ML20234C645

Safety Evaluation Report Related to the Final Design of the Standard Nuclear Steam Supply Reference System.Cessar System 80.Docket No. 50-470.(Combustion Engineering,Incorporated)

ML20234C645
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Site:	05000470
Issue date:	12/31/1987
From:	Office of Nuclear Reactor Regulation
To:
References
NUREG-0852, NUREG-0852-S03, NUREG-852, NUREG-852-S3, NUDOCS 8801060284
Download: ML20234C645 (78)

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- Supplement No. 3  ;

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Safety EvaSuation Report  !

related to the final d.esign of the i Standard Nuclear Stearn Supply Reference System CESSAR System 80 Docket No. STN 50-470 Combustion Engineering, incorporated u

r U.S. Nuclear Regulatory Commission i

Office of Nuclear Reactor Regulation December 1987 i

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I NOTICE Availability of Reference Materials Cited in N'R'C Publications o

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Most documents cited in NRC publications will be available from one of the following sources: j

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NUREG-0852 Supplement No. 3 -

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Safety Evaluation Report related to the final design of the Standard Nuclear Steam Supply Reference System ,

CESSAR System 80 Docket No. STN 50-470 Combustion Engineering, incorporated U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation l

Decernber 1987

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ABSTRACT Supplement No. 3 to the Safety Evaluation Report for the application filed by Combustion Engineering, Inc., for a Final Design Approval for the Combustion j Engineering Standard Safety Analysis Report (CESSAR).(Docket No. STN 50-470) l I

has been prepared by the Office of Nuclear Reactor Regulation of the Nuclear Regulatory Commission. The purpose of this supplement is to update the Safety Evaluation by providing (1) the evaluation of additional information submitted ,

by the applicant since Supplement No. 2 to the Safety Evaluation Report was issued and (2) the evaluation of the matters the staff had under review when Supplement No. 2 to the Safety Evaluation Report was issued.

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,9 CESSAR SSER 3 iii E---

CONTENTS Pagg ABSTRACT....................... ........... .......................... iii i

1 INTRODUCTION AND GENERAL DISCUSSI0h................. ............. 1-1 i

1.1 Introduction.................. ............................. 1-1 1.8 Confirmatory Issues......................................... 1-1 1.10 Interface Information.......... ............................ 1-2 1.11 Other Issues................................................ 1-2 3 DESIGN CRITERIA--STRUCTURES COMPONENTS, EQUIPMENT AND SYSTEMS..... 3-1 1

3.6 Protection Against Dynamic Effects Associated With the Postulated Rupture of Piping-............................... 3-1 3.6.2 Determination of Break Locations and Dynamic Effects Associated With the Postulated Rupture i of Piping............. ..... ...................... 3-1 3.6.2.1 Introduction. ............................. 3-1 3.6.2.2 Parameters Eva'cated by Staff.............. 3-3 3.6.2.3 Staf f Criteria ir the Evaluation. . . . . . . . . . . 3-3 3.6.2.4 Staff Evaluation and

Conclusion:

........... 3-4 3.9 Mechanical Systems and Components........... ............... 3-5 ,

l 3.9.1 Special Topics for Mechanical Components............ 3-5 '

1 3.9.1.1 Design Transients.......................... 3-5 i 3.9.2 Dynamic Testing and Analysis of Systems, 1 Components, and Equipment........................... 3-5 3.9.5 Reactor Pressure Vessel Internals................... 3-6 3.11 Environmental Qualificatfor of Safety-Related Electrical Equipment...... . ............................... 3-6 4 REACT 0R................... ......... ............................ 4-1 4.2 Fuel fystem Design....... ................................ 4-1 4.3 Nuclear Design.............................................. 4-1 4.3.2 Description......................................... 4-1 CESSAR SSER 3 v

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CONTENTS (Continued)

Page 4.4 Thermal-Hydraulic Design.................................... 4-1 4.4.6 Statistical Combination of Uncertainties............ 4-1 4.4.11 Instrumentation for Detection of Inade Core Cooling..........................quate .............. 4-3

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5 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS...................... 5-1 5.1 Summary Description......................................... 5-1 5.1.1 Schematic Flow Diagram.............................. 5-1 5.1.2 Nuclear Steam Supply System--Balance-of-Pl ant Interface Requi rements. . . . . . . . . . . . . . . . . . . . . . . . 5-1 5.2 Integrity of Reactor Coolant Pressure Boundary.............. 5-1 5.2.2 Overpressure Protection............................. 5-1 5.4 Component and Subsystem Design.............................. 5-2 5.4.2 Steam Generators.................................... 5-2 5.4.3 Shutdown Cooling (Residual Heat Removal) System..... 5-2 3 5.4.3.1 C o n f i rma to ry I s s u e No . 1. . . . . . . . . . . . . . . . . . . 5-2 5.4.3.2 Reactor Coolant Piping..................... 5-6 5.4.3.3 Decay Heat Removal Without Power-0perated Relief Va1ves.............................. 5-6 5.4.5 5.4.6 Pressurizer.........................................

Safety and Relief 5-7 Valves............................ 5-7 i

6 ENGINEERED SAFETY FEATURE 5..................... .................. 6-1 6.2 Containment Systems......................................... 6-1 6.2.1 Containment Functional Design....................... 6-1 6.2.1.2 Mass and Energy Release Analysis for Postulated Secondary System Pipe Ruptures Inside Containment................ 6-1 6.2.2 Containment Spray System............................ 6-1 6.2.4 Containment Isolation System... .................... 6-1 6.3 Emergency Core Cooling System............................... 6-1 6.3.1 System Design....................................... 6-1 6.3.1.3 Interface Requirements..................... 6-1 CESSAR SSER 3 vi

CONTENTS (Continued)

Page 6.3.3 Testing............................................. 6-2 6.3.3.1 Small-Break Analysis....................... 6-2 6.3.3.2 Post-Loss-of-Coolant- Accident long-Term Cooling.......................... 6-2 6.5 Containment Spray as a Fission Product Cleanup System....... 6-3 7 INSTRUMENTATION AND C0NTROLS........................ ............. 7-1 7.1 Introduction................................................ 7-1 l 7.2 Reactor Coolant System...................................... 7-1 1

