board, s
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Pacific Gas & Electric Company - 3-M 29 577 (H The second possibility would be request for an exemption from the requirements of- the regulations under 10 CFR Part 50.12.
Here the regulations indicated that the requested exemption would have to be shown to:
(1) be authorized by law, (ii) not endanger life or property or the cosmon defense and security,and; (iii) be othenvise in the public interest.
In this case, the subsequent procedures to be followed by the Licensing Board were not ex'plicitly spelled out.in the regulations. ,
(c) The third possibility would be a showing that the requested action would not constitute a deviation from the. regulations in that the pertinent. regulations already contemplated and allowed for an ' applicant to propose and justify alternate approaches.' Language to this effect was contained in Appendix A to 10 CFR Part 50,10 CFR Part 100 and Appendix A to 10 CFR Fart 100.
(3) In any of these methods the oversiding and fundamental consideration would be whether or not an acceptable 1esel of safety had been shown.
(4) In any of these methods we would went any specific passage or section -
of the regulations that might not be met to be clearly identified and the reasons end justifications to be clearly stated.
(5) We indicated that 10 CFR Part 50.57, " Issuance of operating license",
would be cited in connection with any of the three methods discussed above.
(6) The staff would not spebify which method of ctating the application '
should be used. This decision would be left to PG&E.
INTERIM LICENSE APPLICATION - STATUS OF SYSTEMS REANALYSIS
'At a previous meeting, on April 29, 1977, PG4E had described to us those systems and portions of systems for which they intended to complete the reanalysis prior to applying for an interim license. In general, these consisted of the systems that PG4E considered necessary to ensure that the plant could be safely shutdown following a major earthquake. At this o,.,c .
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y Pacific Gas & Electric Company JUN 2 81977 meeting, on June 2,1977, we informed PG&E that we cmsidered the pro-bability study to be the primary tool that would be used in judging whether or not the plant would have an acceptable level of safety for an interim operating license. Accordingly, we did not consider that the interim license would be contingent upon whether or not the reanalysis had been completed for any particular system or part of a system at the new earthquake level of 0.759 SEISMIC REEVALUATION - C0ku11 NATION OF LOADS In order to justify a full term operating license, PG&E would need to complete the seismic reevaluation at 0.759 We informed PG&E that in performing this reevaluation they shguld combine the calculated loads resulting from a postulated loss-of-coolant accident with the calculated loads resulting from the postulated earthquake at 0.75g. These loads should be combined by direct addition, as is the usual practice in nuclear plant design, rather than by using the square root of the sum of the squa ns. PG&E would then be expected to demonstrate to us that structures, systems and components important to safety can perform their required safety functions under the combined loading conditions.
If, for any particular item, functional capability could not be demonstrated l and PG&E should believe that modifications to demonstrate functional !
capability would be impractical or unwarranted, then PG&E would be expected to describe the situation fully and to justify its acceptability.
Original Signed By 4 Dennis P. A!Ilson Dennis P. Allison, Project Manager Light Water Reactors Branch No. 1 Division of Project Management l
Enclosure:
Attendance List cc w/ enclosure: /
See Page 5 l I
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Pacific Gas & Electric Corgany cc Philip A. Crane, Jr., Esq. Mr. Jonn Forster Pacific Gas and Electric Company 985 Palm Street 77 acale Strect San I,uis obispo, California 93401-San Francisco, California 94106 Mr. William P. Cornwell l
An:frew J. Skaff, Esq. P. O. Box 453
, California Public Utilit ies commission Morro Bay, California 93442 350 t'ollister Street San Francisco, California 94102 Mr. James 0. Schuyler, Nuclear ~ Project Engineer Mr. Frederlen Eissler, President. Pacific Gas & Electric Company.
i scenic Shoreline Preservation /77 Beale Street Conference, Inc. San Francisce, Galifornia 94106 4623 More Mesa orive Santa Baroars, california 93105 Mrs. Thelma Hirdler 811 Fair Oaks Avenue Ms. Sandra A. Silver Arroyo Grande,. California 94420 5055 Radford Avenue North Hollywood, California 91607 Mr. W. C. Gangloff Wagtir.ghouse Electric Corporation Mr. Gordon A. S11ver P.~0. Box 355 5055 Badford Avenne Pit t sburgh, Pennsylvan'ta 15230 North Hollywood, California 91607
'lale I. Jones, Esq.
Paul C. Valentine, Esq. 100 Van Ness Avenue 400 Channing Avenue 19th Floor i Palto Alto, California 94301 San Francisco, California 94102 P.s. Raye flecing David F. Fleischaker, Esc.
1746 Charro Street 1025 15tn Street, d.-W.
San Luis Obispo, California 93401 Washington, D. C. 20005
, 1 Neil Goldbarg, Esc;. '
Ms. Elizabeth E. Apfelberg Wilmer , Cutler & Pickering 1415 Caza6aro 1666 K Street, N. W. San Icis Ooispo, California 93401 Washireton, C. C. 20006 .
H. H. Hewit. ark Consulting fagineering Services j 1211 Civil Engineerfrg Building i University of Illinois j
- Urbana, Illino!s til801 l
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! NRC PORM $18 (9 76) NRCM 0240 W u. s. ooVERNMENT PRINTING OFFICES 1970 m ess.ead l
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ENCLOSURE ATTENDANCE LIST PACIFIC GAS 8 ELECTRIC COMPANY (PG&E)
JUNE 2, 1977 4
i PACIFIC GAS & ELECTRIC COMPANY
'M. Furbush i P. Crane J. Hoch
- J. Gomly ,
WESTINGHOUSE W. Gangloff 1 T. Esselman NUCLEAR REGULATORY COMMISSION _--
E. Case R. DeYoung -
I. $1hwe11 D. Jeng D. Vassallo P. Kuo J. Stolz D. Allison J. O'Brien J. Knight A. Fratoni !
R. Dosnak W. Gammill P. Chen M. Grossmen T. Sullivan J. Tourtellotte L. Davis
/
~ CENTER FOR LAW IN '?HE PUBLIC INTEREST (Interves A Counsel) _
D. Fleishaker 4~.
. ~ . .
- i. .
)
-70 EG 0240 ;
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- u. s. oovenwueur eneurswa orrecei teve . see ea4 - - - - - - - - - ~ ~ ~
1 MEETING
SUMMARY
Docket File ( N )'
NRC PDR Local PDR TIC NRR Reading l LWR 1 File 1 E. G. Case R. S. Boyd 1 R. C. DeYoung J. Stolz ,
K. Kniel
- 0. Parr S. Varga -
L. Crocker ,
D. Crutchfield F. Williams R. Heineman H. Denton D. Muller Project Manager D. Allison i Attorney, ELD E. Hylton IE ( 3)
ACRS (16) -
L. Dreher ,
NRC Participants E. Case {
R. DeYoung {
l D. Vassallo j J. Stolz l D. Allison j J. Knight .l R. Bosnak l P. Chen l T. Suljlivan !
I. Sibweil l D. Jeng ,
P. Kuo J. O'Brien A. Fratoni W. Gammill M. Grossman J. Tourtellottee L. Davis ",
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