ML20236C367
| ML20236C367 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon, 05000000 |
| Issue date: | 09/18/1975 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20236A877 | List:
|
| References | |
| FOIA-87-214 NUDOCS 8707300031 | |
| Download: ML20236C367 (40) | |
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September 18, 1975 I
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'[,, ojb5 SUPPLEMENT NO. 3 TO THE SAFETY EVALUATION REPORT BY THE OFFICE OF NUCLEAR REACTOR REGULATION l 1
I U.S. NUCLEAR REGULATORY COMMISSION l
i IN THE MATTER OF l
PACIFIC GAS AND ELECTRIC COMPANY DIABLO CANYON NUCLEAR POWER STATION, UNITS 1 AND 2 DOCKET NOS. 50-275 AND 50-323_
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8707300031 870721 PDR FOIA CONNOR87-214 PDR v .-
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_ TABLE OF CONTENTS PAGE
1.0 INTRODUCTION
................................................. 1-1 3.0 DESIGN CRITERIA - STRUCTURES, COMPONENTS, EQUIPMENT, AND SY ST EM S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 - 1 3.5 Hiss 11e Protection Criteria............................ 3-1 4.0 REACT 0R...................................................... 4-1 4.2 Mechanical Design...................................... 4-1 4.2.1 Fuel............................. ............. 4-1 4.4 The rmal and Hyd raulic Design. . . . . . . . . . . . . . . . . . . . . . . . . . . 4-2 5.0 REACTOR COO LANT SYST EM. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5- 1 5.2 Integrity of Reactor Coolant Pressure Boundary. . . . . . . . . 5-1 5.2.1 Design of Reactor Coolant Pressure Boundary Components.......................... 5-1 6.0 ENGINEERED SAFETY FEATURES................................... 6-1 I
6.2 containment Systems.................................... 6-1 )
6.2.1 Containment Functional Design.................. 6-1 6.2.2 Containment Heat Removal Systems............... 6-2 l
6.3 Emergency Core Cooling System (ECCS)................... 6-3 j
6.3.1 Design Bases................................... 6-3 1
j 6.3.3 Perf o rmanc e Evalua tion . . . . . . . . . . . . . . . . . . . . . . . . . 6- 5 l 6.3.5 Conclusion...................................... 6-5 10.0 STEAM AND POWER CONVERSION SYSTEM. . . . . . . . . . . . . . . . . . . . . . . . . . . 10-1 i J
10.4 Other Features........................................ 10-1 I I
15.0 ACCIDENT ANALYSES........................................... 15-1 l i
15.1 G e n e r a 1. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15-1 i
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1 TABLE OF CCHTENTti (Cont'd) 'l 1
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PAGE 17.0 QUALITY ASSURANCE.........................................., 17-1 17.3; Quality Assurance Program............................. 17-1 j 18.0 REVIEW BY THE ADVISORY COMMITTEE ON REACTOR SAFf6UARDS (ACRS)........................................ 18-1 l l
22.0 CON CLU S I ONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22-1 I l
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APPENDICES PAGE l APPENDIX A - CONTINUATION OF THE CHRONOLOGY OF THE RADIOLOGICAL REVIEW................................... A-1 APPENDIX B - REPORT OF THE ADVISORY COMMITTEE ON l REACTOR SAFEGUARDS, Dated June 12, 1975............... B-1 ]
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l 1-1 1.0 1 INTRODUCTION l
The Commission's Safety Evaluation Report in the matter of the '
1 Diablo Canyon Nuclear Power Station, Units 1 and 2, was issued on j October 16, 1974. In the Safety Evaluation Report it was stated that I supplemental reports would be issued to update the Safety Evaluation ]
Report in those areas where the staff's evaluations had not been completed. Supplements 1 and 2 to the Safety Evaluation Report, issued on January 31, 1975 and May 9, 1975, respectively, documented the resolution of several outstanding items, and summarized the status of the remaining outstanding items.
The purpose of this Supplement Number 3 is to further update the i Safety Evaluation Report by providing the staff's evaluation of certain j matters which were not resolved when Supplement Number 2 was issued.
Each of the following sections of this supplement is numbered the same ,
I as the corresponding sections of the Safety Evaluation Report that are being updated.
Appendix A of this supplement is a continuation of the chronology of the Nuclear Regulatory Commission staff's principal actions with respect to radiological matters related to the processing of the appli-cation. Appendix B is a report by the Advisory Committee on Reactor Safeguards. l l
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'3-1 3.0 DESIGN CRITERIA - STRUCTURES, COMPONENTS, EQUIPMENT, AND SYSTEMS 3.5 Missile Protection Criteria In a letter to the applicant dated July 11, 1975 we requested additional information describing the feasibility and cost of making improvements to the tornado missile protection for three areas of the
-i l- plant which are still being reviewed. These areas are as follows:
(1) The 4.16167 switchgear roons in the turbine building at eleva-l tion 119 do not appear to have substantial protection' f rom tornado missiles which might penetrate the turbine building j walls in the vicinity of these rooms. From certain directions, the existing protection consists of relatively light weight. .
exterior wall panels. ;
(2) The diesel generators and their radiators appear to be vulner-able to small missiles which might go through the spaces between the 1 - beams in the missile protection shield for the air in-takes.
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(3) It is not clear what missile resisting capability is available for the pipes which connect to the component cooling surge tanks j l
which are placed on top of the auxiliary building. Also, it is 4
not clear whether the component cooling water system could be operated if damage to the surge tanks or piping were sustained.
In Amendment 33 to the FSAR, the applicant provided (1) a revised design for the diesel generator air intake missile shield, and (2)' a description of the consequences of tornado missile damage to the com-
, ponent cooling surge tanks and connected piping. The applicant is l
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3-2 expected to submit additional information regarding the protection for the switchgear rooms. We will report the results of our evaluation of all of this information in a future supplement to the Safety Evaluation Report.
