ML20210J605
| ML20210J605 | |
| Person / Time | |
|---|---|
| Site: | 05000447 |
| Issue date: | 09/30/1976 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-0124, NUREG-0124-S01, NUREG-124, NUREG-124-S1, NUDOCS 9708180158 | |
| Download: ML20210J605 (63) | |
Text
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NUREG 0124 h (Suppl.1 to NUREG 75/110)
NIIllII hIIl DlI$II Regulatory o m5s fo[i rcicted to the preliminary design of the Office of Nuclear Reactor Regulation i GESSAR-238 Nuclear Island
- Docket No STN 50-447 Standard Des.ign SEPTEMBER 1976 Gcneral Electric Company Cupplement No.1 1
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NUREC-0124 Supp. 1 September _1976 SUPPLEMENTNO.1 TO THE_
SAFETY EVALUATION REPORT BY THE OFFICE OF NUCLEAR REACTOR REGULATION U. S. NUCLEAR REGULATORY COMMISSION IN THE MATTER OF GENERAL ELECTRIC i STANDARD SAFETY ANALYSIS REPORT (GESSAR-238 NUCLEAR ISLAND)
DOCKET NO. STN 50-447 I
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i TABLE OF CONTENTS Page 1 1.0 INTRODUCT !DN AND GENER AL DESCRIPTION OF THE PL ANT . . . . . . .'. . . . . . . . . . . . . . . . . . . . . .
1-1 1.1 Introduction...........................................................
1.7 f acility Hodifications as a Result of Regulatory Staff Review............ 1-1 1.7.2 racility Modifications Required by the 5taff...................... 1-1 r
1-2 1.8 Requi rements f or Future Techni cal information. . . . . . . . . . . . . . . . . . . . . . . . . . . .
1-2 1.8.2 Post PDA Review...................................................
17 1.1 1 Co n c l u s i o n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
3-1 3.0 DESIGN OF STRUCTURES, SYSTEMS, AND COMP 0NENTS.................................
3.2 cla ssificat ion of Struc tures, Systern , and Componen's. . . . . . . . . . . . . . . . . . . . 3-1 3-1 3.2.1 Seismic Classification............................................
3.5 M i s s il e Protec ti on C ri teri a . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32
' 5.3 T o rn a do M i s s 11 e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-2 1
+? r.echa ni ca l Sys tems and Components . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . , 3-3 3.9.1 Dynamic System Analysis and Testing............................... 3-3 3.10 Seismic Design of Category I instrumentation and Electrical Equipment.... 3-3 3.11 Environmental Design of Mechanical and Electrical Equipment.............. 3-4 3.11.1 Electrical Penetrations in the Shield Building, Con t a i nmen t a nd Drywel l Wa 11 s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-5 3.12 Separation Criteria for Safety-Related Mechanical and Electrical Equipment................................................. 3-5 4.0 REACT 0R....................................................................... 4-1 i
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Pagg l 4 . 2 - Mec ha n i c a l De s i g n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-1 1
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4.2.3 f a s t S c r am . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 4.3 N uc l e a r De s 1 g n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 41 4.3.7 Analytical Methods................................................ 41 5.0 REACTOR COOLANT SYSTEM........................................................ 51 5.4 -Component and Subsystem Design........................................... 5-1 5.4.5 Residual Heat Removal System...................................... 5-1 6.0 ENGINEERED SAFETY FEATURES.................................................... 6-1 6.2 Containment Systems...................................................... 61 6.2.1 Containment Functional Design..................................... 6-1 1 i
6.2.3 Secondary Contai nment Func t ional 0es ign. . . . . . . . . . . . . . . . . . . . . . . . . . . 6-4 6.2.4 Co n ta i nmen t i so l a t i on Sy s t em. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-5 l
6.3 Emergency Core Cooling System............................................ 6-6 6.3.1 System Description................................................ 6-6 7.0 I NST RUME NT AT I ON AND CO NT R0L S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ~1-7.1 Introduction............................................................, 7-1 7.2 Reactor Trip System...................................................... 71 7.2.1 Fa s t S c ra m . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-8 ,
7,3 Engineered Sa fety Fea tu res Sys tem. . . . . . . . . . . . . . . . . . '. . . . . . . . . . . . . . . . . . . . . . 79 7.3.1 Introduction............ ......................................... 7-9 '
7.3.2. Emergency Core Cooling System..................................... 7-9 7.3.3 - Containment and Reactor Vessel isolation Control System........... 7-12 7.3.4 Essential Service Water System.................................... 7-13 7.3.5 Fl amma bi l i ty Con t rol Sys tem. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-13 7.3.6 S t a ndby Ga s T rea tmen t Sy s t em . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-14
- 7.3.7 Suppression Pool Makeup System.................................... 7-14 7.3,8 Co nta i nmen t S p ray Sy s t em . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-14 7.3.9 I n d i ca t i on o f Bypa s s e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-14 11
[a21 7.4 Sa f e S hu t down Sy s t em. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-14 7.4.1 Reac tor Core I solation Cooling System. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 14 7.4.2 Standby Liquid Control 5ys'em..................................... 7-15 ,
7.5.. Safety-Related Display Instrunentation................................... 7-15 7.5.1 General............. .... ........................................ 7-15 7.5.2 Nuc1enet.......................................................... 7 16 7.5.3 Power Genera tion Con t rol Comple x . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-16 7.6- All Other Systems Required for Safety.................................... 7 17
- 7. 6.1 Genera 1........................................................... 7-17 7.6.2 Reac to* Pressure Relief Instrumenta tion. . . . . . . . . . . . . . . . . . . . . . . . . . . 7-18 7.6.3 Reci rcula tion Pump Tr ip System. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-19 7.7 C o n t ro l S y s t e- s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-19 7.7.1 Rod Control and Information System and Instrumentation............ 7 19 7.7.2 O t he r Co n t ro l $y s t ems . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-21 7.8 Instrurentation Interfaces with Balance-of-Plant Systems................. 7-22 8.0 ELECTRICAL POWER SYSTEMS.. ................................. ................. 8-1 9.0 AUKILIARY SYSTEMS...... ....................................................., 9-1 9.3 Process Auxiliaries...........................,.......................... 9-1 9.1.1 Main Steam isolation Valve Leakage Control System............ .... 9-1 11.0 RADIDACTIVE WASTE MANAGEMEN1.................. .......................... .... 11-1
- 11. 3 Ga seou s Wa s te T r ea tmen t Sy s t em. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11-1 11.3.3 Conclusions................................................ ...... 11-1
- 15. 0 A C C I DE NT ANAL Y S 15. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15-1
..15.2 Abnormal Operational Transients.......................................... 15-1 15.4 An tic i pa ted Trans ients Wi thout Scrain. . . . . . . . . . . -. . . . . . . . . . . . . . . . . . . . . . . 15 2 15.5 Failure of inputs From Turbine Butiding to Reactor Protection System..... 15-3 111
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APPENDICES i bt
. APPENDIX A N
'CO TINUAT ION OT CHRONOLOGY OF RA010 LOGICAL REV J tW. . . . . . . . . . . . .
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LIST Of TABLES Page 1-3 1ABLE 1-1 POST ' ITEMS......................................................
TABLE 1-2 SAFETY MATTERS THAT REMAIN TO BE RESOLVE 0 PRIOR TO A DECISION ON 1-6 THE ISSUANCE OF A CONSTRUCTION PERMIT TO A REFERENCING PLANT.....
6-3 TABLE 6 1 QUENCHER BUBBLE PRESSURE FOR THE CESSAR-238 NUCLEAR ISLAND DESIGN...
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l llST OF FIGURES Page FIGURE 7-1 PREVIOUS GENE RAL ELECTRIC COMPANY DESIGN. . . 72 ......................
FIGURE 7 2 GCSSAR-238 REACTOR TRIP SYSTEM L0GIC. . . . . . . . . . 73 ......................
FIGURE 7 3 RECIRCULATION PUtiP TRIP 5YSTEM...................................... 7-20 t
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1.0 INTRODUCTION
AND GENERAL DESCRIPTION OF THE PLANT 1.1 Introduction On December 22,1975, the United States Nuclear Regulatory Cornission issued the Safety Evaluation Report (NUREG 75/110) and the Preliminary Design Approval for the General Electric Standard Safety Analysis Report (GESSAR-238 Nuclear Island) design (Docket Number STN 50-447). In our Safety Evaluation Report on the GESSAR-238 Nuclear Island design, we identified nineteen items (in Table 1-3) that we indicated would require continued review af ter the issuance of the GESSAR 238 Nuclear Island Preliminary Design Approval and four staff design requirements imposed as conditions to the GESSAR 238 Nuclear Island Preliminary Design Approval.
Since the issuance of the Safety Evaluation Report, the General Electric Company has sutaitted five amendments (Amendments 40, 41, 42, 43 and 44) to the GESSAR-238 Nuclear Island Safety Analysis Report. The purpose of Supplement Number 1 to the Safety Evaluation Report is to update the Safety Evaluation Report by providing the staff's evaluation of the additional information received since the issuance of the Safety Evaluation Report. Each of the following sections in this supplement is numbered the same as the section of the Safety Evaluation Report that is being updated.
Appendix A to this supplenent is a continuation of the chronology section of the Safety Evaluation Report.
1.7 racility Modifications as a Result of Regulatory Staff Review 1.7.2 Facility Modifications Required by the Staf f in the Safety fvaluation Report, we identified four staff requirements which were imposed as conditions to the Preliminary Design Approval for the GESSAR 238 Nuclear Island design. Since that time, the General Electric Company has pro-vided us with acceptable connitments to the staff positions on two of these conditions, and an acceotable resolution to a third condition.
The statyt of the four conditions and the section in this suppleinent where each condition is discussed are:
(1) Tornado missile velocities - Resolved (See Section 3.5).
(2) Containment pool dynamics - Under review (See Section 6.2.1.9),
(3) Continuous purging of containment - Resolved (See Sections 6.2.4 and 11.3).
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(4) Main steam isolation valve leakage control system - Resolved (See Sections f,.2.3and9.3.1).
1.8 Requirements for Future Technical Information 1.8.2 Post-PDA Review Table 1-3 of the Safety Evaluation Report for the GESSAR 238 Nuclear Island contained a list of items which we planned to continue to review af ter the issuance of the Preliminary Design Approval. Table 1-3 contained several items which are not normally reviewed at the construction pemit stage of review along with those items which would nortnally be reviewed. However, at that time, no distinction was made between these two types of items.
Since the issuance of the Safety Evaluation Report, we have been working with the General Electric Company in an effort to complete the review of these nineteen items. During the course of our review, it became apparent that the list pre-sented in Table 1-3 was inappropriate to identify the work being done to satisfy the requirements for the issuance of a construction pemit to a referencing plant since:
(1) Table 1-3 identifies several items which we plan to continue to review on the GESSAR 238 Nuclear Island application although sufficient information exists at the present time for the issuance of a construction pemit.
(2) Item 14 in Table 1-3 is in actuality a group of issues concerning a particular area of review. In order to clearly understand the extent of the outstand-ing safety matters, item 14 should be separated into individual items requiring resolution.
For improved ease of use, we have decided to separate Table 1-3 in the Safety Evaluation Report into two diffcrent lists. The first. Table 1 1. identifies the same items as Tabla 1-3 of the Safety Evaluation Report and shows the present status of each item. The status for thirteen of the items in Table 1-1 is given as " Acceptable for construction permit stage of review" and a section of this supplecent is referenced where a discussion is provided on whether the review is completed or whether resolution of the item is not required for the issuance of a construction permit. For the remaining six items, the current review status is listed.
In Table 1-2, we have provided a detailed list of those specific items which remain outstanding. as of the date of this supplement, and which require resolu-tion prior to the issuance of a construction permit to a referencing plant.