7.3 Engineered Safety Features Actuation System................. 7-1 7.5 Safety-Related Display Instrumentation...................... 7-2 9 AU X I LI A RY SY ST EMS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9-1 9.1 Fuel Storage Facility....................................... 9 9.1.4 Fuel Handling System................................ 9-1 9.2 Water Systems............................................... 9-1 9.2.3 Demineralized Water Makeup System................... 9-1 9.3 Process Auxiliaries......................................... 9-1 9.3.4 Chemical and Volume Control System.................. 9-1 10 STEAM AND POWER CONVERSION SYSTEM................................. 10-1 10.3 Circulating Water System.................................... 10-1 10.3.1 Secondary Water Chemistry........................... 10-1 14 INITIAL TEST PROGRAM......... ............... .................... 14-1 14.1 Introduction..... .......................................... 14-1 14.2 Specific Information To Be Included in FSAR...... .......... 14-1 1

14.2.1 Cc-nformance of Initial Test Programs With Regulatory Guides and Industry Standards............ 14-1 14.2.2 Initial Fuel Loading and Initial Criticality........ 14-2 14.2.3 Individual Test Programs............................ 14-2 14.3 Impact of Palo Verde 1 Hot Functional Testing on Other CESSAR Chapters. ........................................... 14-2 14.3.1 Reactor Coolant Pump Diffuser and Impeller Repairs.. 14-3 l

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CONTENTS (Continued)

Page 14.3.2 Resistance Temperature Detector Therm Design Revision.......................owell ............... 14-4 14.3.3 Safety Injection Nozzle Thermal Removal..... ................... Sleeve.................... 14-4 14.3.4 Control Element Assembly Guide Removal.............. 14-4 15 ACCIDENT AND TRANSIENT ANALYSIS.................................... 15-1 15.1 Introduction and Analytical Techniques...................... 15-1 15.1.1 CESSAR Consistency Review Changes................... 15-1 15.3 Limiting Accidents........................................... 15-1 15.3.1 Steam Piping Failures Inside and Outside Containment..........................................

15.3.7 Steam Generator Tube 15-1 Rupture........................ 15-2 15.4 Radiological Consequences of Design-Basis Accidents......... 15-4 15.4.5 Steam Generator Tube Rupture........................ 15-5 16 TECHNICAL SPECIFICATIONS.......................................... 16 17 QUALITY ASSURANCE................................................. 17-1 17.3 Quality Assurance Program.......... ........................ 17-1 17.3.1 CESSAR Amendments 8 and 9................ .......... 17-1 17.3.2 CESSAR Inconsistencies................... .......... 17-1 22 TMI-2 REQUIREMENTS................................................ 22-1 22.1 Introduction.................... 22-1 22.2 Evaluation..................................................

........................ .... 22-1 I.D.2 Plant Safety Parameter Display Console..........

II.D.1 22-1 Performance Testing of Boiling Water Reactor and Pressurized Water Relief and Safety Valves.. 22-2 II.F.2 Instrumentation for Detection o Inadequcte Core Cooling....................f ................ 22-3 II.K.2.13 Effect of High-Pressure Injection on Vessel Integrity for Small-Break LOCAs With No Auxiliary Feedwater............................. 22-3 II.K.2.17 Potential for Voiding in the Reactor Coolant.  ;

System During Transients........................ 22-3 II.K.3.30 Revised Small-Break LOCA Methods To Shos Compliance With 10 CFR Part 50, Appendix K...... 22-4 i

CESSAR SSER 3 viii

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CONTENTS (Continued)

APPENDICES A CONTINUATION OF CHRONOLOGY OF.CESSAR REVIEW B BIBLIOGRAPHY D ABBREVIATIONS E PRINCIPAL CONTRIBUTORS J ENVIRONMENTAL QUALIFICATION OF CLASS 1E ELECTRICAL EQUIPMENT i

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,\y 1 INTRODUCTION AND GENERAL DISCUSSION 1.1 Introduction On November 20, 1981, the U.S. Nuclear Regulatory Commission (NRC) staff issued its Safety Evaluation Report (SER) relating to the application filed by Combus-tion Engineering (CE or applicant) for a Final Design Approval (FDA) based on the final Combustion Engineering Stancard Safety Analysis Report (CESSAR).

Supplement No. 1 tc the SER was issued on March 30, 1983, and Supplement No. 2 was issued on September 30, 1983.

In the SER and its two supplements, the staff identified issues for which either further information was required from CE or additional staff effort was neces-sary to complete the review of the application. The purpose of this third sup-plement is to update the SER by providing (1) the evaluation of the additional information submitted by CE since Supalement No. 2 was issued and (2) the evalu-ation of matters that the staff had under review when Supplement No. 2 was issued. This information addressed the resolution of the issues identified in FDA-2; reflected startup experience at Palo Verde Unit 1, the first plant that referenced CESSAR to become operational; and corrected minor discrepancies in CESSAR.

The NRC Licensing Project Manager for CESSAR is Mr. Guy S. Vissing. Mr. Vissing may be contacted by calling (301) 492 8208 or writing: Division of Reactor Projects III, IV, V, and Special Projects, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555.

Each of the following sections of thio supplement is numbered the same as the section of the SER that is bein0 updated, and, unless otherwise noted, the dis- l cussions are supplementary to and not in lieu of the discussions in the SER.

Appendix A to this supplement is a continuation of the chronology of the corre-spondence between the NRC and the applicant that updates the correspondence in ,

j the SER and Supplement Nos. 1 and 2. Appendix B lists the references used during the course of the review.* Appendix D lists abbreviations used in this l report. Appendix.E is a list of principal contributors to the staff review.

Appendix J contains the staff's evaluar, ion of environmental qualification of Class IE electrical equipment (C2NPD-2b5-A, Revision 3, and Amendment 9 to CESSAR Chapter 3). ,

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1. 8 Confirmatory Issues Section 1.8 of Supplement No. 2 to the SER contains a list of issues that had been essentially resolved to the staff'n satisfaction, but for which certain confirmatory information was to be provnded by CE. Since the issuance of Sup-piement No. 2, CE has provided the required information for a number of those issues. Two are listed below, along with the sections of this third supplement wherein the issues are discussed.