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4-1 4.0 REACTOR 4.2 Mechanical Design
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4.2.1 Fuel In the Safety Evaluation Report and in Supplement Number 2 to the Safety Evaluation Report, we stated that Westinghouse would docu-ment the justification for applying the results of certain tests made on 17 x 17 fuel assemblies with seven spacer grids to 17 x 17 fuel assemblies with eight spacer grids.
The tests in question, which were perfor=ed with 17 x 17 fuel assemblies with seven spacer grids, were reported in WCAP-8278,
" Hydraulic Flow Test of the 17 x 17 Fuel Assembly", and WCAP-8236,
" Safety Analysis of the 17 x 17 Fuel Assembly for Combined Seismic and Loss-of-Coolant-Accident". We have previously concluded, as described in t%a Safety Evaluation Report, that these reports are acceptable for seven grid fuel assemblies.
j In a letter dated May 20, 1975, Westinghouse documented the justi-fication for applying the results of the seven grid tests to 17 x 17 fuel assemblies with eight spacer grids. Our primary concern with these tests was that the fuel assembly design be able to withstand flow-induced vibration. The information submitted by Westinghouse provides justification for demonstrating that the design of the 17 x 17 fuel '
assembly is adequate to withstand the effects of flow-induced vibration under both normal operating and transient conditions. We have evalu-ated this justification and found it to be acceptable. Based on our evaluation of the additional information which was provided, we con-sider this matter to be resolved.
R 4-2 Based on our evaluation, as described in the Safety Evaluation !
i Report, Supplement Number 2, and this report, we have concluded that the mechanical design of the Diablo Canyon fuel is acceptable.
4.4 Thermal and Hydraulic Design 4
In the Safety Evaluation Report (see page 4-15) we stated that I l
the f acilities were designed to operate at a higher heat output l
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and temperature than comparable units. This conclusion was based l l
on the use of the THINC code which permitted a more detailed l
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analysis of tne thermal and hydraulic characteristics of the core.
However, we required verification of the calculational accuracy l
of the code, based on the Zion facility. In Supplement Number 2 I
to the Safety Evaluation Report, we stated that the results of the j i
verification test program for the THINC code had been documented I l
in topical reports WCAP-8453 (Proprietary) and WCAP-C454 (Non- j Proprietary), " Analysis of Data from the Zion (Unit 1) THINC Verification Test," December 1974. Since our evaluation of these reports was not completed at that time, we further stated that, if sufficient verification could not be obtained from the combined test and analytical programs, restrictions would be included in the Diablo Canyon technical specifications to maintain required margins to fuel rod damage.
We have now completed our evaluation of the above mentMned
, topical reports; this evaluation is documented in a letter to l
Westinghouse dated July 9, 1975. Based on our evaluation, we have concluded that no further in-core testing is required for the THINC code, since the THINC calculations gave conservative results with
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respect to the Zion test results. Because the Zion and Diablo l Canyon cores have similar thermal and hydraulic characteristics,
- we have concluded that no special operating restrictions will be required for the Diablo Canyon Units. We consider this matter-to be resolved. l 1
I In Supplement Number 2 to the Safety Evaluation Report, we stated that we considered the effects of fuel rod bowing on departure from nucleate boiling heat flux af ter the first fuel cycle to be unresolved, and that we would report the resolution of this matter l l
in a future supplement to the Safety Evaluation Report. Westinghouse l l
has submitted additional information concerning fuel rod bowing after the first fuel cycle.
l We have not completed our evaluation of this information, and we expect additional data to be submitted. We will report the resolution of this item in a future supplement to the Safety Evaluation Report.
In Supplement Number 2 to the Safety Evaluation Report, we-stated that, if the results of the non-uniform departure from nucleate boiling tests were not available when the technical specifications for Diablo Canyon were finalized, we would require that the minimum allowable departure from nucleate boiling ratio be increased 5 percent above that requi"ed to satisfy the 95/95 criterion (see page 4-12 of the Safety Evaluation Report) . These results have been documented in WCAP-8536 (Proprietary), " Critical Heat Flux Testing of 17 x 17 Fuel Assembly Geometries with 22 Inch Grid Spacing."
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4-4 We have not completed our evaluation of this information, and will report the resolution of this item in a future supplement to the Safety Evaluation Report.
Based on our evaluation, as described in the Safety Evalution Report, Supplement Number 2, and this report, we have concluded, subject to favorabic resolution of the outstanding items described l
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above, that the thermal and hydraulic design of the Diablo Canyon i'
reactors is acceptable, and that these reactors can operate at the proposed core power levels.
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i 5.0 REACTOR COOLANT SYSTEM !
5.2 Integrity of Reactor Coolant Pressure Boundary 1 J
i 5.2.1 Design of Reactor Coolant Pressure Boundary Components 1 Regarding the blowdown forces acting on the reactor pressure vessel in the unlikely event of a loss-of-coolant accident, we were informed on May 7, 1975, by a Licensee of a pressurized water reactor, Virginia Electric and Power Company, that an asymmetric loading re-sulting from a postulated pipe rupture at a particular location in the reactor coolant loep had not been taken into account in the original i
design analysis of the reactor vessel support system. This loading 1
{l results from the forces induced on the internals within the reactor l
vessel during such a loss-of-coolant accident caused by transient j i
differential pressure conditions within the vessel. In addition, the asymmetric loading from the transient differential pressures that would exist around the exterior of the reactor vessel from the same postu-lated pipe rupture was not included in the original design analysis.
However, the symmetric loadings from such a pipe rupture were included in the original analysis of the reactor vessel support system.
It is our opinion that these factors related tc the adequacy of vessel support systems are generic in nature, and apply to the Diablo Canyon Units.
It is also our opinion that the original design of the reactor vessel support system for these Units was developed using con-servative methods of analysis for the symmetric leads that were considered.