Following each item. a reference to a section in this supplement is provided where the item is discussOd in detail.
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N TABLE 1-1 POST-PDA ITEMS Status Discussed in Section Item Analysis of consequences 3.2.1 , provided July 1976; under staf f review
- 1. Leakage characteristics of prirary coolant pum seals Acceptable for construction 3.9.1.4 permit stage of review
- 2. Description of cortbined effects of safe shutdown earthquake and steam line break
- 3. a. List of specific equipmer.t to be seismically qualifled 3.10 Acceptable for construction permit stage of. review
- b. The qualification r:ccedures to be used 4 .a. 1.ist of specific equipment to be environ-mentally qualified 3.11 Acceptable for construction permit stage of review
- b. The qualification procedures to be used Acceptable for construction I 3.11.1
- 5. ' Preliminary design of drywell penetrations permit stage of review !
I 3.11.1 Acceptable for construction permit l O" 6. Procedures and methods to be used to qualify stage of reveiw the shield building, containment and drywell penetrations 3.12 Staff review to be accomplished coincident 7, ' implementation methods of separation criteria with review of item 14 for safety-related electrical equipment 4.3.7 Acceptable for construction
- 8. Detailed information on: perwit stage of review
- a. Lattice physics methods.
- b. Boiling water reactor simulation code.
'c, Verification of core calculational methods.
6.2.1.6 Acceptable for construction
- 9. Confirming data from large scale' permit stage of review Mark Ill tests for short-tern containment response, l
TABLE 1-1 (Continued)
Item Discussed in Section Status
- 10. Assumptions used to size 6.2.1.5 Response to staff concems containment vacuum breakers provided July 1976; under staff review
- 11. Environmental design criteria for 6.2.4 Acceptable for construction isolation valves and other safety-related pemit stage of review equipment in the drywell
- 12. Address question 6.125 (manual operator 6.3.1 Acceptable for construction action on emergency core cooling systems pemit stage of review following a loss-of-coolant accident)
- 13. Proprietary version of 8 x 8 6.3.1 Acceptable for construction zirconium spray cooling test pemit stay of review 14 New modified instrumentation and cw. trol systems 7.0 Areas a through f are acceptable preliminary design review for: for the construction permit stage a.
of review with the exception of Reactor trip system sub-items 3 through 12 in Table 1-2; b.
The July 1976 submittal on these i Engineered safety features actuation system items is under review
- c. Safe shutdown system
- d. Safety related display instricentation
- e. All other instrumentation required for safety
- f. Control systems - reactor manual control system, recirculation flow control, gaseous and itquid rad-waste control, feedwater flow control and inter-action between safety and non-safety control '
systems
- 15. Scope of onsite electrical system 8.0 July 1976 submittal under review
- 16. Review of high pressure core spray power 8.0 system Being reviewed separately as Topical Report NEDO-10905, "High Pressure Core Spray Power Supply Unit"
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l TABLE l-2 SAFETY MATTER $ THAT REMAIN TO BE RE501.VE0 PRIOR TO A DECISION Ort TjiE_ ISSUANCE._OF A__ EONSTRUCTION PERMIT FOR A REFERENCING PLANT (The number in parentheses following each item denotes the item in Table 1-1 ueer which this area was previously classified) ltem Olscussed in Section
- 1. Leakage characteristics of primary coolantpumpseals(1) 3.2.1 ,
- 2. Containment vacuum breakers (10) 6.2.1.5 Automatic depressuritation system testability (14) 7.3.2.2 3.
4 High water level trip (14) 7.2 Two pump trip (14) 7.6.3 5.
- 6. ' Low pressure core spray and low pressure (14) coolant injection system interlocks (14) 7.3.2.3 Electrical review of essential service water system U,14) 7.3.4 7.
Electrical review of flamability control system (7,14) 7.3.5 8.
Electrical review of standby gas treatment system (7, 14) 7,3.6 9.
- 10. Electrical review of suppression pool makeup system (7,14) 7.3.7
- 11. Electrical review of containment spray system (7, 14) 7.3.8
- 12. Review of Nuclear Island / balance-of-plant electrical (14) interfaces 7.8
- 13. Review of onsite power systems (15) 8.0 16
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1.11 Conclusion Based on our review of the information submitted by the General Electric Company since the issuance of the Preliminary Design Approval, we conclude that, except for those items identified in Table 1 2 of this supplement, the General Electric Company has supplied sufficient information on the post-PDA items identified in Table 1-3 of the Safety Evaluation Report 13 provide a suitable basis for the in addition, we issuance of a construction permit to a referencing plant, conclude that the coninitments provided by the General Electric Company, in Amendment 43 to the GES$AR-238 huclet.' island Safety Analysis Report, provide an acceptable resolution to the Preliminary Design Approval conditions with the exception of Condition Two identified in Section 1.7.2.
We will address the resolution of Condition Two to the Preliminary Design Approval and those items in Table 1-2 of this supplement in a future supplement to the Safety Evaluation Report prior to the issuance of a construction permit to a referencing plant.
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d 3.0_- DESIGN OF STRUCTURES. $YSTEMS AND COMPONENTS 3.2 - 11assification of Structures. Systems and Components 3.2.1 -seismic Classif1(ation i
In Section 3.2.1 of the Safety. Evaluation Report 'we provided a discussion on j two systems which we felt did not have the proper seismic classification. These were the r,rimary coolant pump seals' cooling water system and the offgas systen.
We stated that we would continue to review the leakage _ characteristics of the ,
primary coolant through the primary coolant pump seals as a result of assumed loss of seal water or co?1ing water to the__ primary coolant pump seals. Specifical- 1 ly. we requested the General Electric Company to provide en analysis to demon-strate that the consequences of recirculation pump seal failure are acceptable.
The General Electric Compa,ny has since informed us tht the consequences of total failure of the recirculation pump seals' cooling water supply would be a gradual deterioration of the seals over a period of several hours finally result-ingintotalsealfailure. They further stated that seal failure would result in a loss of primary coolant at a rate of less than 50 gallons per minute.
- We' found the General Electric response unacceptable as it was not supported by either tests or analyses. General Electric has since supplied us with the pump vendor's saalysis of the consequences of pump seal cooling water failure. We are reviewing the analysis supplied by General Electric and will report on our conclusions in a future supplement to the Safety Evaluation Report.
In the Safety Evaluation Report, we stated that the General Electric Company had proposed to design the offgas system delay tank supports to the seismic design criteria listed in Effluent Treatment Systens Branch Technical Position 11-1
" Design Guidance for Radioactive Waste Management Systems Installed it Light-Water-Cooled Nuclear Power Reactor Plants." This statement requires fur *her clarification. The General Electric Company. in Amendment 31 to the GESSAR-238-Nuclear Island Safety Analysis Report. modified their application ".o include the following design criterlat "The support elements. _ including the skirts, legs and anchor bolting for the charcoal absorber tanks of the of' gas system shall be designed as.follows:
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(1) The fundamental frequency of the charcoal absorber tanks, including the support elements. _is greater than 33 Hert2.
(2) .The charcoal absorber tanks are mounted on the base mat of the building j; housing the tanks.
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- (3) The charcoal absorber tanks including the support elements are designed
-with a horizontal static coefficient of 0.15g.
(4) The stress levels in the support elements of the charcoal absorber tanks shall not exceed 1.33 times the allowable stress _ levels pennitted by the AISC Manual-
.of $ teel Construction, Seventh Edition.1970."
items 1 and 2 of the General _ Electric Company's design criteria assure that the ground level acceleration resulting from a seismic event will not be magnified t'y the absorber tank supports since the amplification factor associated with structures having fondamental frequencies exceeding 33 Hertz is unity. _-Since the static co-efficient fpecified for the absorter tanks (item 3) is compatible to the operating bases earthquake intensity specified for the CE$$AR 238 Huclear Island design, and since item 4 of the General Electric Company's design criteria is identical to the stress criteria specified in Effluent Treatment Systems Branch Technical Position 11
- 1. we conclude that the General Electric Company's design criteria are as conserva.
tive as those specified in Effluent Treatment Systems Branch Technical Position 11-1 and are therefore acceptable. Since the issuance of the Safety Evaluation Report, we have recalculated the offsite dose which could result from a failure of the offgas tank supports (assuming a relative concentration value of I x 10'3 seconds per cubic meter which is consistent with the relative concentration value used for the 0-2 hour accident analysis calculations) and found it to be less than two rem.
Since the offgas tank supports are being designed with a seismic capability to meet the operating basis earthquake requirements, we would postulate a failure of the of fges tank supports only for seismic events exceeding the intensity of the operating basis earthquake.. events which have a low probability of occurrence. $1nce the probability of the failure of these supports is low, and since the consequences of failure is a relatively small fraction of 10 CFR Part 100 guidelines on whole body doses, we conclude that the intermediate level of seismic design is acceptable for the offgas tank supports.
3.5_ Missile' protection Criteria 3.5.3 Tornado Missiles in the Safety Evaluation Report, we stated that the tornado missile velocity spectrum
- proposed by the General Electric Company was unacceptable as it was unconservatively low and unsupported by data. We therefore required the General Electric Company to adopt our tornado missile velocity spectrum as a design basis for the GESSAR-238 Ncclear Island design. This requirement was made a condition of the Preliminary Design Approval issued to the General Electric Company.
~ The General Electric Company has since modified their application with Amenenent -
43 to the GESSAR.h8 Nuclear Island Safety Analysis Report. to incorporate our 32
position on tornado missile velocities as a design basis for the CE55AR 238 Nuclear Island design. We therefore consider this conditior- resolved.
3.9 Mechanical Systems and Components 3.5.1 Dnamic System Analysis and Testing 3.9.1.4 Analysis Methods for loss-of toolant Acr.ident loadings in the $afety Evaluation Report, we indicated that General Electric was to provide us with a dynamic analysis of the ccenbined effects of a safe shutdown earthquake plus a steam line break u the reactor internals to confirm the structural adequacy of the internals. The results of this analysis will be used to demonstrate the conservatim present in the design loads for the internals. The methods and models for determin-ing these loads have, however, already been sutaitted and approved. That provides an acceptable basis for the issuance of the Preliminary Design Approval to General Electric and is, therefore. acceptable for use by a referencing applicant in a Con-struction permit application. We will report on the results of the dynamic analysis during our review of the final design.
3.10 Seismic Design of Category i Instrumentation and Electrical Equipment in the Safety Evaluation Report. we identified the following as open items to be resolved as part of our review following the issuance of tre Preliminary Design Approval:
(1) The list of equipuent to be seismically qualified.
(?) The qualification procedures to be used.
The General Electric Company has resolved these items in the following manner. It has been agreed between the General Electric Company and the staff that. for the final design stage of review, we will conduct a detailed review of the methods and procedures used in implementing the seiwic design criteria and of the scope of the seismic qualification program.
Pursuant to this detailed review:
(1) The General Electric Company has Sutnitted a typical list of equipment which will be seismically qualified.
(2) The General Electric Company will sutinit. for review and approval, a topical report which will describe the detailed qualification procedures which will be used in implementing the seismic design qualification criteria.
(3) The General Electric Company will sutait to an audit, by the staff. of the qualification of selected instrumentation and electrical equipment.
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Acceptable implementation of criteria described in the detailed procedures of item 2 will tie verified during this audit, We have concluded that these comitments provide an acceptable basis for the Pre-liminary Design Approval.
3.11 inviromental_ Des _1 3 n of Mechanical __and Electrical Equipment in the $afety Ivaluation Report, we identified the following as open items to be resolved as part of our review following the issuance of the Preliminary Design Approval:
(1) The list of electrical equipment to be environmentally qualified.