• Availability of all material cited is described on the inside front cover of this report.

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(1) statistical combination of uncertainties (4.4.6)

(2) instrumentation for detection of inadequate core cooling (4.4.11)

These issues have now been acceptably resolved.

At this time, Confirmatory Issues No.1 (shutdown cooling system analysis--

5.4.3) and No. 2 (steam generator rupture analysis--15.3.7, 15.4.5) remain un-resolved because of the deficiencies associated with the auxiliary pressurizer spray (APS) system. The resolution of these remaining issues will be addressed 1 in the next supplement to the SER. l 1.10 Interface Information Section B.3.b.1 of the Commission's Policy Statement on Severe Reactor Accidents Regarding Future Designs and Existing Plants (50 Federal Register J 153, August 8, 1985) stated:

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Each of the two reference design applicants with existing FDAs }

[ Final Design Approvals] must request that their FDAs be amended to permit their designs to be referenced in new CP [ construction permit] and OL [ operating license) applications. The request .

i must either (i) include the information needed to satisfy each of the criteria stated in Section B.2 or (ii) provide suitable interfase requirements to ensure that CP and OL applications l referencing the design will satisfy each of the criteria in 1 Section B.2. Requests in either case need not include an evaluation of how the design conforms to the Standard Review Plan (10 CFR 50.34(g)). '

i By letter dated August 30, 1985, CE requested that FDA-2 for CESSAR be amended I to permit its reference in new CP and OL applications. The request provided it,terface requirements to ensure that CP and OL applications referencing the design will satisfy each of the criteria in Section B.2 of the policy. The interface requirements were included in Am ndment 11 of CESSAR.

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The staf f has reviewed the proposed interface requirements and has deterained that they are suitsble to satisfy each of the criteria of Section B.2 of the

} policy and,are, therefore, acceptable.

1.11 Op er Issues After Supplement No. I to the CESSAR SER was issued, two new issues were identified- pipe whip restraints and safety parameter display sy; tem (SPDS).

Supplement No. 2 states that staff approval of those iss:Ms would result in changes to the CESSAR scope, but that the staff review wn incolnplete. Because the staff considered these issues to be neither. outstanding nor confirmatory, the Final Design Approval excluded the features corresponding to these issues from the CESSAR scope, pending their final resolution.

Pipe whip iestraints are discussed in Section 3.6.2 of this supplement, the SPDS in Section 22.2. These issues have now been acceptably resolved.

CESSAR SSER 3 1-2

i Af ter the Final Design Approval was issued, two other issues were identified--

resolution of the anticipated transients without scram (ATWS) issue (10 CFR 50.62) and the steam generator tube vibration issue.

By letter dateo February 27, 1987, CE submitted its proposal for resolving the )

ATWS issue. The staff is currently reviewing this issue. By letter dated April 10, 1987, CE committed to provide its program for resolving the steam generator tube vibration issue. ByletterdatedJulyJ3,1987,CEprovidedits program for resolving the steam generator vibration problem. The staff is cur-rently reviewing this issue. The resolution of these two issues will be ad-dressed in the next supplement to the SER.

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3 DESIGN CRITERIA--STRUCTURE, COMPONENTS, EQUIPMENT AND SYSTEMS 3.6 Protection Against Dynamic Effects Associated With the Postulated Rupture of Piping 3.6.2 Deter'minat'on of Break Locations and Dynamic Effects Associated With the Postulated Rupture of Piping 3.6.2.1 Introduction j Supplement No. 2 to the CESSAR SER states that, by letter dated June 14, 1983, Combustion Engineering (CE) had submitted a report entitled " Basis for Design of P'icnt Without Pipe Whip Restraints for RCS Main Loop Piping," and that the staff had not finished reviewing this report. This submittal was made to sup-po-t requests by CESSAR appliccnts for an exemption to General Design Criterion (GDC) 4 of Appendix A to Part 50 of Title 10 of the Code of Federal Regulations (10 CFR 50) in regard to the need for protection against the dynamic effects of postulated pipe breaks. At a meeting with CE on October 26, 1983, the staff.

had a number of comments and questions on the submittal. In response to the staff's concerns, CE submitted a revision to the report by letter dated Decem-ber 23, 1933. On the basis of deterministic fracture mechanics analyses, CE l

contends that postulated double-ended guillotine breaks (DEGBs) of the primary loop reactor coolant piping will not occur in CESSAR facilities and, therefore, need not be considered as a design basis for installing protective devices such as pipe whip restraints and jet impingement shields to guard against the dynamic effects associated with such postulated breaks. No other changes in design requirements are addressed within the scope of the referenced reports. For example, no changes are proposed to the definition of a loss-of-coolant accident or its relationship to the regulations addressing the design requirements for the emergency core cooling system (10 CFR 50.46), containment (GDC 16 and 50),

other engineered safety features, and the conditions for environmental qualifi-cation of equipment (10 CFR 50.49).

The Commission's regulations require that applicants provide protective measures against the dynamic effects of postulated pipe breaks in high-energy fluid system piping. Protective' measures include physical isolation from postulated pipe rupture locations, if feasible, or the installation of pipe whip restraints, jet impingement shields', or compartments. In 1975, concerns arose about the asymmetric loads on pressurized-water-reactor (PWR) vessels and their internals that could .

result from these large postulated breaks at discrete locations in the main primary coolant loop piping. This led to the establishment of Unresolved Safety Issue (USI) A-2, " Asymmetric Blowdown Loads on PWR Primary Systems."

The NRC staff, af ter several review meetings with the Advisory Committee on Re-actor Safeguards and a meeting with the NRC Committee to Review Generic Require-ments, concluded that an exemption from the regulations may be granted as an al- i ternative to the resolution of USI A-2 for 16 facilities owned by 11 licensees in the Westinghouse Owners Group (one of these facilities, Fort Calhoun, has a CESSAR SSER 3 3-1 1

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CE nuclear steam supply system). This position was stated in Generic Letter 84-04, published on February 1, 1984. The generic letter states that the affected licensees must justify an exemption to GDC 4 on a plant-specific basis.