Accordingly, it is likely that,when a systematic estimate of the conser-vatism used in the original analysis is completed, and the asymmetric
5-2 loads are properly taken into account, it will be confirmed that, although t'ie margins of safety will be less than previously thought, the vessel support system can safely withstand the postulated effects i
of this particular loss-of-coolant accident.
We are continuing our review of this matter, and will report ,
the results of our evaluation in a future supplement to the Safety Evaluation Report.
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i 6.0 ENGINEERED SAFETY FEATURES 6.2 Containment Systems 6.2.1 Containment Functional Design With ragard to subcompartment pressure calculations, in Supplement Number 2 to the Safety Evaluation Report we stated that the applicant must provide additional information on the geometry of the system to justify the assumed limitation on the size of the opening that can result I
from a break at the reactor vessel nozzle weld. We further stated that, I with regard to the pressure response of the reactor coolant system pipe penetration ;. rough the reactor shield structure, the applicant must analyze the response of the. piping penetration to a pipe break within the penetration using an acceptable method of analysis, or else justify that a break need not be postulated inside the penetration.
In Amradment 30 to the FSAR, the applicant provided additional in rarmation regarding (1) the limited displacement rupture of the reactor coolant pipe at the nozzle weld, and (2) justification that a pipe break need not be assumed inside the piping penetration. The applicant demonstrated that the geometrical configurati:. limits the size of the opening that can result from a break at the nozzle weld. In
, addition, the applicant has used the appropriate guidelines in hguinory Guide 1.46 to justify that a break need not be postulated inside the piping penetration.
We have reviewed the information submitted and, based on our evaluation, we have concluded that the applicant has adopted acceptable criteria for postulated break locations and type, and has provided an acceptable basis for the assumed limited displacement of the ruptured pipe. We consider these matters regarding subcompartment pressure calculations to be resolved. ;
6-2 6.2.2 Containment Heat Removal Gystems Because of the possibility that insulation could be torn from j piping or equipment 1.nside containment during a loss-of-coolant acci-dent and block the flow of water through the sump screens, we asked the applicant to perform an evaluntion to determine what amounts of i i
insulation might become detached due to jet effects created by a pipe break inside containment. In Amendment 29 to the FSAR the applicant provided information regarding the design of piping and equipment insulation used within the containment, and the expected behavior of i
this 1nsulation during a postulated loss-of-coolant accident. I 1
Both reflective metal panel insulation and stainless steel jacketed calcium silicate insulation are used inside containment. Jet effects .
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ment. However, the arrangement of floors, walls, and compartments would prevent some portion of this dislodged insulation from reaching the containment sump. ,
i To prevent blockage of the recirculation flow at the sump, the de- l sign of the Diablo Canyon containment sump includes a debris curb, trash rack, double screens and large screen areas which cause very low flow velocities as the water approaches and passes through the screens. This permits most of the debris to settle out before reaching the containment sump. The applicant has also performed a jet force ' analysis to estimate the maximum amount of piping and equipment insulation which could be removed in the vicinity of a pipe rupture. The analysis shows that the total surface area of insulation removed is significantly less than the sump coarse screen area.
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6-3 Based on our evaluation of the information submitted by the applicant, we have concluded that the use of reflective metal insu-lation and steel jacketing around the conventional insulation, coupled with the design features of the containment samp, provide adequate assurance that the sump will not be blocked by displaced insulation during a loss-of-coolant accident. We consider this matter to be resolved. !
6.3 Emergency Core Cooling System (ECCS) 6.3.1 Design Bases In Supplement Number 2 to the Safety Evaluation Report, we stated that the applicant had provided a commitment to lock out power to those motor-operated valves identified by the staff as af fecting the function of the ECCS if spurious operation were to occur. The valves for which locking out of power would be required were those identified in item 7.12 of a request for additional information (see items 42 and 45 in Appendix A to the Safety Evaluation Repert). At: the time we con-sidered this matter to be resolved.
Since that time, two additional valves, the containment sump isolation valves (Valve Numbers 8982, A and B), have been identified as a potential problem where spurious operatio could result in the loss of system function. These valves are located in two lines from the sump, and are required to operate for realignment to the recirculation mode.
Each line communicates with one of the two resid 2al heat removal pump suction lines. These in turn connect to the refueling water storage tank through a common line which contains a check
6-4 valve. Each containment sump isolation valve is interlocked with a motor operated valve in its individual pump suction line to prevent l communication between the containment sump and the common line to the refueling water storage tank. However, in the event of an electrical failure causing spurious mechanical motion to open either sump isolation valve during the early safety injection phase following a locs-of-coolant accident, containment pressure (up to 47 psig) could force closure of the check valve in the common line to the refueling I
water storage tank. This could result in cavitation and damage to j both residual heat removal pumps, one of which is required for l
long-term cooling following a loss-of-coolant accident.
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l We have discussed this matter with the applicant, and requested ad- )
i ditional information regarding our concerns with this matter. In Amend- !
ment 33 to the FSAR, the applicant proposed to lock out power to the )
1 containment sump isolation valves mentioned above (Valve Numbers 8982, j A and B). We arc evaluating this information submitted by the applicant, j i
and will report the final resolution of this item in a future supplement l l
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l 6.3.3 Performance Evaluation l In Supplement Number 2 to the Safety Evaluation Report,.we stated that the applicant would submit a revised analysis of the emergency core cooling system performance.
In Amendment 33 to the FSAR the applicant submitted a revised analysis. However, the analysis submitted was valid only for Diablo ,
Canyon Unit 1, and it did not contain all of the information which we i had previously requested (see item 136 in Appendix A of this report).
1 We will review the analysis in Amendment 33 as well as the additional information to be submitted by the applicant, and will report the results of our review in a future supplement to the Safety Evaluation Report prior to a decision concerning the issuance of operating licenses for Diablo Canyon Units 1 and 2.