(2) The qualification procedures to be used, In Amendment 41 to the GES$AR-23B Nuclear Island Safety Analysis Report, the General Electric Company responded to all of our outstanding requests for additional infor-nation concerning environmental qualification of electrical equipment, in Tables 3.11,1 through 3.11.4 of GESSAR 238 Nuclear Island Safety Analysis Report, they have provided the expected post accident themal and radiological conditions for safety-related equipment within the drywell, containnent and auxiliary building, With regard to the environmental qualification of Class lE equipnent, the General Electric Company has cammitted to conformance to the requirements of the Institute of Electrical and Electronic Engineers Standards 323-1974, " Qualifying Class it Electrical
[quipment for Nuclear Power Generating Stations
- and has supplied their qualifica-tion test program which describes thei, proposed method of confonning to this standard.
The program outlined by the General Electric Company includes temperature-humidity testing, seismic testing and life testing (aging) in a sequence which is in confonn-ance with the requirements. The justification for excluding any environmental parameter in the testing will be provided in the qualification report for the specific equipment involved.
The above conriitments and outlined program are suitable bases for the developnent of the more detailed qualification program, We have concluded that this is acceptable for this stage of review and that there is reasonable assurance that General Electric can develop a detailed qualification program acceptable to the staff, We will repor on the results of uur evaluation of the detailed qualification program at the fini design stage of review, 34
3.11.1- Electrical Penetrations in the Shi_ eld __Bu_ilding. Containment and Drywell Walls
_In the Safety Evaluation Report, we identified two items in which the General Electric Company had not provided adequate information for the review to t>e completed. Ttese were the procedures and methods to be used to qualify the Shield building, contain-ment and drywell penetrations, and the preliminary design for the drywell penetrations.
With regard to the qualification of the shield building and containment penetrations, the General Electric Company meets the requirements of the Institute of Electrical and Electronics Engineers Standard 317-1974. " Standard for Electrical Penetration ~
Assemblies in Containment $tructures for Nuclear Power Generating Stations." for the containment penetrations and for the electrical considerations associated with the shield building penetrations. We find this acceptable.
With respect to the drywell penetrations, the General Electric Company has provided us with a preliminary design for the drywell penetrations and a description of their proposed qualification program.
The qualification program for the drywell penetrations includes results from pre.
viously conducted tests.to demonstrate the sealing capability of the potting material used in the penetrations. Once the penetrations are constructed, the drywell structural proof test and the periodic drywell leak tests will provide continued assurance that the penetrations are intact.
Based on the previous discussion, we conclude that the design and testing provisions for the electrical penetration in the shield building, containment and drywell are acceptable.
3.12 Sfparation Criteria for Safety-Related Mechanical and Electrical Equipment in the Saf ety Evaluation Report, we concluded that the General Electric Company's proposed design criteria for the separation of safety-related equipment met our requirements. However, the review of the General Electric Company's implernenta-tion of these criteria had to await tne subnission of the preliminary design of many of the instrumentation systems. The General Electric Company has since sutrittrd this inforrration in Amendment 41 to the GESSAR 238 Nuclear Island Safety Analysis Report.
We have utilized Regulatory Guide 1.75. " Physical Independence of Electrical Systems."
in our revi,w of the preliminary designs (elementary wiring diagrams) for the instru-mentation and control systems. The specific conclusions for these systems are reported in the applicable portions of Section 7,0 of this supplement.
All the instrumentation and control preliminary designs utilized isolation devices wherc signal 3 are transmitted tetween redundant divisions of equipnent 3-5
- and where signals are transmitted frcel division equipnent to non divisional equip.
ment. . Though isolation devices have been employed in the manner on other plant designs, we required the Genero. Electric Company to distinctly identify the type of devices they intend to use and to define the test program which will be utili2ed to l demonstrate the isolation capabilities of these devices, lhe General Electric Company has since provided us with the test program which they intend 19 use to qualify optical isolators, which is the isolation device they will uttltre to isolate signals within their proposed solid state protection system. (An optical isolator is basically a light emitting diode and a light sensitive diode separated by a transparent membrane.) We have reviewed the proposec test program and canclude that it provides a suitable basis for qualifying the isolation devices. We will report on the results of this test program for isolation devices at the final design stage of review.
We have not completed our review of all the systems which are outside the GESSAR-238 Nuclear $ team Supply System scope of design yet inside the GESSAR.238 Nuclear Island scope of design. The systems involved are the essential service water system, the standby gas treatment system. the flammability control system, the contairinent spray system. the suppression pool makeup system and the onsite power system. We will re y rt on our conclusions concerning the physical independence of these systems in the applicable portions of Section 7.0 of a future supplement to the Safety Evalva- ,
tion Report.
We conclude that the electrical. instrumentation and control systerrs within the GESSAR 238 Nuclear Steam Supply System portion cf the GESSAR-238 Nuclear Island design have adequate separation of redundant safety related equipment.
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4.0' REACTOR 4.2 i ; _ Mechanical Design 4.2.3 Fas t $ cram, -
In the Safety Evaluation Report we stated that we would require General Electric' toi provide a detailed description of (and comitment to) a test program for the fast scram system.
.The General Electric Company has comitted to a test progr 'n for the fast scram
-. System to verify the design bases used in the analyses in Section 15.0 of the ,
GE55AR.238 Nuclear Island Safety Analysis Report. - The test program will consist of ;
four phases:
=(1) Developnent Testing
- This phase of the program is complete and was conducted to identify and optim12e those perfomance objectives for the fast scram system.
--(2) Design Acceptance Testing - This phase of the program is scheduled for completion in 1976 and is intended to verify the final design using, as close as possible, actual production control rod drive components.
(3) Production Qualification Testing This phase of the program is scheduled for completion in 1917 and is intended to establish a firm statistical base on control rod drive perfomance by testing preproduction control rod drives to the design requirenents.
(4)-. Production Verification Testing - This phase of the program is scheduled for completion in 1978 and is intended to verify that the procedures and techniques used in manufacturing produce control rod drives that meet all of the requirements.
We have reviewed the test program and conclude that, if adequately implemented. it will verify the design objectives for the fast scram system. We will review the various phases of the test program and report the results of our evaluation at the ,
final assign stage of review. We conclude that the proposed test program is acceptable for the Preliminary Design Approval.
4.3. Nuclear Design-4.3.7 Analytical Methody
~
< in the Safety Evaluation Report we stated that' General Electric had comitted to
,1 provide us with a series of-topical reports to address lattice physics methods, the ,
!' boiling water reactor simulation code, and verification of lattice physics and core calculational methods, o
41
These topical reports deal with the final design details of the core configuration and were not needed to reach the conclusion that an appropriate core design can be developed for the final design. We will report on our conclusions concerning these reports during the final design stage of review.
42
_. - ___ __ . . _ m _ _ . _ _ _ . __ _ . _ -.
5.0 ptACTOR COOLANT SYSTEM 5.4 gponentandsubsystemDesign, 5.4.5 pesidual Heat Removal system During our review of the GESSAR 238 fiuclear Island design, we noted that the residual heat removal system could be disabled by a single failure in the shutdown cooling mode. General Electric agreed to provide an alternate method of achieving cold shutdown which employed safety grade equipment and which was not susceptible to the same single f ailure which could disable the residual heat removal syste.n. .This comitment formed the basis for the issuance of the preliminary Design Approval to General Electric.
$1nce that t he General Electric has described how they plan to achieve cold shutdown-in the event of failure of the residual heat removal system. This involves the use of an alternate heat removal path through the automatic depressuritation system valves to the suppression pool. We have reviewed the proposed method and conclude the following; (1) The alternate path employs safety-grade equipment.
(2) ho single failure will disable both the residual heat removal system and the alternate heat removal path.
(3) Either the residual heat removal system or the alternate shutdown path is capable of bringing the plant to cold shutdown assuming the loss of either onsite or offsite power and taking credit only for those actions capable of being perfonned from the control room.
We therefore conclude that the proposed residual heat _ removal system design combined with the alternate shutdown path is acceptable.
51
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6.0 . ENGINEERED SAFETY FEATURES 6,2_ -Eg taintnent Systems.
f6.2.11 Containment Functional Design
- 6. 2.1.5 . External Pressure DesigqE in the Safety Evaluation Report we stated that certain of General Electric's assump-tions used in the string aralyses for.the containment vacuum breakers may not be sufficiently conservative. We have continued to review this item with General Electric and they have since revised their analysis. We have this information under review and will report our findings in a future supplement to the Safety Evaluation Report.
6.2.1.6 Test program
.In the safety Evaluation Report we stated that we would require the General Electric Company to complete a series of confirmatory tests on the Mark !!! design prior to the issuance of an operating license for any Mark til plant, These tests do not
= represent design governing conditions nor are they expected to escalate into design basis considerations. The design criteria for the GESSAR-238 Nuclear Island con-tainment were previously established by tests and analyses as discussed in Section 6.2 of the Safety Evaluation Report, We will review the results of these tests, when submitted, and will report our find-ings at the final design stage of review, 6.2.1.9 Pool Dynamics In the Safety Lyaluation Report we stated that the General Elect.'ic Company had agreed to comply with the staff's load criteria for structures' located within and
- above the suppression pool with the exception of two areasi (1) - The dynamic loads ger.erated during the clearing of the safety relief valve dis-charge lines.
(2) The impact loads on pipes at elevations between 17 and 19.5 feet.
These two exceptions were made conditions of the Preliminary Design Approval issued to the General Electric Company,-
.The General Electric Company has since.provided us with additional information, in Amendment 43 to the GES$AR.238 Nuclear Island Safety Analysis Report, in an effort to provide a basis for the use of a different type of safety / relief valve discharge
_ ~
device fquencher) which results in lower design loads than the device evaluated by the staff (ramsheae): As a result of our review of Amendment 43, we bave concluded 0+1
that the statistical method proposed by the Gmeral Electric Company and the load criteria shown on Table 61 are acceptable. This conclusion is based on the following:
(1) The 6.ethod has properly treated all available test data and is based essentially on the large-scale data with correction terms that take into account the influence of non '.arge scale variables. Since the large scale tests were perforned in an actual reactor with a suppression containment conceptually similar to the GES$AR 238 Nuclear Island containment, estrapolatfor, from the large scale by statistical technigse is appropriate and ecceptable.
(2) The method has been conducted in a conservative manner. The primary conservatisms are:
4 (a) The calculation is based on the most severe parameters. For example, the maximum air volume initially stored in the line, the maximum initial pool temperature and the highest primary system pressure were selected to estab-lish quencher load criteria.
(b) for the cases of multiple val"e actuation, the load criteria are based on the assumption that the maximum pressures resulting fron each valve will occur simultaneously. We believe that the assumption is conservative since different lengths of line and safety / relief valve pressure set points w'il result in the occurrence of maximum pressures at different times and.
consequently, lower loads. l (3) The proposed load criteria in Table 61 are acceptable. The criteria were established by using a 95-95 percent confidence limit. Our consultant, the Broothaven National Laboratory, has performed an analysis of the effect of this confidence limit. The result of this analysis indicates that for a 95-95 percent confidence limit, approximately one percent of the number of safety / relief valve actuations may result in containment loads above the design value. lie believe that this low probability is acceptable considering the conservatism of the method of prediction. The actual loads should not exceed the design value.