Other PWR applicants or licensees may request similar exemptions from the re-quirements of GDC 4, provided they submit an acceptable technical basis for eliminating the need to postulate pipe breaks.

The acceptance of the technical bases for an exemption was made possible by the development of advanced fracture mechanics technology. These advanced fracture '

mechanics techniques deal with relatively small flaws in piping components (either postulated or real) and examine their behavior under various pipe loads. The objective is to demonstrate by deterministic analyses that the detection of small flaws by either inservice inspection or leakage monitoring systems is en- ) <

sured long before the flaws can grow to critical or unstable sizes, which could '

lead to large-break areas such as the DEGB or its equivalent. The concept under-lying such analyses is referred to as " leak before break." Tnere.is no implica-tion that piping failures cannot occur, but rather that improved knowledge of the failure modes of piping systems and the application of appropriate remedial measures, if indicated, can reduce to insignificant values the probability of catastrophic failure.

Advanced fracture mechanics technology was applied in Westinghouse topical reports (WCAP-9558, Rev. 2; WCAP-9787; Letter Report NS-EPR-2519) submitted to the staff on behalf of the licensees belonging to the USI A-2 Owners Croup. k Although the topical reports were intended to resolve the issue of asymmetric blowdown loads that resulted from a limited number of discrete break locations, the technology in these topical reports demonstrated that the probability of j breaks occurring in the main loop piping in the primary coolant system is suf-ficiently low so that these breaks need not be considered as a design basis for requiring installation of pipe whip restraints or jet impingement shields. The i staff's evaluation of this technology is attached to Generic Letter 84-04 as

! Enclosure 1.

Probabilistic fracture mechanics studies conducted by the Lawrence Livermore National Laboratory (LLNL) on both Westinghouse and CE nuclear steam supply i system primary coolant system piping (LLNL,1984) confirm that the probability of leakage (e.g. , undetected flaw growth through the pipe wall by fatigue) and of a DEGB is very low. The results in the LLNL report (1984) are that the best-estimate leak probabilities for Westinghouse nuclear steam supply system main loop piping range from 1.2 x 10 8 to 1.5 x 10 7 per plant year and the best-estimate DEGB probabilities range from 1 x 10 12 to 7 x 10 12 per plant year.

Similarly, the best estimate leak probabilities for CE nuclear steam supply system main loop piping range from 1 x 10 8 to 3 x 10 8 per plant year, and the best-estimate DEGB probabilities range from 5 x 10 H to 5 x 10 13 per plant-year. These results do not affect core melt probabilities in any significant way.

During the past few years it has also become apparent that the requirement for installation of large, massive pipe whip restraints and jet impingement shields is not necessarily the most cost-effective way to achieve the desired level of safety, as indicated in Enclosure 2 to Generic Letter 84-04. Even for new plants, these devices tend to restrict access for future inservice inspection of piping, or if they are removed and reinstalled for inspection, there is a potential CESSAR SSER 3 3-2

risk of damaging the piping and other safety-related components in this proc-ess. If installed in operating plants, high occupational radiation' exposure (ORE) would be incurred, and reduction of public risk would be very low. Re-moval and reinstallation for inservice inspection also entail significant ORE over the life of a plant.

3.6.2.2 Parameters Evaluated by the Staff The primary coolant system of CESSAR facilities as described in the CE letter of December 23, 1983, has two main loops, each consisting of a 42.0-inch-diam-eter hot leg and two 30-inch-diameter crcssover and cold legs. The materials in the primary loop piping are SA 516 Grade 70 (pipes) and SA 508 Class 1, 2, or 3 (safe ends and nozzles). The piping system is clad on the inside surface with stainless steel. In its review of the December 23, 1983, letter, the staff evaluated the CE analyses with regard to 1

) (1) the location of maximum stresses in the piping, associated with the l combined loads from normal operation and the safe shutdown earthquake (SSE) l (2) potential cracLing mechanisms (3) the size of throughwall cracks that would leak a detectable amount under normal loads and pressure (4) the stability of a leakage-size crack under normal plus-SSE loads and the expected margin in terms of load (5) the margin based on crack size (6) the fracture toughness properties of carbon steel piping and weld material 3.6.2.3 Staff Criteria Used in the Evaluation The staff's criteria for evaluating the above parameters are given in Enclosure 1 to Generic Letter 84-04, Section 4.1, "NRC Evaluation Criteria,"

and are:

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(1) The loading conditions should include the static forces and moments (pres-sure, deadweight, and thermal expansion) resulting from normal operation and the forces and moments associated with the SSE. These forces and mo-l ments should be located where the highest stresses, coincident with the l poorest material properties, are induced for base materials, weldments, and safe ends.

(2) For the piping run/ systems under evaluation, all information that demon-strates that degradation or failure of the piping resulting from stress corrosion cracking, fatigue, or waterhammer is not likely should be pro-vided. Relevant operating history should be cited, such as systems opera- o tional procedures; systems or component modification; water chemistry param-l eters, limits, and controls; resistance of material to various forms of stress corrosion, and performance under cyclic loadings.

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i (3) A throughwall crack should be postulated at the most highly stressed loca- l tions determined from item 1 above. The size of the crack should be large enough so that the leakage will be detected with adequate margin, using the minimum installed leak detection capability when the pipe is subjected to normal operational loads.

(4) It should be demonstrated that the postulated leakage crack is stable under normal plus-SSE loads for long periods of time; that is, crack growth, if any, is minimal during an earthquake. The margin, in terms of applied loads, should be determined by a crack stability analysis, that is, that the leakage-size crack will not experience unstable crack growth even if larger loads (larger then design loads) are applied. This analysis should i demonstrate that crack growth is stable and the final crack size is lim-ited so that a double-ended pipe break will not occur. f i

(5) The crack size margin should be determined by comparing the leakage-size crack to critical-size cracks. Under normal plus-SSE loads, it should be demonstrated that'there is adequate margin between the leakage-size crack and the critical-size crack to account for the uncertainties inherent in the analyses and for leakage detection capability. A limit-load analysis may suffice for this purpose; however, an elastic plastic fracture mechanics (tearing instability) analysis is preferable.