6.3.5 Conclusion Our evaluation of the acceptability of the emergency core cooling system has not been completed. The items which are currently unresolved
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are described in Sections 6.3.1 and 6.3.3 of this report. We will I
report our conclusions regarding the acceptability of the emergency core cooling system in a future supplement to :he Safety Evaluation Report.
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10.0 STEAM AND POWER CONVERSION SYSTEM 10.4 Other Features 1
- l In June of this year (see item 135 in Appendix A of this report),
1 we asked the applicant to provide information regarding the potential effects of secondary system fluid flow instability. Events i
sur.h as damage to the feedwater system piping that occurred at !
Indian Point 2 on November 13, 1973, could originate as a consequence I
a of the uncovering of the feedwater sparger in the steam generator or j uncovering of the stean. generator feedwater or auxiliary feedwater 1
inlet nozzles. -
We have asked the applicant to provide information which demonstrates that unacceptable damage will not result from feedwater hammer on the Diablo Canyon Units. This information can be provided in the form of. ;
4 adequate testing which would be performed in accordance with plant !
procedures.
l The applicant provided the information on feedwater hammer in August of this year (see item 142 in Appendix A of this report). We ;
i are currently evaluating this information, and will report the results of our evaluation in a future supplement to the Safety Evaluation Report. l This matter must be recolved before the plant can be operated at power.
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I 15-1 15.0 ' ACCIDENT ANALYSES 15.1 General In Supplement Number 2 to the Safety Evaluation Report, we stated that the applicant must justify the assumption that air entering the I auxiliary building would be heated 15 degrees Fahrenheit before it reaches the auxiliary building charcoal filters. This was the only outstanding issue remaining with regard to dose reduction in the event of residual heat removal leakage following a postulated loss-of-coolant accident.
In Amendment 28 to the FSAR, the applicant provided additional j
information with regard to this matter.- The applicant has described '
the heat loads used to calculate the 15 degree temperature rise. We have reviewed this information and found it to be acceptable.
Based on our evaluation of the applicant's calculations, we have concluded that adequate humidity control has been provided for the charcoal filters in the event of residual heat removal system leakage i
following a postulated loss-of-coolant accident. We consider this 1
matter to be resolved.
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, 1 17-1 17.0 QUALITY ASSURANCE 17.3 Quality Assurance Program In Section 22 of Supplement Number 2 to the Safety Evaluation Report, we stated that we had not completed our evaluation of the applicant's commitment to the guidance in certain WASH documents I
which pertain to the operational quality assurance program.
1 In Amendment 30 to the FSAR the applicant revised this commit-ment. The applicant stated that the quality assurance program for plant operation of Diablo Canyon, Units 1 and 2, complies with the guidance contained in WASH-1283 (May 24,1974), " Guidance on Quality Assurance Requirements During Design and Procurement Phase of Nuclear Power Plants - Revision 1"; WASH-1284 (October 26, 1973), " Guidance j on Quality Assurance Requirements During the Operations Phase of Nuclear Power Plants"; and WASH-1309 (May 10, 1974), " Guidance on Quality Assurance Requirements During the Construction Phase of Nuclear Power Plants." This complies with our peaition on the imple-mentation of guidance in quality assurance programs and is, therefore, 4 acceptable.
We consider this matter to be resolved. j Basedonourevaluati$nasdescribedintheSafetyEvaluation i' Report and supplemented in this report, we nov conclude that the Diablo Canyon quality assurance program has the necessary controls to comply l
with the requirements of Appendix B to 10 CFR Part 50 and is, therefore, acceptable for controlling the operational phase of Diablo Canyon, Units 1 and 2.
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18.0 REVIEW BY THE ADVISORY COMMITTEE ON REACTOR EAFEGUARDS (ACRS)
The Advisory Committee on Reactor Safeguards completed a partial review of the application for operating licenses for Diablo Canyon, Units 1 and 2, at its 182nd meeting, June 5-7, 1975. The Diablo Can-yon Units were previously consi'dered at Subec=mittee meetings held in Washington, D.C. on September 12, 1974, in San Luis Obispo, California on February 18-19, 1975; and in Los Angeles, California on May 23, 1975.
A copy of the Committee's report on this partial review, dated June 12, i
liFS, is attached as Appendix B to this report. )
i We will report the actions we have taken or plan to take in response )
l to the Committee's comments in a future supplement to the Safety l 1
Evaluation Report.
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22.0 CONCLUSION
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In Section 22 of Supplement Number 2 to the Safety Evaluation Report, we stated that several items were still outstanding, and that f avorable resolution of these items would be required before operating licenses 1
for Diablo Canyon Units 1 and 2 could be issued. A number of these items have been resolved in this supplement. A revised status report is given below. It includes those items which were net resolved in Supplement l Number 2 (itemsl through 14 below), as well as new unresolved items which have atisen since Supplement Number 2 was issued (items 15 ,
I through 19 below). )'
l (1) The applicant has provided additional information on the effects l
of tsunamis caused by near-shore generators; however, our evaluation of this information has not'been completed (Sections 2.4.2, 2.4.3 and 2.4.5 of Supplement Number 1),
(2) Our comparative evaluation of the Hosgr1 and Santa Lucia Bank faults, and our evaluation of the earthqvake potential of the Hosgri Fault have not been completed. Tne applicant must pro-vide additional information before our evaluation can be completed (sections 2.5.1 and 2.5.2 of Supplement Number 1). j i
(3) The applicant has provided information on the potential consequences '
of pipe breaks outside containment. During our evaluation of this inf ormation we have determined that additional information must be provided (see Item 143 in Appendix A of this report). (Section 3.6 of the Safety Evaluation Report). !
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- 5) The applicant has submitted some of the required information con-I cerning the seismic qualification of Category I instrumentation
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before our evaluation can be completed (Sections 3.10 and 7.8 of the Safety Evaluation Report). j l
(5) Our evaluation of the justification for applying the results of l l
tests of 7-grid assemblies to 8-grid assemblies has been completed, J and this matter is resolved (Section 4.2.1 of this report).