(4) With regard to the subsequent actuation, the load criteria are based upon a single safety / relief value actuation. The General Electric Company has estab-lished this basis by regrouping the safety / relief valves in each group of pressure set points. As indicated in Anendment 43, there are three groups of pressure set points for the 19 safety / relief valves--namely, one safety / relief valve at a pressure set point of 1103 pounds per square inch gauge, 9 safety /
relief valves at 1113 pounds per square inch gauge, and the remaining 9 safety /
relief valves at 1123 pounds per square inch gauge. Only one safety / relief valve is now set at the lowest pressure set point. Based on this pressure set point arrangement for the 19 safety / relief valves, the General Electric Company has analyzed the most severe primary pressure transient--a turbine trip without bypass. Results of the analysis show that initiation of reactor isolation will 6-2
k TABLE 6-1 QUENCHER BUBBLE PRESSURE FOR THE GESSAR 238 NUCLEAR ISLAND DESIGN 95-95 PERCENT CONFIDENCE LEVEL Design Value ikximum Pressure (pounds per square inch differential)
Case Description Politive pressure Negative pressure
- 1. - Single valve first actuation at 100 degree' fahrenheit pool temperature 13.5 -8.1
- 2. t Single valve subsequent actuation at 120 degree Fahrenheit pool tenperature 28.2 -12.0
- 3. Two adjacent valves first actuation at 100 degree Fahrenheit temperature 13.5 8.1 4 10 valves (one low set and nine next
' level low set) first actuation at
, 100 degree fahrenheit pool temperature 16.7 -9.3
- 5. 19 valves (all valve case) first actuation at 100 degree Fahrenheit p531 temperature 18.6 -9.9
- 6. 8 automatic depressurization system valves first actuation at 120 degree Fahrinheit pool temperature 17,4 -10.4 6-3 1
-+ v -v - -
. - . - . - . ~. - - . -- . -- .
activate all or a portion of the 19 safety / relief valves which will release the !
stored energy in the primary system. Following the initial blowdown, the energy l
- generated in the primary system consists primarily of decay heat which will l cause the lowest set safety / relief valve to reopen and reclose (subsequent actuation). The time duration between subsequent actuation was calculated to be 1 a minimum of 62 seconds and increasing with each actuation. The tire duration of each blowdown decreases from 51 seconds for the initial blowdown to three seconds at the end of the period of subsequent actuations, which is 30 minutes af ter in_itiation of reactor 1 solation. ;
We find the result of the General (lectric Company analysis reasonable. There-fore, the assumption of only the lowest set safety / relief valve operating in subsequent actuation is justified and acceptable.
We note, however, that the tests performed to date lack complete dynamic or geometric similarity with the Mark Ill quencher system. Therefore, the test results had to be extrapolated. Though the resulting loads are conservative in ccuparison with the test data, we feel that these loads should be verified through in-plant testing.
Therefore, we will require that an in-plant test be conducted, by either General Electric or the utility applicant referencing the GESSAR-238 Nuclear Island design, at the first reference plant to complete construction. If tests are conducted at other Park Ill design plants prior to the completion of construction of 5.he first reference plant, those test results may be sutnitted in lieu of the test requirements I on the reference plant, provided the tests had been conducted at a plant having similar containment parameters.
With regard to the exception on pipe loads between 17 and 19.5 feet, General Electric provided us with additional information in Amendment 40 to the GESSAR-238 Nuclear Island Safety Analysis Report to justify lower loads on these pipes. The staff is reviewing this information to determine the applicability of the scaling relation of the small scale tests to a full scale system. We will report on our conclusions in a future supplement to the Safety Evaluation Report.
6.2.3 Secondary Containment Functional Design in the Safety Evaluation Report we stated that the General Electric Company committed to provide design provisions to eliminate bypass containment leakage. This was to be accomplished through the use of positive leakage control systems, upgrading some piping systems to seismic Category 1.to achieve credit for closed Ic ys or water seals, and identifying those water legs or loop seals which could pmently perform a sealing function. This comitment and functional description formed the basis for the issuance of the preliminary Design Approval to the General Electric Company.
The General Electric Company has since provided, in an April 13, 1976 letter from
- 1. Stuart of the General Electric Company to B. Rusche of the Nuclear Regulatory Comission, specific information on the functioning of the systems used to eliminate bypass containment leakage.
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- ~ _ - - - _ - - . . - - - . - - - ----------
l 1he April 13. 1976 letter also co, taihed a proposal to replace the present testage control systm on the main steam lines with a positive sealing systm. The main l
steam line positive stating systm will te designed in a similar fashion to the ses1 ing system on the other lines penetrating contatreent in that the space betwc-en the inbo6rd and outboard isolation valvst will te pressurlied via a safety grade air !
systte, to a pressure enceeding the peak calculated pressure that the inboard isolation valve will esperience during the accident. Thus, the inward pressure on the isola.
tion valves will always exceed the outwerd pressure, theretiy precluding outward leslage. We conducted a preliminary evaluation of this new systa and concluded that it could resatt in a signif nent decrease in the radiological doses associated w4h The Gene al the loss of. coolant accident and therefore warranted additional redew.
Ilectric Company was informed of our position at a July 1,1976 setting, and of our decision to review the systems dt.igned to preclude bypass contaltnent lea 6 age and the proposed positive sealing systm for the main steam lines, concurrently.
In their suloittal, the General Electric Company addressed each of the lines penetrat.
179 the primary contairment and the system proposed to preclude lealage through each line foi, Ing a loss of coolant accident. However, specific information regarding system actuation times, systm designs, and the justification for taHng credit for certain closed loops to perform a sealing function was not provided.
1he General tiectric ;ompany has $1nce agreed to provide us with a topical report to We will report on the results of our review of this report at address these matter s.
tie final design stage of review. :
The General llectric Company hat, however, provided us with sufficient inforwation on the contatraent bypass scaling systems and the main steam Itne positive sealing s) sten to enable us to conclude that the concept is technically f easible, using T stre the art technology, and that there is reasonable assurance that the General (1ecti . tempony can devetap a final design acceptable to the staff. We conclude that these designs are acceptable for the preliminary Design Approval and are, therefore.
acceptable for use by a referencing applicant in a construction permit application.
We will require that the results of our review of the General (hetric Company topical report be incorporated into the accident analysis calculations performed at the final design stage of review for any plant referencing the Gl$5AR.238 Nuclear Island design.
6.2.4 Containment Isolatinn System J
in the $afety (valuation Report, we stated that we would require the General Electric '
Company to qualify isolation valves tnd other safety related equipment in the drywell for a negative drywell pressure representative of Ilmiting post loss.of. coolant accident conditions. We consider this to be a part of the overall detailed qualift.
cation program. We did not require the General Electric Company to specify a test prassure for the issuance of the preliminary Design Approval, nor will it te required for the issaance of a construction pernit to a referencing plant. We will evaluate 6.b
_ ,.~
the test pressure which the General (lectric Company proposes against the environmental conditions espected in the drywell and report on our conclusions during the fihal design stage of review.
In the $afety (valuation Peport, we noted that it was our position that continuous purging of the containment through large penetrations was undesirable, but that we could find continuous purging acceptable if General "lectric would adopt Containment
$ystems Branch technical position 6 4. 'Contairment purging During Normal plant Operations. as a design basis for their purge system. Tht'. position was also made a condition of the preliminary Design Approval issued to the Ger.aral (1ectric Company.
the General (lectric Ctepany has since modified their application to incorporate the branch fosition as a design basis for the Gl$$AR.238 Nuclear island design. We now consider this condition resolved. We will review General (lectric's renner of imple.
menting the branch position and report on our conclusions at the final delige stage of review.
6.3 (nergency Core tooling System 6.3.1 $1ss, tem Descript10,n.
In the $4fett ivaluation Report we stated that we would continue to review the follow.
ing areas related to the emergency core cooling systml (1) Overall role of manual actions required to mitigate the consequences of a loss-of. coolant accident.
(2) Results of the spray distribution tests for the Gl$$AR.238 Nuclear 1sland reactor.
(3) The final proprietary version of the topical report on the 8 x 8 rirconium spray cooling test.
With respect to the first item. General (lectric had already specified the general functions required of the reactor opeti ,or. This was found to be an acceptable basis for the issuance of the Preliminary Design Approval to the General (lectric Company and is an acceptable tests for the issuance of a construction pemit to a referencing plant. We will review the 11st of specific actions required of the operator when the design is finaltred and will report on our conclusion dutIng the review of the final design.
With rpspect to itm two, General Electric has provided us with the results of spray distributica tests for an earlier reactor design. However, the present design has a slightly different configuration. General (lectric has agreed to conduct confinna.
tory spray distritstion tests with the new configuration to verify the predicted distribution. $1nce we consider these tests to te confirmatory in nature, we will not require that the tests be completed prior to the issuance of a construction 66
i We will resort on the results of the tests during the permit to a referencing plant.
review of the final design. l With respect in item three, the General Electric Company hat provided the final 4 propri)Qry version of the topical report on the 8 a 8 firconium spray cooling test.
Topical Report NIDE 20*31. *[mergency Core Cooling Tests of an Internally Prettu We found the topical report accept-Zirc61oy Clad. 8 m 8 $imulated CWR fuel Bundle.*
able, and our evaluatinn of the report has t>een incorporated into the acceptance version of the report.
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7.0 INSTRL*1JNTAT10NANDCONTRO11, 7.1 Lnt roduc tion in the $afety (valuation Report. we concluded that the design bases and criteria for the instrumentation and control systems were acceptable for the Preliminary Design Approval but the review of the preliminary designs had to await the sutoission of additional inforrution by the General Electric Ccmpany.
The review of the preliminary design of the instrumentation and control systems has been conducted on the Gl$$AR 251 Nuclear $ team Supply $ystem doclet (Dociet Number $fN
$0631). However. with Amendment 41 to the GC$$AR 238 Nuclear Island Safety Analysis Report, the General Electric Company updated the GE$$AR 238 Nuclear Island Safety Analysis Report to the same detail as found in the Gts$AR 261 Nuclear Steam Supply
$ystem Safety Analysis Report. The results of our review efforts are sumarized in the following sections, 7.2 kactorTrip$ystem In the Safety Evaluation Report, we concluded that the design bases and criteria of tha reactor trip system were acceptable for the Preliminary Design Approval but the review of the preliminary design coJld not proceed until the General Electric Company supplied us with additional infonnation. Therefore, the discussion in the safety (valuation Report was related mainly to the functional design. In addition, we indenti-fled two design changes which the General Electric Company had proposed in Amendment 24 to the GE$$AR 238 Nuclear Island Safety Analysis Report which appeared to result in undesirable design features.
The General Electric Company has since sutnitted the preliminary design of the reactor trip system in Amendmerit 41 to the GES$AR 238 Nuclear Island safety Analysis Report which incorporates the design cases and ~ addresses the staff concerns discussed in the
$afety tvaluation Report. The following discussion details the staff review of the preliminary design of the reactor trip system and how the design bases and criteria have been implemented by the prelim; nary design.
The preliminary design of the reactor trip system (consisting of the system elementary wiring diagrams) was reviewed to detennine conformance to the design criteria approved by the staff during the initial GE$$AR 238 Nuclear Island review. In addition, the preliminary design was analyzed for operational capabilities with regard to the safet; design bases. Both the equipment and the logic implementation are vastly different from previous boiling water reactor designs. Figures 7 1 and 7-2 show these design differences.
71
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The system utillies solid state circuitry which is divided into four separate and l independent divi 6 tonal logic channels. lie four divisional and independent divi +
stonal logics (located in the four divisional output cabinets) receive inputs froin each di the four instrunentation channels which nonitor each critical plant parareter.
Both analog and digital instruaentation channels are used, in general. an analog instrunentation channel includes the sensor / transmitter, signal conditioning circuitry.
and the trip units and logic isolation devices located in the divisional input cabinett. ]
A digital instrunentation channel includes the sensor /switr.h and the lignal condition. !
ing circuitry and logic isolation devices located in the divisional input (abinets, the following parameter 6 are conttored to provide inputs to the reactor trip systeel (flAMJ,J{R 1YP( IN5TRLN(N1- l (1) Scram discharge volume Analog. !
high water level.