(6) The data on materials should include types and specifications of materials used for base metal, weldments, and safe ends; the properties of materials including the J-R curve used in the analyses; and long-term effects, such as thermal aging, and other limitations to valid data (e.g., J maximum and maximum crack growth).

3.6.2.4 Staff Evaluation and Conclusions On the basis of its evaluation of the analysis in the CE submittal of Decem-ber 23, 1983, the staff finds that CE has presented an acceptable technical justification, addressing the above criteria, for not installing protective devices to deal with the dynamic effects of large pipe ruptures in the primary coolant system piping of CESSAR facilities. This finding is based on the l following:

l l (1) The loads associated with the most highly stressed locations in the main loop primary system piping were provided and are within the allowable limits of the American Society cf Mechanical Engineers Boiler and Pres- ,

sure Vessel Code (ASME Code).  !

(2) For CE plants, there is no history of cracking failure in reactor primary l coolant system piping. CE reactor coolant system primary loops have an operating history that demonstrates their inherent stability. This in-  ;

cludes a low susceptibility to cracking failure from the effects of corro- '

sion (e.g. , intergranular stress corrosion cracking), waterhammer, or fa-tigue (low and high cycle). This operating history includes several plants with cany years of operation.

(3) FortheleakratecalculationsperfohmedforCESSAR,initialpostulated throughwall flaws were used that are equivalent in size to that specified l CESSAR SSER 3 3-4 4

in Enclosure 1 to Generic Letter 84-04. CESSAR facilities are expected to have a reactor coolant system pressure boundary leak detection system that is consistent with the guidelines of Regulatory Guide (RG) 1.45 so that it can detect leakage of 1 gpm in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. This will be verified during the case-by-case review of each applicant's submittal. The calculated leak rate through the postulated flaw is large relative to the staff's required sensitivity of plant leak detection systems. The margin is at least a factor of 10 for leakage.

(4) The expscted margin in terms of load for the leakage-size crack under normal plus-SSE loads is greater than a factor of 3 when compared with the limit load. In addition, the staff found a significant margin in terms of loads larger than normal plus-SSE loads.

(5) The margin between the leakage-size crack and the critical-size crack was calculated. Again, the results demonstrated that a crack-size margin of at least a factor of 3 exists.

l l In view of the analytical results in the CE submittal of December 23, 1983, and the staff's evaluation findings discussed above, the staff concludes that the probability or likelihood of large pipe breaks occurring in the primary coolant system loop of a CESSAR facility is sufficiently low so that protective devices associated with postulated pipe breaks in the CESSAR primary coolant system need not be installed.

Applicants or licensees with CE System 80-designed nuclear steam supply system (NSSS) who intend to use the leak-before-break approach to eliminate the need to install protective devices associated with postulated pipe breaks in their primary coolant systems must confirm that their as-built facility design sub-stantially agrees with the design described in the CE submittal of December 23, 1983; specifically, the piping loads should be no greater than those cited in that submittal. Also, applicants or licensees must confirm that their leak detection systems meet the staff's requirements in item 3 above.

3. 9 Mechanical Systems and Components 3.9.1 Special Topics for Mechanical Components 3.9.1.1 Design Transients CESSAR Table 3.9.1-1 summarizes the transients used in the stress analysis of ASME Code, Class 1 components. By letter dated October 22, 1984, CE submitted changes to this table to indicate that the primary system hydrostatic test and leak test design transient cycles are between 120 F and 400 F and not between 100 F and 400 F. These changes were proposed as a result of a consistency re-view of the technical specifications for the first System 80 plant against CESSAR.

The staff determined that these changes do not change the conclusions reached in previous evaluations and are, therefore, acceptable.

3.9.2 Dynamic Testing and Analysis of Systems, Components, and Equipment In Supplement No. 2 to the SER, the staff found that CE's Comprehensive Vibra-tion Assessment Program conforms to RG 1.20 and is, therefore, acceptable.

CESSAR SSER 3 3-5

However, CE was required to submit a final report summarizing the program results after the testing was completed at the tirst System 80 plant, Palo Verde Unit 1.

This report was also expected to compare the test results with the revised fatigue curves that appear in the Winter 1982 Addendum to the ASME Code (Section III, Table I-9.0 of Appendix I).

CE's letter of February 16, 1984, forwarded a preliminary report on the program results, which the staf f found acceptable. CE submitted a final version on March 7, 1985. The staff has reviewed it and concludes that it meets the above conditions in a satisfactory manner and is, therefore, acceptable.

3.9.5 Reactor Pressure Vessel Internals i In a letter dated March 22, 1985, CE submitted revisions to Section 3.9.5 of CESSAR. Two figures in Section 3.7 (seismic design) were also to be modified as a result of these revisions. These changes were intended to reflect CESSAR design changes to correct deficiencies revealed during hot functional testing at the first System 80 plant, Palo Verde Unit 1. The guide brackets of the con-trol element assemblies (CEAs) were found susceptible to vibration cracking.

I This problem was solved by removing the guide brackets and~ placing snubbers between the upper guide structure barrel and the CEA shroud structure. The CEA guide brackets serve only during assembly and disassembly of the reactnr internals to align the CEAs and extension shafts. Their removal, therefore, does not affect normal plant operation. Their function can be performed during assembly and disassembly of the internals, by a separate alignment tool.

The staff found these corrective measures acceptable for Palo Verde Units 1, 2, and 3 in Supplement No. 7 to the Palo Verde SER (NUREG-0857).

The staff has reviewed the corresponding revisions to CESSAR, has determined that they reflect the design changes approved for Palo Verde Units 1, 2, and i

3, and concludes that they are adequate and acceptable.