(6) The matters regarding uncertainties in the thermal and hydraulic design have been partially resolved (Section 4.4 of this report).
(7) We 1. ave completed our evaluation of the matters regarding subcom-partment pressure calculations using the Transient Mass Distribution Program. This matter is resolved (Section 6.2.1 of this report).
(8) We have not completed our evaluation with regard to single failures in the containment sump isolation valves (Sections 6.3.1 and 5.3.5 of this report).
(9) The applicant has submitted part of the required emergency core 1
cooling system analysis in accordance with the Final Acceptance l Criteria. Additional information must be submitted before our evaluation can be completed (Sections 6.3.3 and 6.3.5 of this i report).
(10) The applicant has provided information regarding physical separation in the process analog system. Additional information must be provided before our evaluation can be completed (Section 7.2.3 of the Safety !
Evaluation Report).
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I (11) Our evaluation of the Westinghouse generic ATWS model has not been completed (Section 7.2.5 of the Safety Evaluation Report).
(12) The applicant has provided some information concerning the environ-mental qualification of Category I instrumentation and electrical i equipment. Additional information must be submitted before our evaluation can be completed (Section 7.8 of the Safety Evaluation Report).
k (13) Our evaluation of the design modifications to the auxiliary build- i ing to bring about a reduction of the doses in the event of residual heat removal system leakage following a postulated loss-of-coolant
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accident has been completed.
This catter is resolved (Section 15.1 of this report).
(14j The matter regarding the operational quality assurance program is resolved (Section 17.3 of this report).
(15) Our evaluation of the potential effects of insulation used inside containment on sump operation following a loss-of-coolant accident has been cumpleted.
This matter is resolved (Section 6.2.2 of this report).
(16)
The applicant has submitted information concerning the potential for damage due to feedwater system water hammer. Our evaluation of this information has not been completed (Section 10.4 of this report). I 1
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22-4 (17) The applicant has submitted additional information regarding tor-nado missile protection. Our evaluation of this information has not been completed, and further information must be provided )i (Section 3.5 of this report).
(18) Regarding fire protection in the vicinity of electrical cabling and equipment, the applicant has not provided the additional l
information that was requested (Item 143 in Appendix A of this 1
report).
(19) Our evaluation of the factors related to the adequacy of the i reactor vessel support systems has not been completed (Section '
5.2.1 of this report). j l
l Subject to f avorable resolution of the outstanding matters de-scribed above, the conclusions as stated in Section 22 of the' Safety Evaluation Report remain unchanged.
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1 A-1 l APPENDIX A l I
CONTINUATION OF THE CHRONOLOGY OF THE RADIOLOGICAL PEVIEW 127. May 9, 1975 Supplement No. 2 to the Safety Evaluation Report issued.
128. May 16, 1975 Submittal of Amendment No. 28 consisting of additional information required for the resolution of outstanding items in the Safety Evaluation Report.
129. May 19, 1975 Letter to applicant transmitting the sched-ule for implementation of.the Westinghouse Standard Technical Specifications.
130. May 23, 1975 ACRS Subcommittee Meeting in Los Angeles, California.
131. June 2, 1975 Submittal of Amendment No. 29 conristing of additional information required for the resolution of outstanding items in the Safety Evaluation Report.
132. June 5, 1975 ACRS Full Committee Meeting in Washington, D.C.
133. June 12, 1975 ACRS letter which constitutes a partial review of the Diablo Canyon operating license application.
l 134. June 13, 19 75 Letter to applicant requesting additional ;
information on boron precipitation effects )
on long term cooling for the emergency core l cooling system. l 135. June 13, 1975 Letter to anplicant requesting additional information on the effects of secondary I system fluid flow instabilities. ,
l 136. July 9, 1975 Letter to applicant requesting additional l information on the emergency core cooling I system.
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A-2 137. July 11, 1975 Letter to applicant requesting additional information on tornado missile protection for the Diablo Canyon Units.
138. July 16, 1975 Letter to applicant requesting additional information regarding the extension of the construction permits for Units 1 and 2.
139. July 29, 1975 Submittal of report on the consequences of seismic-induced actuation of protection I system relays on the Diablo Canyon Units. l 140. July 29, 1975 Submittal of Amend:nent No. 30 consisting of additional information required for the resolution of outstanding items in the Safety Evaluation Report.
141. July 31, 1975 Meeting with applicant to discuss geology and seismology.
l 142. August 5, 1975 Letter from applicant transmitting the ,
I information on the effects of secondary system fluid flow instabilities that was requested in our letter of June 13, 1975.
143. Augus t 7, 1975 Letter to applicant requesting additional information on electrical instru:nentation
! and control syste=s, pipe break outside
! containment, and fire protection.
144. August 15, 1975 Submittal of Amendment No. 31 consisting of a partial response to the staf f's re-quest for additional information on geology and seismology dated February 12, l
1975.
145. August 19, 1975 Submittal of Amendment No. 32 consisting
, of a partial response to the staff's re-quest f( r additional information on !
l I geology and seismology dated February 12, l l 1975. j I
146. August 20, 1975 Submittal of a marked-up copy of the Westinghouse Standard Technical Specifi-cations for the Diablo Canyon Units.
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l 147. August 28, 1975 Meeting with applicant to discuss geology l and seismology. j 148. August 29, 1975 Submittal of Amendment No. 33 consisting f primarily of a revised emergency core j cooling system analysis.
149. September 4, 1975 Letter to applicant transmitting our review of the Westinghouse Protection System Noise i Test Report, and requesting additional
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information on this item. i 150 September 16, 1975 Submittal of Amendment No. 34 consisting l of a partial response to the staff's {
request for additional information on j geology and seistology dated February 12, ?