(2) Reactor vesse) Analog.
water level (10w and high).
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, (3) Main steam line Analog.
high radiation. l (4) Drywell high pressure. Analog.
1 (5) Nuclear system high pressure. Analog.
(6) Turbine control valve Digital.
fast closure.
(7) Main steamline isolation Digital.
(closureofmainsteamline isolationvalves).
(8) Turbine stop valve closure. Digital.
(9) Neutron monitoring system inputs Upscale and inoperathe trips (startuprangemonitoring, via neutron monitoring system.
average power range monitoring and intermediate range monitoring).
74
- . _ . . . -. ~__ _ _ . _ . . . . _ . _ , . . . _ . . _ _ - . , , . . _ _ . _ . . . . , - . _ , _ - _ _ -
The Outputs of the four divisional ir$u+ cabinets are transmitted to the four divisional logics.+directly to the late division and through isolation devices to the other three divisions, jn this manner, adversa division interaction is prevented. The suitability of these devices is contingent on successful demonstrations by tests of their effective.
nesse as discussed in $sciton 3.17 of this supplement. Each of the four divisional logics (located in the division output cabinets) performs a two.out-of four clincident check of the logic level signals from each set of four instrument inputs, in addition, certain interlots and e.anual functions are perforned in the divisional logics.
The outputs of each of the four divisional logics provide an input directly to its associated scram actuator logic and through an isolator to each of the other three divisional scram actuator logics. Thos. divisional separation is again Nintained through inulation devices.
In each of the four scram actuator logics, the four input signals are combined into two one.out of two logic configurations. The division 1 or 4 logic configuration produces a scram actuator logic trip (de.energises a load driver) for the "A" solenoid c0115 of rod groups 1 cnd 4 and f or the "B" solenoid coils of rod groups ? and 3. The division ? or 3 logic configuration produces a similar trip for the "A" solenoid coils or rod groups 2 and 3 and for the 'B" solenoid coils of rod groups 1 and 4.
The-de-energiring of both the "A" and "D" solenoid coils of each rod produces a scram therefore, the overall scram logic combining the divisional logic outputs to produce a rcactor scram is one out of.two talen twice.
The power for the 'A" solenoid coils of each of the four rod groups is supplied respectively from one of the four ihdependent Class II reactor trip system power supply busses. The power for the *B" solenoid coils of all four rod groups is sup.
plied from one comion non. Class It bet. This power supply configuration provides the necessary redundancy for the reactor trip system safety functions and in addition provides protection against inadvertent scrams on loss of any one power supply.
1he manual scram action for the new design includes four manuel scram switches which are arranged in a one.out-of two taken twice logic. By actuating one switch of r
either the 'A" channel (division 1) or the "D" channel (division 4) and one switch l of either the 0" channel (division 2) or the "C' channel (division 3). the unual '
scram is initiated through the load drivers. Additional a nual scram capability is provided through the rode switch which provides a stram when placed in the shutdown ,
mode.
Another portion of the design for manual scram utillies the same four manual scram
' switches, but each switch is used to open the m in circuit brealers which feed the -!
load drivers and solenoid coils. By actuating any one switch (division 1. 7. 3 or
- 4) the associated group of rods (group 1. 2. 3. or 4) can be inserted. Therefore, this melbod of unual scram relies on a minimum of equipment comon to the automatic scram action.-
75 L
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4 The testing and maintenance capabilities for the reactor trip system have been igraved in the sense that now instrurentation channels can be conpletely bypassed (removed from service if necessary) one at a time and the logic will become two.out.
of three and still seet the single f ailure criterion during the bypassed line inter.
val. Operationally, this is an improvement in the sense that a failure (spurious signal) during the bypassed interval will not cause an unnecessary scram.
The results of our review of specific areas of the reactor trip system are presented below:
(1) The bypass circuitry provided in the preliminary designs includes provisions for indication of the bypassed condition and provisions for preventing simultaneous bypass of more than one instrument channel. Ibe circuitry provisions include a channel bypass awitch with a 6ey lock (one for each of the four channels) which when placed in the bypass position sends a signal (through an isolator) to each of the other channels to prevent bypass of other channels and provides indication of the channel bypass condition.
The bypass circulty is acceptable under the condition that the isolation devices utilised are acceptable an outilned in Section 3.12 of this supplenent.
(2) The testing provisions for the reactor trip system consist of six overlapping tests that can check the operation of the reactor trip system through the solenoid coils during reactor operaticn. These include manual scram, simulated input calibration tests. singic rod scram test. analog instrupent channel test.
sensor check for digital channels and a system pulse test. All these tests.
except the pulse test, are $1milar to tests used on previous designs. The pulle test prov'stons are utilised for all the solid state safety systems which include the General Electric Company reactor trip system and engineered safety teatures systems.
The system introduces a short duration pulse into the solid state logic at the instrunent input cabinets for the purpose of checking the logic circuitry. To preclude an inadvertent scram during testing, the test equipnent will deliver a pulse long enough to propagate through the logic but not long enough to actuate the equipment. The pulse output is monitored at the output of the signal con =
ditioning equipment (analog comparator unit or bistable trip unit) in the input cabinets to the output of the load driver. The load drivers can then be tested separately to verify their ability to energize and/or de.energite the respective connected loads. -
We have concluded that the General Ciectric Campsny has defined the necessary design criteria (which include the institute of (1cctrical and flectronics Engineers Standards 2791971. " Criteria for protection Systems for Nuclear power Generating Stations." 3791972. " Guide for the Application of the Single f ailure Criterion to huclear power Generating Stations protection Systems." and 338-1971 7-6 j
l l
' Criteria for the Feriodic festing of Nuclear p(wer Generating Stations Protection systems') and has provided suf ficient preliminary design information for the testing '
of the solid State safety system. We find this acceptable for the Preliminary Design Approval.
During the final design review we will review the pulse test portion of the final design, including the qualification of the isolation devices used in this ,
system (see Section 3.12 of this supplement). in order to determine that this new design feature will not degrade the overall safety function of the solid state safety system.
(3) We have reviewed the distribution scheme of the 120 Volt. alternating current power for the scram solenoid valve coils. During the review, we noted a possible concern regarding a comon failure node of the load drivers and their ability to de.energile when required. The General Electric Company has addressed this We concernwithadescriptionO(theisolationcapabilityoftheloaddrivers.
believe that this is an acceptable approach. The isolation (,apability of the load drivers will be a part of the qualification program required under $cction 3.12 of this supplecent.
(4) The General (lectric Company has esde a change in the areas of power supply voltage for the load drivers. Previously the logic voltage for the load drivers in a divisional output cabinet was supplied by the same divisional power supply.
The General Electric Company has changed this such that there is interconnection for this pcwer between division cabinets. (Division 1 supplied a portion of the d.c logic voltage for division 4 and vice versal division 2 supplied a portion of the d c voltage for division 3 and vice wrsa.)
We required the General Llectric Company to , justify this change. The General Electric Company has docupented that, by this design modification, an operational advantage is achieved while maintaining the channel independence. The operational advantage is that failure of the power supply voltage of one logic division will not cause an unnecessary scram of one group of rods as would have previously occurred. The channel independence is maintained by confomance to Regulatory Guide 1.75. ' Physical ?ndependence of Electrical Systems," recommendations in this area of the modified de'ign and therefore ($ acceptable.
(5) During our review of the reactor trip system, we noted that certain inputs to the reactor trip system (!riginated from the turbine building area which is not nomally categorized as a seismic Category I structure. The inputs provide a signal to scram the reactor on a turbine trip even( and include the turbine stop valts closure signal and the turbine control valve fast closure signal (and associatedturbinefirststagepressurepermissives).
77
. - ~ _ _ _ _ ._ . - _ _ -
. _ _ . _ _ . - _ _ _ _ _ _ . ~
Because our polition with regard to reactor trip system inputs has maintained that all the inputs must weet the Institute of [lectrical and Electronics Engineers Standard 2791971 (including the seismic qualification requirerents), the inputs from the turbine building area were unacceptable as proposed.
Therefore, we required the General (lectric Consiany to demonstrate that safety grade bactup trip inputs to the reactor trip system did exist. which could pro +
vide a reactor scram when required in order to prevent unacceptable consequences.
The General [lectric Company has perfonned an analysis to denonstrate that safety grade trips are available to prevent unacceptable consequences. We have teviewed the analysis and the results ar: presented in Section 15.5 of this supplerent.
i Based on our conclusion on the General Electric Company's analysis of the f ailure of these turbine trip inputs and the General Electric Company's comit-pent that these inputs satisfy all ttte requirements of the Institute of [lectri-cal and [lectronics Engineers Standard 279 1971, with the exception of seismic qualification, we conclude that these inputs to the reactor trip system from the turbine butiding area are acceptable.
(6) The General Clectric Company has stated that the reactor vessel high water level input to the reactor trip system will be designed in a manner identical to other safety grade inputs to the reactor trip system. We find this comitnent accept-able but will require formai docurentation of the high water level trip prior to the issuance of a construction permit to a referencing plant. We will report on this docuuntation in a future supplement to the Safety (valuation Report.
We have concluded that, except for the previously nentioned item 6, the prelitninary design of the reactor trip system satisfies the Commission's regulations and is acceptable for the Preliminary Design Approval.
7.2.1 Fast Scram (New Section)
The General [lectric Company has removed all reference to the prompt relief trip system and has proposed to utiltre a "f ast scram" system. The General Electric Company has stated that the
- fast scram" system does not affect any of the instrunenta-tion and control sy.tems in the plant except for the control rod drive system's hydraulic lines.
However, we noted in Amendment 41 to the Gt55AR-238 Nuclear Island Safety Analysis Report that the scram solenoid arrangement for each Control rod drive has changed from two individual solenoid valves to a single dual coil solenoid valve.
78
We conclude that this change does not alter our conclusions with regard to the plant instrumentation and control systems since there are no electrical functional changes from previous plant designs. as power must be removed from both coils to scram as in pre'ious designs. We find that the instrutentation and control systems aspects of the 'f ast scram" system are acceptable for the Preliminary Design Approval.
1.3 {n11peered SafeMatures Systems 7.3.1 Introduction in the Safety tvaluation Report, we concluded that the design bases and criteria for the engineered safety features systems were acceptable for the Preliminary Design Approval, with the exception of those specified for the standby gas treatment system.
We could not make any conclusion 45 to the preliminary design of the engineered safety features systens at that time, for the designs had not been tutnitted for review, the General llectric Company has since Sutnitted the preliminary design of these systems in Amendment 41 to the GC$$AR 238 Nuclear Island Safety Analysis Report.
The preliminary design of the engineered safety features systems (consisting of the elementary wiring diagrams) was reviewed to determine conformance to the design criteria established during the initial GE$$AR.238 Nuclear Island review, in addition, the preliminary designs were analyzed for operational capabilities with regard to the system's safety design bases.
Although the functional performance requirements of the engineered safety features systems are fundamentally the same as previous boiling water reactor plant designs, the instrumentation and control systems (that 15. the sensors logic and actuators) for the engineered safety features systems are not similar to any previous boiling water reactor designs. The engineered safety features instrunentation and control systems utillie solid state equipment and, coupled with the reactor trip system, form what the General [lectric Company refers to as the solid State safety system. The testing for the engineered safety features includes the same pulse test provisions as the reactor trip system described in Section 7.2. item 2 of this supplement.
As in the case of the reactor trip system, the preliminary designs were reviewed based on the criteria established during the review of the conceptual designs. The results of our review of the preliminary designs are sumnartted in the following sections.
1.3.2 Eneroenry Core Cooling System F.3.2.1 !!,!ah pressure Core Spray in the Safety Evaluation Report on the Gt55AR.238 Nuclear Island design, we concluded that the design bases and criteria for the high pressure core spray system were 79
acceptable for the Preliminary Design Approval but the review of the preliminary design could not be completed without additional information from the General llectric Company.