3.11 Environmental Qualification of Safety-Related Electrical Equipment i

Supplement No. 2 to the CESSAR SER states that the generic approach to environ-mental qualification of mechanical and electrical equipment described-in CESSAR (Amendments 1 to 8) meets the relevant requirements of General Design Criterion l

(GDC) 4 of Appendix A to 10 CFR 50, and is, therefore, acceptable. However, Section 3.11 in CESSAR references CE's Topical Report CENPD-255-A, Revision 3,

" Environmental Qualification of Class IE Electrical Equipment." When Supple-ment No. 2 to the CESSAR SER was issued, the staff had not made a final decision on the acceptability of that report. )

)

On September 8, 1983, the staff notified CE that CENPD-255-A, Revision 3, was acceptable for referencing in license applications. Subsequently, Amendment 9 to CESSAR was submitted. This amendment includes a' revised Section 3.11 and Appendix 3. 1A incorporating new information on the methodology for radiation qualification below 104 rads and aging, as well as clarification of its appli-cability to both electrical and mechanical equipment. This material had been originally submitted in a letter dated May 10, 1983, and found acceptable by the staff in Supplement No. 2 to the SER. The staff's safety evaluation of Section 3.11 (Amendment 9) and CENPD-255-A, Revision 3, is provided in Appendix J to this report.

CESSAR SSER 3 3-6

I 4 REACTOR  !

1 4.2 Fuel System Design Amendment 9 to CESSAR corrects editorial errors in Section 4.2.3.2, " Fuel Rod Design Evaluation," and Section 4.2.3.4, " Control Element Assembly." These corrections do not change the previous staff analyses and are, therefore, acceptable.

4.3 Nuclear Design  !

4.3.2 Design Description In a letter dated May 31, 1985, CE submitted additional clarifications to CES.5AR including corrections of typographical errors in Section 4.3.2.2.3, "Expectea Power Distributions," and Section 4.3.2.3.2, " Moderator Temperature Coefficients."

These corrections do not change the previous staff analyses and are, therefore, acceptable.

4.4 Thermal-Hydraulic Design D_escriptior. of the Thermal and Hydraulic Design of the Reactor Coolant System In a letter dated October 22, 1984, CE submitted changes to CESSAR that were identified as a result of a review of the technical specifications for Palo Verde Unit 1, the first System 80 plant. Among those changes were two correc-tions to Table 4.4-9, " Thermal and Hydraulic Characteristics Table," of CESSAR Section 4.4.3.6. One was a correction of a previous typographical error; the other was the deletion of the word " core" from the average coolant enthalpy listing. ,

i This change was made so that the values in Table 4.4-9 correctly correspond J' to the reactor inlet and outlet coolant conditions. The staff finds these changes acceptable.

4.4.6 Statistical Combination of Uncertainties In Supplement No. 2 to the CESSAR SER, the staff identified the statistical combination of uncertainties (SCU) as Confirmatory Issue No. 3. More speci- ]

fically, this issue comprises three components:

(1) a review of the safety analysis by CE to confirm that the minimum departure from nucleate boiling ratio (MDNBR) limit of 1.23, imposed as a result of j the staff review for fuel burnup up to 20,000 MWD /MTU, rather than the CE-proposed limit of 1.22 is not violated (2) a commitment from CE to perform a test program to monitor pertinent param-eters during the power ascension test of the first plant referencing CESSAR to determine if a hot-leg flow stratification anomaly exists in the CESSAR plants CESSAR SSER 3 4-1

(3) identification of the interface items for comparison of the plant-specific uncertainty values with the bounding values used in the CESSAR analysis By its letter dated August 31, 1983, CE addressed these issues. With regard to the MDNBR limit of 1.23, CE indicated that the departure from nucleate boiling ratio (DNBR) limit currently installed in the core operating limitr. supervisory system (COLSS) and core protection calculators (CPCs) for CESSAR plants will be greater than or equal to 1.23 and that the conclusions of the CESSAR safety analyses remain valid for the installed COLSS and CPC DNBP, limit. The staff, I therefore, concludes that the DNBR limit issue has been properly addressed for fuel burnup not exceeding 20,000 MWD /MTU. Because the DNBR limit of 1.23 is based on a rod bow penalty of 0.8% DNBR for fuel burnup of 20,000 MWD /MTU, an additional rod bow penalty should be added to the DNBR limit for higher burnup. J In Amendment 9 to CESSAR, a table of rod bow penalty as a function of burnup based on the approved methods described in CENPD-225, Revision No. 3, has been incorporated into the technical specifications (Chapter 16). If a CESSAR plant is operated with burnup exceeding 20,000 MWD /MTV, the DNBR limit shall be increased by a value that is the difference between 0.8% and the rad bow penalty corresponding to the appropriate burnup. This increased DNBR limit shall be the basis for the plant-specific safety analysis and the COLSS and CPC DNBR limits.

With regard to a possible hot-leg flow strati 4 cation anomaly, CE has committed to conduct a test program at Palo Verde Unit 1, the first System 80 plant, to detect the presence or absence of a hot-leg flow stratification anomaly. The test program will consist of recording several hot-leg resistance temperature detector (RTD) signals during the power ascension tests. The selected RTDs will provide sufficient data to determine if a hot-leg temperature transient occurs ,

j while the plant is running at equilibrium conditions. Hot-leg RTD temperatures

)

will be recorded by a strip chart for a period of at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at each of the power levels, 50%, 80%, and 100%. If a hot-leg anomaly is detected, CE also has committed to assess the characteristics of the anomaly so that an appropri-ate strategy can be devised and implemented to handle it. The staff shall be notified about how the uncertainty associated with the anomaly will be accounted for in the CPC calculation. On the basis of this commitment, the staff con-cludes that the possible hot-leg flow stratification anomaly issue has been properly addressed.

l With regard to the measurement uncertainties of the pertinent parameters such as reactor coolant flow rate, inlet temperature, pressurizer pressure, and re-actor power, CE has provided a detailed breakdown of the uncertainty components, their error distributions, and the methods for combining these errors. Because of the bounding nature of the generic parameter uncertainties used in the CESSAR SCO analysis, only those parameters that may vary from plant to plant require verification on a plant-specific basis. CE has identified the interface uncer-tainty parameters that will be verified by plants referencing CESSAR to ensure that they are within the bounding values. The verification will be completed and documented according to the CE quality assurance procedures. If any uncer-tainty values of a plant referencing CESSAR exceed the basic values, the appli-cant shall notify NRC and provide a safety analysis performed with the proper values for that plant. The staff finds this acceptable.