1975. I l
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, APPENDIX B ADVISORY COMMITTEE ON REACTOR SAFEGUARDS NUCLEAR REGULATORY COMMISSION W ASHINGTON. D. C. 20665
,jg i2 $75 Honorable William A. Anders Chairman U.S. Nuclear itquiatory Commission Washington, D.C. 20555
Subject:
REPORT ON PARTIAL REVIEW OF DIAPID CANYON t0 CLEAR POWER STATION UNITS 1 AND 2
Dear Mr. Anders:
At its 182nd meeting, June 5-7, 1975, the Advisory Committee on Reuter Sieguards conpleted a partial review of the application of the Pacific Gm aM Electric Conpany for authorization to operate the Diablo Canyca Aclear Power Station Units 1 and 2. %eproject was oreviously considered at Subcommittee mP ' ings in Wisiington D.C.
on September 12, 1974; in San Luis Obispo, La.ifornia on February 18-19, 1975; and in Ios Angeles, California on May 23, 1915. During its review, the Comittee had the benefit of discussions with representatives and consultants of the Pacific Gas and Electric Company, the Westinghouse Electric Corporation, and the Nuclear Regulatory Ccmission (NRC) Staff. We Comittee also had the bene-fit of the documents listed. W e Committee reported on the aoolica- !
tion for a construction permit for the Diablo Canyon Relear Power l Station m it 1 in its letter of December 20, 1967, and for thit 2 in its letter of m tober 16, 1969.
%e site is located on 750 acres adjacent to the Pacific Ocean in San Inis Obispo, County, and is approximately 12 miles west- !
southwest of the city of An Inis Obiso.
,% e two units at the Diablo Canyon Station are essentially identical. Each includes a four-loop Westinghouse nuclear steam supply system similar in most respects to that for the Trojan Ntelear Plant, on which the ACRS reported on ibrember 20, 1974.
%e design core power level for' Unit 1 is 3338 MW(t)'and thit 2 is 3411 *(t) . %e slight difference in output for the two units is due to the upgraded turbine generator design for thit 2.
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lionorable William A. Anders JUN 12 N W e Committee has not completed its review of the seinic desian bases, the adequacy of the seismic design, or of the requirements with regard to protection.against tsunamis. %ese, and some additional matters dise"---d below, will be reviewed by the Comittee following completion c. teview of seismic-related topics by the WC Staff.
W e Diablo Canyon Units 1 and 2 are scheduled to be among the l first to go into operation using a full-core of 17x17 fuel. t ile j many of the various required verification procrans have been com-.
pleted and reviewed by the NRC Staff,'other tests and analyses are still to be documented and. reviewed. % ese include: INR tests for non-uniform heat flux, fuel assembly flow tests, and the effect i of fuel rod bowing on DNB after the first fuel cycle. We results I of such tests and analyses should be evaluated fully by the NRC Staff, and resolved to it's satisfaction, exior to the full-core j i use of 17x17 fuel to produce power. Prototyoe 17x17 fuel rod ;
l assemblies are to be loaded into operating pressurized water ;
I reactors in the near future; the results of these irradiations - R l should be followed closely. W e Committee wishes to be kept in- I formed concerning the rerlts of the various ongoing 17x17 test : ;
and analytical programs, and any design changes which may be l proposed in the future.
- 1 l Following each cycle of operation,17x17 fuel assemblies will l be examined for fuel rod integrity, fuel rod and assembly di ension i and alignnent, and surface deposits. In view of the fact that the -
17x17 fuel array is a new design and that no prototype irradiations are planned for 17x17 fuel containing eight soacer-grids, the results of surveillance programs for this tyee fuel saould be followei closely.
%e Conmittee wishes to be kept informed. ]
We recently pronosed rethod of constant axial offset control will he used for core power distribution monitorino and control. %e l NPC Staff shouli review the effectiveness of this nethod in i mrotectinq aqainst adverse consequences of mstulated reactor ,
I ttansienta andi accidents. We Committee wishes to be kept informed. ;
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1 Nveral chanqes have been made in the Westinghouse ECCS evaluation Wel to bring it into conformance' with the Cor.nission criteria as given in 10 CFR 50, ADpendix K. %e. performance of.the emergency core cooling systems will be reevaluated with the approved evaluation nodel, and appropriate operuina linits and procedures :
for ensurinq ronitoring of the oower distribution are to be ,
incorporated in the %chnical Specifications. W e Comm.ttee i wishes to be kept informed. i 0
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Honorable Willian A. Anders JUN 1 2 & a
%e evaluation of Anticipated Transients Without Scram has been made generically for Westinghouse plants, and the applicant has made com-parisons indicatina that the results obtained are apolicable to Diablo Canyon Units.1 and 2.
NRC Staff review should be completed and this matter resolved in a manner satisfactory to the NRC Staff and the ACRS.
Diablo Canyon Units 1 and 2 may be amona the f.'est reactors of this type to operate at a power as high as 3411 MW(t). Because there is limited operating experience with very large, high-power density reactors, the ACRS has previously recommended a more cautious-than-normal approach to full power, with longer periods of operation at power levels in the range of 70 to 90% of full power, and with additional nonitoring of core and systems performance throughout the life of the first core.
%e applicant discussed with the Committee an augmented startup program, which is proposed for implementation on several of the first plants to operate with a full-core employment of the 17x17 fuel assembly. We Committee believes that the augmented program is desirable and recomends that the NRC Staff evaluate the results of this program, as well as overall operatirn experience with large high power-density reactors, prior to sustained operation at full power.
Certain ascects of the protection against tornadic missiJes are still under evaluation.
satisfactory to the Bis imCmatter Staff. should be resolved in a ranner
%e applicant has not provided adequate information to confirm the environmental eouipment. qualification of Class I instrumentation and electrical
%is matter should be resolved in a m'nner satisfactory to the NRC Staff and the ACRS.