The General (1cctric Company has since suteltted the prelimiaary design for the system in Arendment 41 to the GL$$AR 238 huclear Island $4fety Analysis Report. We j have reviewed this information and conclude that all aspects of the design bases and criteria, which have been previously approved, have been appropriately incorporated into the preliminary design. We conclude that the instrueentation and controls for the high pressure core spray are acceptable for the Preliminary Design Approval.
7.3.2.? Automa,t,1L D efressuri t et,t_o LS 1st m in the $afety Evaluation Report, we identified two generic issues of long standing which w& required the General (lectric Company to address. These were the capability for on line testing of the solenoid valves and the consequences of inadvertent actuation of the automatic depressuritation system resulting from a single failure, The General Electric Company has included the single failure design criterion as part of the design bases for the inadvertent actuation of the automatic depressurtration system. The preliminary design of the actuation system for the automatic depressurita-tion system includes two logic trains within each division. A coincidence of the two logic trains of a division must be satisfied to achieve actuation, tach logic train is identical except that one train of each division contains a delay timer to delay automatic depressuritation system actuation for I?O seconds af ter the loss of coolant accident signals sre present.
We identified the 120 second tiner as a potential single equipment failure that could cause inadvertent actuation during the first I?O seconds of a loss of coolant accident.
The General Electric Company has verified that this inadvertent actuation (during the I?0 second tise interval) has acceptable consequences and concludes that multiple failures must occur to cause inadvertent actuation of the automatic depressuritation system during plant operation. We find this acceptable.
In Amendment r* to the Gt$$AR ?38 Nuclear Island Safety Analysis Re,nort, the General Electric Company comitted to perfoming a study of methods to inyrove the testability of the automatic depressurtration system. We found this commitrent acceptable for the issuancw of the Preliminary Design Approval to the General Electric Company, subject to our revtew and approval of the preliminary design for the eutomatic depressurisation system. including the design provistons for testing.
In conjunction with this program, the General Electric Company perforned reliability analyses for various alternate conceptual designs for the automatic depressurl2ation system which included on line testing capability, Based on these analyses, the General Electric Company concluded that the esisting system design is more reliabic than any of t!* alternate designs that were evaluated.
7 10
_ _~ _
l We are reviewing this information and are not able to concur with the General Electric Company's conclusion because of concerns with sone of the assumptions used in the analyses, f or esemple, we do not believe that the General Electric Company assumption for the solenoid valve failure rate has been adequately justified.
We are currently pursuing this aspect of the automatic depressurization system design with the General flectric Company and will report on our conclusions in a future supplenent to the $afety (valuation Report.
7.3.2.3 QwAj}ufe. Core _ $prjy_ and lowdressure_ Coolant injectton System in the $atety ivaluation Report, we concluded that the design bases and criteria for the low prellure core spray and low pressure coolant injection systems were acceptable for the Preliminary Design Approval, but the review of the preliminary designs of these systems required additional information from the General [lectric Company. Part of the bests for the acceptance of the design bases and criteria was the General (lectric Company convnitnent. in Anendment 24 to the GCS$AR 238 Nuclear Island Safety Analysis R(port, to modify the designs to provide diverse initiation signals that were not dependent on a non diverse interlock and to evaluate the designs to assure that adequate protection against high pressure is provided for the low pressure portions of these systems.
The design of the emergency core cooling system injection lines was reviewed to confirm that the isolation provisions at the interface with the reactor coolant system were adequate. The number and type of valves used to fonn the interface between low pres. s r
sure portions of the emergency core cooling system and the reactor coolant system must provide adequate assurance that the energency core cooling system will not be subjected to a pressure greater than its design pressure. This may be accomplished by any of the follcwing provisionsi
)
(1) One or more check valves in series with a normally closed notor. operated valve.
The motor operated valve is to be opened upon receipt of a safety injection signal once the reactor coolant pressure has decreased below the energency core cooling system design pressure.
(2) Three check valves in series.
(3) Two check valves in series, providet, that there are design provisions to permit periodic testing of the check valves for leaktightness and that testing is per.
formed at least annually.
The General Electric Company has utilized a check valve in series with a normally closed motor-operated valve. The motor operated valve is to te opened upon receipt of an accident signal once the reactor coolant pressure has decreased below the design 7 11 r -
i l
l design pressure of the low pressure core spray system and the low pressure coolant injectionsystem. The General Electric Company has also agreed to provide a pressure ;
interlock on the motor + operated valve to preclude the opening of the valve untti the ;
i reactor system pressure is sufficiently low. The General L)actric Coetany has agreed to ma6e the interlock diverse in order that the diverse accident initiation signals i (10w reactor water level and high drywell pressure) are not dependent on a non diverse interlock. However, the General (1ectric Company has not supplied the preliminary design informati6n to demonstrate diversity. We will report on this matter in a ,
future supplement to the $sfety Ivaluation Report prior to the issuance of a construc. {
Oton permit to a referencing plant. '
7.3.3 Containment and Reactor vessel isolation Control Syysta_m in the $4faty [valuttion Report, we concluded that the design bases and criteria for the containment and reactor vessel isolation control system were acceptable for the Preliminary Design Approval, subject to the staf f review of the preliminary design. We i
also agreed witt the General Electric Company that the review of the control arrangement for the main sisam isolation valves would be conducted through the review of the l General Electrit Company Topical Report APID 6790, " Design and Performance of General tiectric Boiling Water Reactor Main Steam Isolation Valves.*
The appitcable nief f positions resulting from the review of that topical report are as .
follows:
(1) The solenoid valves rast be qualif ted to the requirements of the Institute of flectrical and tiectronfts Engineers Standard 3231974.
- Qualifying Class I flectrical Equipment for Nuclear Power Generating Stations." for the specific ,
plant in which they will be used. 7 (P) The solenoid valves must be physically separated and electrically isolated in order to preserve the electrical independence of the initiating logic and satisfy f the requirements of the Institute of Electrical and [lectronics (ngineers $tandard l 279 1971 " Criteria for Protection Systems for Nuclear Power Generating $tations.*
(3) A method of testing the valves for proper response time which satisfies the !
rwirementsofGeneralDesignCriterton21(asclarifiedbyRegulatoryGuide i.22. " Periodic Testing of Protection System Actuation Functions") must be provided. . Testing shall be conducted on a routine basis during reactor operation.
With' regard to item 1 on solenoid valve qualification, the General Electric Company has submitted their qualification program which will include this equipment. A dis.
cussion of the review of this qualification program is included in Sections 3.10 and 3.11 of this supplement.-
I
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7 12
With regard to item 2 on separation and electrical isolation, physical separation between solenoids on the same rein steam isolation valve is not required to maintain the independente of the initiating logic circuitry since these circuits are 6ept independentthroughisolationdevices(loaddrivers).
We find this an acceptable approach ard therefore it is concluded that the isolation capability of the load drivers has to be established as part of the qualification program required under Section 3.12 of this suppleeent.
With regard to itt'm 3 on the response time testing of the valves during reactor operation. the G15$AR.238 Huclear Island design incorporates provisions for each main steam isolation valve to tie response. time tested during reactor operation.
s Other portions of the containment and reactor vessel isolation control system will function similar to previous designs, t>ut the eQuilment (sensors, attuation and logic) are not similar to previous plants and were included in the review of the pre.
liniinary design of the overall solid state safety system.
We conclude that the preliminary design of the containment and reactor vessel isola.
tion control system is acceptable for the preliminary Design Approval.
7.3.4 ($sential_ Service WatE $ystem the preliminary design review of the essential service water system has noi as yet been ctripleted. The General [lectric Company provided preliminary design information on this system in Amendment 4) to the GE$$AR.238 Nuclear Island Safety Analysis Report.
We have reviewed this information and have issued requests for additional information to the General [lectric Company. When cleplete responses to these requests for additional information are provided, we will complete our review and report our con.
clusions regarding this system in a future supplement to the Safety Evaluation Report.
7.3.6 Finmat ility Control System We have reviewed the preliminary design information which the General (lectric Company has supplied on the flammability control system through Amendment 41 to the G($$AR-238 Nuclear Island Safety Analysis Report and have issued requests for additional infonnation to the General Electric Company. When complete responses to these requests for additional infonnation are provided, we will complete our review and report our conclusions regarding this system in a future supplement to the Safety (valuation Report.
7 13
i 7.3.6 itandbyGatTreatment$ystem We have reviewed the information supplied to date by the General flectric Company and have issued requests for additional information on the design criteria of the ,
systm and on certain espects of the preliminary design. When ccnplete responses to I these requests for additional information are provided, we will $omplete our reytew and report our conclusions regarding this system in a future supplerent to the $4fety (valuation Report.
7.3.7 }uppressionpoolMa6eupSygem, I l
The General Electric Company has not as yet provided us with that portion of I the logic and control for the suppression pool makeup system describi- the specific l source of the loss.of. coolant accident or emergency core cooling System operating ,
signals. Informetion demonstrating channel independence for the logics and controls ,
for the suppression pool makeup system is similarly lacking. When this information !
is provided, we will cornplete our review and report our conclusions in a future f
supplement to 8.he Safety tvaluation peport.
l 7.3,8 Containment $rr_sy $ystem i
1he General llectric Company has sutaitted preliminary design information on the ,
containment spray system in Amendment 41 to the GL$$AR.238 Nuc1 car Island Safety Analysis Report. We have reviewed this information and issued requests for additional I infomation to the General Electric Company. Once complete responses to our requests for additional information are provided, we will complete our review and report our i conclusions in a future supplement to the Safety (valuation Report. !
7.3.9 Indication of Bypasses P
1he review of the preliminary designs included a review of the provisions for the indication of bypassed conditions as outlined in Regulatory Guide 1.47 ' Bypassed and inoperable $tatus indication for Nuclear powsr Plant safety Systems." ,
The General [lectric Company has nade provisions for this indication to satisfy the requirements of kegulatory Gylde 1.47. We conclude that this is acceptable for the Preliminary Design Approval, ,
7.4 Safe Shutdown System 7.4.1 Reactor Core Isolation Cooling System. L The preliminary design (elerentary wiring diagram) of the reactor rore isolation f cooling system was reviewed to determine the conformance to the design criteria and
-staff requirements established during our initial review, as reported in the Safety [
tvaluation Report.
l 7 14 l
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In that review, we detemined that the reactor core isolation cooling system would be required to t,e classified as an engineered safety feature tecause, together with the high pressure core spray system, it proviced the protection necessary in the event of a rod drop accident. Therefore, the review of the preliminary design concen-trated on the Conformance to the established design criteria and the sane requirements of other engineered safety features such as the high pressure core spray system.
The preliminary design, as propo$ed by the General (Itctric Company, did not provide a feature to transfer the luction of the reactor core isolation cooling system from the condensate storage tank to the suppression pool.
The General tiectric Company has revised the preliminary design to include the auto-matic transfer feature and has designed it similar to the high pressure core spray system feature. With the above modification, we conclude that the preliminary design of the reactor core isolation cooling system is etceptable for the Preliminary Design Approval.
7.4.2 $tenditj, quid _ Control sy,s,le,m We reviewed the preliminary design (elerentary wiring diagram) of the standby liquid control system to determine the conformance to the design criteria and staf f require.
- ments established during our initial review as reported in the $afety Lyaluation keport.
In accordance with the review, the General llectric Con'pany has submitted the pre.
Ilmintry design which provides for redundant actuation of the standby liquid control system pumps. In addition, they have submitted the results of an analysis which demonstrate that fatture of the reactor water cleanup system to isolate will not prevent the standby 11guld control system from performing its function.