CESSAR SSER 3 4-2

Da the basis of the acceptable resolution of the three components constituting Confirmatory Issue No. 3, as discussed above, the staff concludes that Confirma-tory Issue No. 3 has been acceptably resolved and is, therfore, closed.

4.4.11 Instrumentation for Detection of Inadequate Core Cooling Supplement No. 1 to the CESSAR SER states that this issue remained open for CESSAR pending receipt of additional information on (1) environmental and seis-mic qualification of the instrumentation equipment and (2) modifications to Emergency Procedure Guidelines (EPGs). Supplement No. 2 also evaluates addi-tional information supplied by CE on t.e first of these two items and finds.it acceptable. Amendment 9 to CESSAR incorporates the information.previously approved by the staff into CESSAR, Appendix B, Item II.F.2. j Regarding the second item (modifications to EPGs), Supplement No. 2 to the CESSAR SER indicates that it is still open in the staff's generic review and labels it Confirmatory Issue No. 4. By a letter dated September 28, 1984, CE summarized the actions already taken and committed to submit additional infor-mation requested by the staff by December 1984. This additional information was provided in Appendix A to CEN-152, Revision 2, " Combustion Engineering Emergency Procedures Guidelines," transmitted by a CE Owners Group letter dated November 26, 1984.

The conclusions of the staff's safety evaluation of Appendix A to CEN-152 are:

(1) The response in Appendix A to CEN-152, Revision 2, satisfactorily addresses the staff concerns rahed during the review.

(2) Section 4 of the appendix clearli describes the performance and indicating characteristics of the CE heated junction thermocouple probes and includes suitable cautions about the potential for misinterpretation while reactor coolant pumps are running.

(3) The information in the appendix is suf ficient to permit plant-specific emergency operating procedures to be written giving the operators specific guidance for interpreting vessel inventory indications.

(4) Figures 18 and 19 of the appendix are apparently reversed in relation to the text on pages 46 and 48. A clear description of the representative tempera-ture with algorithm is not provided in the appendix. The errors should be corrected, and a description with algorithm of the representative tempera-ture should be provided as a condition of approval for Appendix A to CEN-152, Revision 2.

(5) Pending the resolution of item 4 above, Appendix A to CEN-152, Revision 2, is acceptable to supplement CEN-152, Revision 2, with respect to the EPGs for use with instrumentation for the detection of inadequate core cooling until Appendix A to CEN-152, Revision 2, is integrated into the text of CEN-152, Revision 3, throughout the procedure guidelines.

Amendment 9 to CESSAR, Appendix B, Item II.F.2, states that the CE Owners Group has developed generic EPGs addressing inadequate core cooling.

CESSAR SSER 3 4-3

On this basis, Confirmatory Issue No. 4 is closed, and Appendix B, Item II.F.2, of CESSAP. Amendment 9, modified to conform to the staff's suggestion (as docu-mented in CE's letter of May 31, 1985), is acceptable.

\

m CESSAR SSER 3 4-4

5 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS 5.1 Summary Description 5.1.1 Schematic Flow Diagram By a letter dated October 22, 1984, Combustion Engineering (CE) submitted changes to CESSAR that were found tn be needed as a result of a review of the technical specifications for the first System 80 plant (Palo Verde Unit 1). In Section 5.1.1 of CESSAR, the proposed change involved the correction of a pre-i vious typographical error in Table 5.1.1-1, " Process Data Point Tabulation."

l Because this change is typographical and does not alter the staff's original basis for acceptance, the staff finds it acceptable.

5.1.2 Nuclear Steam Supply System--Balance-of-Plant Interface Requirements In a Isolation Valve Closure Time|letter dated December 5, 1984]], CE submitted additional changes to CESSAR that were found necessary as a result of the review of Pa'o Verde Unit 1 Tech-nical Specifications. In particular, the interface. requirements for main steam isolation valve (MSIV) and main feedwater isolation valve (MFIV) cicsure times were revised to ensure consistency with the assumptions of the safety analyses.

The MSIV and MFIV closure times also had to be corrected in Section 5.4.5 of CESSAR, " Main Steam Line Isolation System," and in Section 6.2.1.4, " Mass and Energy Release Analysis for Postulated Secondary System Pipe Ruptures Inside Containment." The staff has reviewed these changes and concludes that the mod-ified closure times are consistent with those used in CESSAR Chapter 15 safety analyses and are, therefore, acceptable.

5.2 Integrity of Reactor Coolant Pressure Boundary 5.2.2 Overpressurization Protection The CE letter dated October 22, 1984, also modified CESSAR Section 5.2.2.4.4,

" Safety Injection System Valves," to more accurately reflect the actual valve design and specifications. The staff finds that these changes do not affect emergency core cooling system performance and are acceptable. I J

As a result of startup testing at Palo Verde Unit 1, the first System 80 plant, CE, in a Isolation Valve Closure Time|letter dated December 5,1984]], identified other changes to CESSAR; in particular, Section 5.2.2.10, " Overpressure Protection During Low Temperature Conditions," and Figure 5.2-2 were modified to reflect a lower maximum temperature difference of 100 Fahrenheit degrees across a steam generator.

Figure 5.2-1 was also modified to reflect the typical difference in height between the pressurizer and the shutdown cooling system.

The staff has reviewed these changes, verified that the 100-Fahrenheit-degree I temperature difference was used in the analysis for developing Figure 5.2-2, and concludes that these changes do not affect the low-temperature overpressure protection and are, therefore, acceptable.  !

l CESSAR SSER 3 5-1 1 l

I

5. 4 Component and Subsystem Design 5.4.2 Steam Generators Amendment 9 to CESSAR included minor editorial changes in CESSAR Section 5.4.2.4, " Steam Generator Materials." The staff has reviewed them, finds that they are editorial and provide the appropriate cross references, and hence concludes that they are acceptable.