Generic problens relatirg to large water reactors have been identified by the iPC Staff and the ACRS and discussed in the Committee's reuort dated March 12, 1975. %ese problem should be dealt vid appropriately are developed. by the imC Staff and the aoplicant as suitable approaches i Several unresolved items were identified by the NRC Staff in'their Supplement No. 2 to the Safety Evaluation Report, and at the !by 23, 1975 Subcommittee Meeting. %e ACRS expects these to be resolved in timely fashion arri plans to review several of them, including matters relating to water-hammer effects and subcompartnent pressures in a postulated WCA, in connection with its further review of seismic-related aspects. Were also remain some systems behavior and inter-actions questions an6 some questions concerning forces on the pressure vessel support structure during blowdown for certain postulated accidents which the Comittee expects to review further.
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Ibnorable William A. Arders jgg , z ha l I
l Excepting the seismic and other matters identified above as !
requiring further Committee. review, the ACRS believes that, if due i regard is given to the items mentioned above, ard subject to satis- ,
factory completion of construction and pre-operational testirg, !
there is reasonable assurance that the Diablo Canyon Nuclear Power Station thits 1 and 2 can be operated at power levels up to 3338 and 3411 Ml4(t), respectively, without undue risk to the health ;
and safety of the public. The Comittee will report in the l future on those matters not reviewed herein.
Sincerely, W. Kerr Chairman l Feferences
- 1. Final Safety Analysis Beport (PSAR) for the Diablo Canyon Nuclear Ibwer Station, thits 1 & 2, and Amendments 1-28 to the PSAR. l
- 2. Safety Evaluation Report dated October 16, 1974, and Supplements 1 & 2 dated January 31, 1975 and May 9,1975, respectively.
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- 3. tetter dated April 7,1975, ucific Gas and Electric (bnpany (PG&E) to NRC, concernirg evaluation of corrosion resistance j of alternate steam generator tube materials. J
- 4. Dnergency Plan for the Diablo Canyon facility dated March 21, l 1974, and Appendix I dated June 1974.
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Regulatory
.LQ PACIFIC GAS AND E LE C T RIC C O M PANY 1POV#3 l 77 BEALE STREET
- SAN FRANCISCO, CAUFORNI A 94106 * (415)781 4211 .
TWX 910 372-6587 August 15, 1975 g .gsL- TE~~,.
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Mr. L. C. Rouse Chief, Fuel Cycle Licensing Branch 1 id 9 Division of Materials and b12 6 sed!O.lIG @ DL-Fuel Cycle Facility Licensing U. S. Nuclear Regulatory Commission D '< G l N QNiJ>' ,;'y'., /
Washington, D. C. 20555 Nh , , l7 Re: Docket 50-275-OL W.P Diablo Canyon Site Unit 1
Dear Mr. Rouse:
Your letter dated June 17, 1975, asked that we propose additional steps to assure that new fuel in , torage will not achie"e criticality, assum-ing that structural failure of the storage facilities might occur leading to flooding and loss of spacing bettieen fuel assemblies.
In order to achieve criticality, both a loss of spacing between fuel essemblies'and flooding of the' fuel assemblies must occur. We have re-viewed the Diablo Canyon design and have identified potential sources of water, failure modes, and flow paths which could contribute to accidental flooding of the fuel storage facilities. It is our conclusion that such flooding _ is ex-tremely unlikely, even if a seismic event is postulated which results in a .
loss of spacing between fuel assemblies.
However, in order to provide further assurance that criticality will not occur, assuming the occurrence of a seismic event resulting in a loss 'of spacing between fuel assemblies, PGLE proposes to take the following additional steps for storage of the initial core loading of new fuel assemblies for Diablo Canyon Unit 1:
- 1. Store all new fuel assemblies in the spent fuel storage pool.
2 Store all burnable poisons and control rod clusters in their respective fuel assemblies.
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- 3. Add an isolation valve in the fire system in order to isolate q,
the fire main feeding fire stations at elevation 140 f t. in
,the fuel handling area. \)
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DIABLO CANYON UNIT 1 SAFEIT EVAUJATION Storage of New Fuel Assemblies for Initial Core Loading August 15, 1975 E
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I SAFETY E7AUJATION The occurrence of criticality is the only event involving the I storage of new fuel at Diablo Canyon which could adversely affect the healta ,
and safe,ty of the public. Tne possibility of this occurrence has been eval- q uated, assuming that the additional steps proposed by PGiE for storage of new fuel assemblies are taken. The results of this evaluation are as follows:
Loss of spacing between fuel assemblies could not result in criticality unless it were accompanied by the addition of unborated water to the spent fuel pool. Potential sources of water, failure modes, and flow ;
paths have been identified which- could result in water entering the spent fuel pool. It is concluded that there is reasonable assurance q that water will not enter the pool in a manner which could reduce boron 1 concentration to 'a level where criticality could occur.
Criticality Analysis - Effect of Boron Concentration An analysis has been made to determine the relationship between boron concentration and Keff for new fuel assemblies stored iri the spent fuel paol, with loss of spacing between fuel assemblies assumed and with the poolin-tentionally flooded with borated water. This analys$ r made use of the " Leopard" computer code and assumed the following:
- 1. 17 - 17 fuel assemblies, Zr-clad,12 f t. active fuel length.
- 2. A surface-to-surface spacing between fuel assemblies of 0.5 in.
( A curface-to-surface spacing of 0.3 in. is optimum for criticality, but the presence of fuel racks results in each assembly essentially being contained in a box constructed from 1/4 in. angle.)
- 3. An infinite array of fuel assemblies, each enriched to 3.1%
U-235 by weight.
- 4. All assemblies completely submerged in water at 68 F and 14.7 psia.
- 5. No poison effects from the presence of stainless steel fuel-racks.
- 6. No burnable poisons or control rod clusters present.
The results of this analysic show that Keff = 0.999 with 2125 ppm boron . At the proposed boron concentration of 4550 ppm, Keff is less than 0.8.