We conclude that the preliminary design of the standby liquid control system conforms to the design bases approved during our initial review and is therefore acceptable for the preliminary Design Approval.
7.5 }efetydeletedOtsplayInstrumentation 7.5.1 f.tniral, in the $afety [ val.iation Report, we concluded that the agreements made with and comitments made by the General Electric Company on the design criteria for safety-related display instrumentation were acceptable fo'r the Preliminary Design Approval.
We also reported that the General (1ectric Company had agreed to provide specific information such as lists of the indications and controls to be provided and the physical arrangements of the control board 0a9015. This information was to be reviewed 7 15
af ter the issuance of the preliminary Design Approval to the Gerieral (lectric Company.
1he Gener31 flectric Company has since provided us with a list of the indication and controls within their scope of supply, the preliminary design and controls within their scope of supply, and the preliminary control board arrangecent drawings. We have reviewed this information and consider it acceptable for the preliminary Design Approval.
The Ct$$AR 738 Nuclear Island design for the control room incorporates both the Nuclenet and power Generation Control Complex design packages which are discussed in the following section.
7.6.2 hyr1cnc,1 We have reviewed the Nuclenet design criteria in the past through individual case reviews and through Topical Report NIDO.10939. "Deshn Criteria and Technical Descrip-tion of plant / Operator Interface of the Nuclenet 1000 Control Complex." in those reviews, we have deterintned that Nuclenet is basically an operator.to plant system interf ace aid which htilites cathode ray tube displays. The General (lectric Ccunpany has submitted the preliminary control board arrangement drawings as part of the G'$$AR.?38 Nuclear toland $afety Analysis Report and has stated their ccepitment to Regulatory Guide 1.75, " physical Indeperdence of [lectrical $ystems.' hased on the General tiectric Company's statement that all instrumentation and control equipment required for plant safety in the area of the main control room are unaf fected by ut11tringNuc1cnet(sinceitremainsherdwiredinNucienet)andthereviewofthe preliminary design information in the Gt5$AR 238 Nuclear Island $afety Analysis Report, we conclude that the Nucienet design is acceptabic for the preliminary Design Approval.
7.b.3 power Generatton Control Comple,(
We have reviewed the power Generation Control Complex concept in Topical Report N!D0-10466, " power Generation Control Comples," and have been participating in on. going discussions in this area with the General [lectric Company. The main subjects involve the provisions for satisfying the recommendations of Regulatory Guide 1.75, ' physical independence of (lectrical Systems," and the acceptability of large quantities and large concentrations of cables in a sinall area immediately beneath the control room floor.
1he General Electric Company will design the power Generation Control Complex to Regulatory Guide 1.75. As a result of our generic study in the area of fire protection criteria, further requirements may be imposed on this system 50 that unacceptable damage will not result from a fire. This area of the control room design is still a subject of on going discussions between the staff and the Genera) Electric Company in o"r evaluation of NIDO 10466, We will provide the results of our review of NCDO 10466 at the final design stage of review.
4 7 16
7,6 Ali Other $ystems liequigred for Safety, 7.6.1 Generg The systems reviewed for this section were (1) Refueling interlocks.
(?) Reactor vessel instrurientation.
(3) Process radiation penitoring system.
(4) Area radiation nonitoring system.
($) Peactor water clean up system.
(6) Leak detection system.
(7) Process computer system.
(8) Containment atmosphere monitoring system.
(9)'Neutronmonitoringsystem, (10) f uel pool cooling and cleanup system.
'he revice of the preliminary designs concentrated on the safety design bases of the e systems and the interaction of these systems with the plant systems required for scfety.
In the ca'e of the process computer systeme this system has no stated safety design basis and, therefore, we have not required a submittal of the preliminary design for our review. The process computer system will be reviewed at the final design stage of review to verify that the process computer system is not relied upon to perform a safety function.
Inr all other systems, we have reviewed the preliminary design tutaitted by the General [lettric Company to determine that the systems are designed in accordance with the safety oesign bases and to verify that the system design will not comoromise the safety systems with which they interact.
The neutron monitoring system supplies inputs to the reactor trip system and receives power from the reactor trip system safety related power supplies buses. The GCSSAR 238 Nuclear Island design will uttltre the reactor trip system for only divisional and associated equipment in accordance with the reconsnendation$ of Regulatory Guide 1.75. *Phy.ical independence of flectrical Systems."
7-17
. . . . . - - -.~ - .. -. . . _ - - _ _ . - - -- . _ - .
t r
We conclude that the preliminary designs of these systems are acceptable for the Preliminary Design Approval.
7.6.2 R,esitor Pressure Relief Instrunentation in the Safety Evaluation Report we stated that the General [lectric Company had com.
~
-mitted to upgrading the reactor pressure relief instrumentation to provide redundancy and independence i 11 to that required for protection systems. This (opinitment formed the basis et ne Preliminary Design Approval issued to the General Electric Company.
The review of the preliminary design concentrated on verifying that the design require.
rents, as stated. were implemented in a satisf actory manner. The preliminary design originally proposed by the General llectric Company included only a single division (energizedtoactuate)forthepressurerelieffunction. We concluded that the system did not meet the single failure criterion and was, therefore, unacceptable.
In a subsequent revision to the preliminary design, the General [lectric Company provided a two division system similar to the automatic depressuritation system function. Now either of the two divisions can actuate all of the valves as required.
As similiarly noted in the automatic depressuritation system evaluation, this type of design allows for the ren. ate possibility that all the safety related valves could be opened due to a spurious division signal, in order to help prevent such a spurious signal, the General Electric Company has included the single failure criterion as part of the design bases for inadvertent actuation of the safety.reifef valves. As in the design of the automatic depressurita.
tion system, the preliminary design of the actuation system for the safety. relief valves includes dual logic trains within each division. A coincidence of the two logic trains (within the sane division) must be satisfied to achieve actuation.
Therefore, multiple failures must occur to cause inadvertent actuation.
$1nce the actuation of the safety relief valves via the pressure relief instrumentation and control system is identical to the actuation of the automatic depressuritation system and since the design criterla of the reactor pressure relief instrumentation will be equivalent to that of protection systems, we will require that the resolution achieved for the testing of the-solenoid valves on the automatic depressuritation system also be applied to the solenoid valves on the reactor pressure relief system.
We will report on the resolution of this item in a future supplement to the Safety Evaluation Report.
l 7 18 r .. , - . , , . , . . , . .,,
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1.6.3 Recircul_ation Pump,frip_$y gem The recirculation pump trip system perfores the function of disconnecting (via redundant circuit brea6ers in series in the motor feeders) the main power to both reactor recirculation pumps when the turbine trips. The equipment of this systee is all safety. grade equipment and is connected to the reactor trip system circuitry (see figure 7 3). When the turbine stop valves close and/or the turbine control valves fast close, indicating a turbine trip. the four recirculation pump trip divisional logics receive signals from the two out of four divisional logics of the reactor trip system, which produce $ a reactor scram fur this event, tech recirculation pump trip divisional logic then energites a load driver which in turn opens the associated circuit breaker. Two circuit breakers are arranged in series in the power feed circuit of each of the two recirculation pump motort and, ther6 fore. e one.
out of two system level actuation in produced.
This actuation logic prevents a single failure from either disabling the safety function when required or inadvertently tripping more than one recirculation pump during operation.
We have concluded that the design criteria and design description provided for the tectreulation pump trip circuitry are acceptable. We will require the General Electric .
Company to submit the preliminary design information (elementary diagrams) for this system prior to the issuance of a construction permit to a referencing plant.
7.7 Control Systres 7.7.1 Rod Control and Information Systems and instrumentation
,(Formerly referred to as Reactor Manual Control System The rod control and information system 15 included in Section 7.7 in " Control $ystem"1 however, portions of the system perform functions related to safety.
The General Electric Company has provided preliminary design information for the rod control and information system which includes functions related to ganged rod motion, rod information monitoring, and rod pattern control.
We have been reviewing the reactor manual control system design for the General Clectric Company's earlier plant designs as part of a generic review of the General Electric Company's reactor control systems. Although the system title has been changed from reactor manual control system to rod control and information system for the GLS$AR 238 Nuclear Island design, these designs are similar.
Therefore, we required the General Electric Company to provide a detailed comparison between these designs in order that the ptevious and on. going generic review efiart can be appropriately applied to the GE$$AR.238 Nuclear Island design.
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- The General Itectric Company has provided information which verifies that areas of the propostd rod control and information system and instrumentation is similar to previous reactor manual control system designs. Both systems use mu1N plesed signals to transponders on the control rod hydraulic tontrol units and both systems use '
similar solid state rod inhibit logic and rod motion timers with automatic fault. finding circuitry. The dif ferences as enumerated by the General flectric Company include new provisions for satisfying the design criteria established during the Gt5$AR.238 Nuclear Island review. Specific dif ferences noted include dual rod position information and redundant , rod action control to satisfy the safety design bases. Other differences include provillons for ganged f od motion and equipment ut111 red for operator control and display functions. Based on the above comparative design information, coupled with the on. going design review of the reactor manual control system of the previous General flectric Company designs and the review of the preliminary dnsign infor1 nation submitted for the rod i control and information systsm. we conclude that this area is acceptablo for the Pre. liminary Design Approval. We will require that the results of the on going generic review be applied to the GL$$AR.238 Nuclear Island when they become available. 7.1.2 (tfler_ControlSyst_emt the systems reviewed for this section of this supplement weret (1) Recirculation flow control system. (2) f eedwater control syster. (3) pressure regulator and turbine generator control system. (4) Gaseous redwaste control system. (6) Liquid radweste control system. We have reviewed the controls for these systems to determine the effects of f ailures or malfunction of the controls on the reactor protection system and other plant safety.related systems. We conclude that failures or es1 functions of the controls would r.ot be expected'to degrade the capabilities of plant safety systems in any sig-nificant degree, or to lead to plant conditions more severe than those for which the safety systems are designed and are, therefore, acceptable. I 7 21 ; I I 7.8 instrumentation Interfaces with Balance.of_ Plant Systees i in the Safety Ivaluation Report, we stated that we could not proceed with the review > of Instrusentation interfaces with balance of. plant systems until the General Electric Company provided the preliminary designs of the instrumentation and control systems within their scope of design. i Mendnent 41 to the CES$AR 238 Nuclear Island safety Analysis Report provided this preliminary design inforination, ar.d we have issued requests for additional infortnation to the (eneral Electric Company on electrical enterfaces. When complete responses to our requests for additional inforination are provided, we will complete our review and report on this area In a future supplement to the Safety (valuation Report. i i 7 22 8.0 Qt(1RICAl POW [R $Y$1[MS In the safety [ valuation Report, we concluded that the proposed design criteria for the standby power instrumentation and control systems form a generally acceptable basis tot developing a design for the electrical power systems. Since the issuance of the Safety [veluation Report. the General Electric Company has submitted additional inforretion on the electrical power system in Anendnent 41 to the GE55AR.238 Nuclear Island Safety Analysis Report. !!c have reviewed this informa. tion and have issued requests for additional infornation to the General Electric Company. Our main areas of concern relate to the load shedding and sequencing logic for the onsite power system. Interfaces betweca the onsite power system and the balance.of. plant, and the exceptions which the General Electric Company is taHng to Topical Report NEDO 10905. 'High Pressure Core Spray $ystem Power Supply Unit." which is referenced by the GC55AR 238 Nuclear Island. We will report on these areas in a-future supplement to the Safety Lyaluation Report. 8 1- i 9.0 AUXillA_RY SYS1tMS 9.3 Process Auxijf eries 9.3.1 th!a59am isolation Valve Jeakane Control System In the Safety tvaluation Report. we concluded that the design of the main steam isolation valve leakage control system was acceptable, subject to two requirements. We required an interlock to preclude the operation of any inboard leakage control . system if the is. board main steam isolation valve associated with that system 6.as not fully closed. In addition, we required that the setpoint of the flow element timers for the inboard leakage control system be set at 11.6 standard cubic feet per hour. Both of these requirements were made a condition to the preliminary Design Approval issued to the General Electric Company. In an April 13. 1976 letter from I. Stuart of the General Electric Company to
- 8. Rusche of the Nuclear Regulatory Commission. the General Electric Company proposed a new positive sealing system to replace the present leakage control system on the main steam lines. As discussed in Section 6.2.3 of this supplement. we have reviewed the General Electric Company proposal and conclude that it is technically feasible using state.cf the-art technology and that ure is reasonable assurance that the General (1ectric Company can develop a final design which we could find acceptable.