The presence of burnable poisons and control rod clusters, the poisoning effect of the stainless steel fuel racks, and the fact that the array of stored fuel assemblies is finite, all contribute to the conservatism of this analysia.
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Potentiel for Poduction of Boron Concentration f The only m?chanism which could lead to a reduction in boron concen- i tration and the possibility of criticality is the addition of sufficient water to the pool in a manner which would reduce the boron ecncentration, at least locally, to a value less than 2125 ppm. All potential sources of such water 5 have been considered in order to identify any mechanism which might lower borun concentration to this level. The potential sources of water have been categorized as follors:
- 1. Rainfall
- 2. The raw water reservoir
- 3. Storage tanks .
i 4 Piping above elevation 140 f t.
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- 5. Piping below elevation 140 f t.
I Rainf all - Rainfall, in combination with a tornado or seismdc event which l dsnages the roof over the spent fuel pool, could result in the addition of l water to the pool and a reduction in boron concentration. If a loss of fuel assembly spacing is also assumed to occur, the possibility of criticality becomes e consideration.
1 In the event of a tornado in combination with rainfall, it was assumed .
that the roof over the spent fuel pool was damaged or removed resulting j in all rainfall falling on the roof area enterinE the spent fuel pool. It was further assumed that 5 f t, of borated water was removed from the spent fuel by the tornado (See G.E. APED-5696) . For the probable maximum pre- i cipitation given in Section 2.4 of the Diablo Canyon FSAR, 26 hours3.009259e-4 days <br />0.00722 hours <br />4.298942e-5 weeks <br />9.893e-6 months <br /> would be required to fill the spent fuel pool to the deck elevation of 140 ft.
Assuming complete mixing, the boron concentration in the pool would still be above 3667 ppm and Keff would be below 0.84. Further rainfall would be expected to overflow from the pool with very little mixing.
Rainf all in combination with a seismic event would have less effect on boron concentration than the combination of rainfall and tornado, since the tornado was assumed to remove the upper 5 ft, of borated water in the spent fuel pool.
Raw Water Reservoir - The possibility of reservoir failure and the poten-tial for flooding has previously been considered and is discussed in the Diablo Canyon FSAR (Sections 2.4.4, 2.5.5, and 9.2.4). Although PG&E believes that reservoir failure as a result of a seismic event is unlike,1y, such an event has been considered. Because of the drainage paths available to accommodate water, the distance between the reservoir and the spent fuel w___-_-__________________________-_____________ __ _. _ - _ . ___-____a
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pool, and the elevation difference of 25 ft between the edge of the spent fuel pecl and the ground elevation at the east side of the fuel handling building, reservoir failure would not result in water entering the spent fuel pool .
Storare Tanke - There are two storage tanks which are potential f1 coding hazards to the fuel hand 3ing building; the refueling water storage tank, which could contain up to 250,000 gallons of water above elevation 140 f t. ,
and the condensate storage tank, which could contain up to 200,000 gallons of water above elevation 140 f t. Both of these tanks are Design Class I.
If it is assumed that a seismic event causes these tanks to fail, failure would be predicted to occur at the points of highest stress near the bottom of the tanks. Such f ailures would not result in water entering the spent fuel pool.
l A failure mode has been postulated which could result in water entering j
the spent fuel pool. It was assumed that the entire superstructure of the i fuel handling building fails as a result of a seismic event, that the main roof support beens collapse or deform sufficiently to allow a tank to bend over at an angle of 30o, and that the tank contacts the concrete deck at elevation 140 f t., and bursts open along a weld seam. It ras further assumed that no loss of water other than that spilling frc the tank onto the 140 f t, deck results. In the event of such a f ailur= , water would fill the 2} ft. portion of the spent fuel pool between the n:rmal operating le cel and the 140 f t deck and then spill over into the cask de-contandnation area, over edges of the deck, and down stairways. Mis?"g in the spent fuel pool would be essentially confined to U1e upper 10 ft. of the pool. Since more than 27 ft. of borated water covers the top :f. the fuel, the boron concentration would remain above 4000 ppm and Keff below 0.81.
1 Pining Abovr E3evation 140 Ft. - There are two 2 in, fire lines and one 1 in. service water line above elevation 140 f t. whose f ailure could result in water entering the spent fuel pool. The fire lines are noroally pressurized.
The service water line is intermittently pressurized. For purposes of this safety evaluation, all three lines were assumed to be pressurized and failure due to a seismic event was assumed. It was further assumei that l
all water discharged from broken lines into the spent fuel pool. The maximum flow possible from these lines is 880 GPM. If complete mixing is assumed in the spent fuel pool, 4y hours would be required to reach a boron concentration of 2600 ppm and a Keff = 6.945. Several means are available to isolate these lines and terminate flow into the spent fuel pool. Our proposed addition of another isolation valve in the fire main would make it poesible to isolate these two 2 in, fire lines without iso-lating other portions of the fire system.
The assumption of complete mixing is conservative, since mixing would be confined largely to the upper portion of the pool. In any event, there is more than adequate time available to isolate these lines before boron.
concentration is reduced sufficiently for criticality to occur.
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I PipinF Polow r'.evation 1/,0 Ft. - A number of lines carrying water are routed below elevation 143 f t. in the. vicinity of. the spent fuel p0:1.
These lines are separated from the pool vertically by at least one con- i crete deck srd horizontally by the spent fuel pool wall. Potential j failure of these lines has been considered. Storage volumes and drainage paths acailat'_e and tlhe elevetion differnnce between these lines and the ;
top of 'the spent fuel pool are more than sufficient to prevent v;ater 1 from reaching elevation 140 f t. and entering the spent fuel pool . Die presence of at least one deck between th se lines and the 140 ft. elevation prevents water discharging from a broken line from spre.ving upward and reaching the spent fuel pool.
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