The General Electric Company has consnitted to provide us with a topical report to address the remaining areas of concern associated with the review of the mal # steam line positive sealing system. We will report on the results of our review of that report at the final design stage of review and will require that the results of our review be incorporated into the accident analysis calculations perfomed for any plant referencing the GESSAR 238 Nuclear Island design. Based on the infomation and comenttments provided, we conclude that the proposed positive sealing system for the main steam lines is acceptable for the preliminary Design Approval and is therefore acceptable for use by a referencing applicant in a construction permit appilcation. l i 91 4 ; g. , _, , _L,- _ ,a._ . _ , _ . , _ __i _- . .. . .. __., _ s; . e 11.0 RAplDAC11YfWA5ffMANAGfM(Ni 11.3 Gaseous Waste Trea_tnent $1 stem 11.3.3 C_onclusioni in the $afety tvaluation Report we stated that continuous purging of the containment directly to the environment without treatment was unacceptable and purge filtration was made a condition of the Preliminary Design Approval issued to the General (1ectric lontany. In Anendnent 43 to the GES$AR 238 Nuclear Island $afety Analysis Report, the General Electric Company modified the GES$AR 230 Nuclear Island design to incorporate a non. safety grade filtration system connected to the containment purge system. A safety. g' %de flitration system is not required since the containnent purge line is isolated in the event of an accident. We consider this modification acceptable and the condition resolved. 11 1 -- - .~. - .- - .. 15,0 ACCIDENT ANALYSIS 15,2 - Abnormal Operational Transients In Amendrent 41 to the GESSAR-238 Nuclear Island Safety Analysis Report, the transient cie ~es have been levised to incorporate fast scram. The types of typical transients evaluated were losses of flow, increases in pressure ai d power, decreases in coolant temperature, and increates in coolant flow. The most limiting transients in these categories were a two pump trip, a generator load rejection without bypass, a loss of feedwaterheating,andflowcontrollerfailure(increasingflow). Of these, the most limiting core-wide transient was generator load rejection without bypass resulting in a change in the minimum critical power retto of 0.16. Addition of the change in minimum critical power ratio to the safety limit minimum critical power ratio gives the operating limit minimum critical power ratto required to avoid violation of the safety limit. Should this limiting transient occur. T6.as, the operating limit minimun, critical power ratio is 1.23, The transient analyses were evaluated with the scram reactivity insertion rates shown in Figure 15.1.1-1 of the GES$Ak 238 Nuclear Island Safety Analysis Report, The initial condition parrmeters used for the analyses are acceptable. The initial rnini-mum critical power ratio assumed in the transient analyses was equal to the established operating limit minimum critical power ratio of 1.23. The analyses were performed using a computer-simulated analytical modei of a generic Girect-cycle bolling water reactor, as described in Topical Report NEDO-10802. 'Analyt. Ical Methods of Plant Transient Evaluations for the GE BWR," Our review of this topical report will be completed prior to the final design review of the GESSAR-233 huclear Island design, thanges on limitations resulting from our review will be reflected in the final design. The rod withdrawal error (limiting local event) t*ansient is discussed in the CE55AR-238 Nuclear Island Safety Analysis Report in terms of worst case conditions. The discussion indicates that if the peak linear power design limits are exceeded, the nearest local power range monitor subsystems will detect this phenomenon and initiate an alarm. However, if the operator ignores the alarm, the analysis considers the continuous withdrawal of the maximum worth control rod or control rod gang (at enanimum drive speed) to the full out position when operating between 20 percent and 70 percent power. and the withdrawal of.a control rod or rod gang over two feet beyond the original rod pattern when operating at greater than 70 percent power. The results of the analysis show that the minimum critical power ' ratio will remain above 1.07 and the cladding will remain under the one percent plastic strain limit. The rod withdrawal error transient is not limiting for the GE55AR-233 Nuclear Island reactor above 70 15 1 percent power at rated flow since the rod pattern control system prevents the motion of any control rod or control rod gang ta excess of two feet beyond the existing rod pattern. A two pump trip system has been added to the GESSAR 238 Nuclear Island design. This system trips the recirculation pumps upon sensing either a fast closure of the turbine control valves or excess coolant inventory in the reactor vessel. We have reviewed tnese two transients allowing credit for the two punp trip feature. As indicated in Section 7,6.3 of the supplement, we ha've reviewed the markup drawings of the two-pump
- trip feature but require that this design change be formally documented.
In addition to the criterion on fuel cladding integrity, there is a requireL c ttet- ' pressure in the reactor coolant and main steam systems should be maintained t'elow 110 percent of the design pressures during transients. For the GESSAR-238 huclear Island reactor this limit is 1375 pounds per square inch gauge, and it is not exceeded during the transients analyzed, i We conclude that the plant design is acceptable with regard to transients that are expected to occur during the life of the plant (minimum critical power ratio will not exceed . 07 and reactor coolant pressures will not exceed 1375 pounds per square ' inch gauge). 15.4 Antic 1 Lated Transients Without Scram in September 1973, the Atomic fnergy Commission published WASH-1270, " Technical Report on Anticipated Transients Without Scram for Water-Cooled power Reactors," establishing acceptance criteria for anticipated transients without scram. In conformance with the requirements of Section 11.0 of Appendix A to WASH-1270. " Technical Report on Antici-pated Transients Without Scram for Water-Cooled power Reactors," the General Electric Company submitted Topical Report NEDO-20626. " Studies of BWR Designs for Hitigation of Anticipated Transients Without scram." This report is referenced in the GESSAR-238 Nuclear Island Safety Analysis Report. 2 We have completed our review of Topical Report NED0 20626, " Studies of BWR liesigns for Hitigation of Anticipated Transients Without Scram," and published on findings in Deceirber 1975 as the report " Status Report on the General Electric Analyses of Anticipated Transients Without Scr t ," This report lists our concerns with the General Electric Company submittal,' and requires the General Electric Compa_ny to update their anticipated transients without scram analyses. For example, the General Electric Company has not addressed the reliability of systems which they assume function in their_present analyses (such as the relief valves or the high pressure --core spray). . If the General Electric Company cannot demonstrate that the combined utmeliability of these systens is low enough to meet the WASH-1270, " Technical Report on Anticipated Transients Without Scram for Water-Cooled power Reactors," . safety objective, the anticipated transient without scram analyses must be revised assuming these systems fail to operate. , 15-2 . ~ . . . . . i . - In our letter of April 7,1976 (R. E. Heinenan to I. Stuart of General Electric), we , required the' General Electric Company to provide the following information by Jano 30, 1976: (1) The results of additional analyses and the further justification of the General Electric Company analysis model identified in our report, " Status Report on the General Electric Analyses of Anticipated Transients Without Scram." 1 (2) Based on these analyses, identification of the design changes needed to assure that the limits specified in WA5H.1270, " Technical Report on Anticipated Trans- , , lents Without Scram for Water-Cooled Power Reactors," will not be violated following an anticipated transient without scram event. The General CIectric Company has since provided a response to our request by letter dated July 2,1976. We are continaing our generic review of this natter and will require that any changes which may be required be incorporated into the design in a t'mely manner. We conclude that appropriate measures to nitigate the consequences of such events are technically feasible and are within the state-of-the-art. We find this acceptable for the preliminary Design Approval. 15.5 Failure of inputs f rom Turbine Building to Reacior Protection System (New Section) In our review of the reactor protection system we noted that certain inputs to the , reactor protection system did not meet all of the safety design criteria required by the Institute of Electrical and Electronics Engineers Standard 279-1971, " Criteria for protection Systems for Nuclear Power Generating Stations." The specific inputs identified were those originating from th? turbine building. The turbine building in the case of most boiling water reactor plants is not seismically qualified. For these reasons we requested that the General Electric Company investigate the probability and consequences of losing these inputs to the reactor protection system. In their response the General Electric Company stated that the probability of losing these inputs to the reactor protection system was extremely small--on the order of 10 per year. The General Electric Company also investigated the percentage of fut.1 rods in the core which are subject to boiling transition prior to termination of the . ~ transient by the backup scram (high flux). The most limiting case was that for , turbine trip with. a bactup scram in which the turbine bypass system failed to operate. For this case, the General Electric Company indicated that seven percent of the fuel rods in the core would experience boiling transition. We have evaluated the General Electric Cnmpany response and conclude that they have adequately established that this is an extremely low probability event and that they have conswrvatively calculated the effects of the transient, in our evaluation of this event we have used the assumptions of the control rod drop accident to provide a very conservative estimate of tne offsite dose consequences. We have also assumed that those fuel rods which experience bolling 16 ansition ' perforate. The quantity and- j 4-
- )
- i behavior of fission products which are released are treated in the same runner as for.
the control rod drop accident. Based on the above, we have concluded that the two hour doses will not exceed 95 rem to the thyroid and 4.5 rem whole body, assuming a relative concentration value of I I x 10"3 seconds per cubic meter. We have also concluded that the doses for the course of the accident will be less than 55 rem to the thyroid and 4.5 rem whole , body, assuming a relative concentration value of a x 10*4 seconds per cubic reter. Based on our review e conclude that the probability of the event is low enough to permit us to use the conservative accident assumptions and compare the offsite doses - to 10 CFR Part 100. We also conclude that the offsite dose consequence', are well within 10 CFR Part 100 limits. 1 i I l t ~ i l I k 15-4 . _ _ . _ _ - _ _ _ _ __ m _ _ 0 APPENDIX A CONTINUATION OF CHRONOLOGY OF RADIOLOGICAL REVIEW December 19. 1975 The General Electric Company filed Amendment 40 which addressed containment dynamic loads. December 22. 1976 The Nuclear Regulatory Commission issued Preliminary Design Approval to the General Electric Company for their GESSAR-238 Nuclear Island design. february 27. 1976 The General Electric Company filed Amendment 41 which updated their Nuclear Island application electrical information to that of their 'luclear Steam Supply System application. April 30,1976 The General Electric Company filed Amendeent 42. June 10-11. 1976 We met with the General Electric Company represen-tatives to discuss the remaining open items on the Nuclear Island application. June ll, 1976 The General Electric Company appealed our position on the *.astability of the automatic depressuriza-tion system to Comission management. June la 1976 The General Electric Company filed Amendment 43 which resolved three of the four conditions on the Preliminary Design Approval. July 1. 1976 We met with General Electric Company representatives to discuss the remaining open items on the Nuclear Island application. July 19. 1976 The minutes of the June 10-11. 1976 meeting with the General Electric Company were issued. July 23.1976 The minutes of the July 1.1976 meeting with the - General Electric Company were issued. A-1 ... -. . . . . . - . .. .- . _ - . - . . . . - . - . . . - _ . . - . . . - - . ~ _ . . . . ( 1 August 9. 1976 The r.eneral Electric Company requested a formal appeal'on the testability of the automatic depressurization systevr. . August 15, 1976 The General Electric Company' filed Amendment 44, p 1 i 'I r l + A-2 j _ _ . - _ _ . . . . , . . _ . , , .- _.