5.4.3 Shutdown Cooling (Residual Heat Removal) System 5.4.3.1 Confirmatory Issue No. 1 .

Supplement No. 2 to the CESSAR SER indicates that CE has not demonstrated, to the complete satisfaction of the staff, that the System 80 design can be cooled down to the shutdown cooling system (SOCS) initiation using only safety-related equipment (one of the requirements of Branch Technical Position (BTP) RSB 5-1 (NUREG-0800)). The staff met with CE on May 25, 1983, to discuss CE's proposal for a reanalysis of the natural circulation cooldown, which would address the staff's concerns. As a result of this meeting, in a letter dated June 7, 1983, CE committed to submit a reanalysis of the cooldown, assuming (1) no letdown is available, (2) the reactor coolant gas vent system is available to help reduce the void in the reactor vessel upper head (RVUH), (3) there are no stuck rods, and (4) there are conservative mixing and condensation in the RVUH. The staff found this commitment acceptable; this issue became Confirmatory Issue No. 1.

In a letter dated August 12, 1983, CE provided a report to fulfill the above commitment. This report contains an analysis of a full natural circulation cooldown from hot standby conditions to temperatures and pressures that permit the initiation of the shutdown cooling system. The analysis was performed using only safety-related equipment concurrent with a loss of offsite power and an assumed single active failure. The results of this analysis indicate that the total time required to take the CE System 80 plant from hot standby con-ditions to temperatures and pressures that permit use of the shutdown cooling system is approximately 10.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. This time includes maintaining the plant in hot standby for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> before commencing a cooldown. The safety related auxiliary spray and charging systems are used for reactor coolant system (RCS) depressurization during cooldown. The reactor coolant gar vent system also is used. During this simulated cooldown process, approximately 220,000 gallons of condensate water is required, less than the 300,000 gallons available according to plant technical specifications. NUREG-0737 specifically states that the reactor coolant gas vent system does not have to meet the single-failure cri-terion of Institute of Electrical and Electronics Engineers (IEEE) Standard i 279-1971. However, the design of the Palo Verde units has included a safety-  !

related RCS vent system that meets the single-failure criterion. To make this  !

CE cooldown analysis applicable to other plants that use CE System 80 nuclear  !

steam supply systems, the staff requires that CE provide an interface require- l ment to specify a safety-related RCS vent system that meets the single-failure criterion for CE System 80 plants. This is needed to validate this cooldown analysis.

During the course of its review, the staff identified a potential single failure in the auxiliary pressurizer spray (APS) system that would render the CESSAR SSER 3 5-2 I

1

__o

system unable to supply charging fluid to the pressurizer spray nozzle. APS flow is initiated from the control room by opening at least one of the redundant'(parallel) auxiliary spray valves (CH-203 or CH-205) in combination with the closure of the loop charging valve-(CH-240). The staf f expressed concern that a failure of CH-240 to close would negate the safety function of the APS system by causing insufficient flow to the pressurizer spray nozzle.

CE was advised of this concern by a request for additional information dated March 27, 1984.

In view of th'e above, the conclusions of this safety evaluation are as follows:

Subject to an acceptable modification of the APS system, an interface require-ment in CESSAR for a safety-related reactor coolant system (RCS) gas vent system, and a demonstration of the natural circulation cooldown process at the first CE System 80 plant (Palo Verde Unit 1), the results of the natural circulation cooldown analysis for CE System 80 meet the requirements of BTP RSB 5-1 and, therefore, are acceptable.

(1) Acceptable Modification of Loop Charging Valves ,

k In response to the staff's request for additional information, CE, in a j letter dated September 18, 1984, committed to add an extra valve in series-with the existing loop charging valve (CH-240) to eliminate the single-failure vulnerability.

The staff takes the position that the CESSAR System 80 APS system should .

be treated as safety related in accordance with Appendix A to 10 CFR 50 l and 10 CFR 100, since it is required for safe shutdown of the plant and also for mitigating the consequences of a steam generato" tube rupture accident should the main pressurizer spray system become unavailable.

Discussions with CE subsequent to the September 18, 1984 submittal revealed that CE planned to use non-Class 1E buses to power CH-240 and the new series valve even though the valves are considered to be safety related. The staff found this unacceptable because the new series valves are required to perform a safety function as discussed &bove.

The staff takes the position that if any equipment (pumps, valves) is re-quired to perform a safety-related function, the mechanical and electrical components including any associated electrical power distribution system should also be treated as being safety related and should be implemented as safety related. Thus, the existing loop charging valve (CH-240) and the additional' series valve used to isolate normal charging flow should satisfy the single-f ailure criterion and should not be dependent on the use of non safety-related equipment, including the power distribution system. Specifically, each valve should be powered from separate, elec-trically independent, Class 1E buses; any exception to this position must be appropriately justified.

The staff notified CE of this position by letter dated February 8, 1985.

CE responded by letter dated March 22, 1985. Review of the information supplied in the March letter shows that CE continues to propose the use of non-Class 1E buses to power the subject valves but has committed, as an alternative to the use of Class 1E power, to provide isolators to the CESSAR SSER 3 5-3

l I

)

non-safety-related power supply control circuits for protection of valves CH-239 and CH-240. The staff considers this commitment acceptable.

However, the staff does require that the isolators be Class IE protective j devices and that they be implemented to ensure that the capability of the ^

valves to perform the intended safety functions is not degraded below an acceptable level as a result of all possible circumstances (i.e., l overvoltage, undervoltage, low frequency, design-basis events,'etc.) I associated with its non-Class 1E power buses. This will require i plant-specific failure modes and effects analyses to support the required i safety functions of the valves. The Class IE electrical protective devices must meet the applicable portions of IEEE Standard 279-1971 (i.e, Sections 4.2, " Single Failure Criterion"; 4.3; " Quality of Components";

4.4, " Equipment Qualification"; 4.5, " Capability for Test and l Calibration"; 4.18, " Identification of Protective Actions"; 4.20, (

. "Information Read-Out"; 4.21, " System Repair"; and 4.22, "Identi fi- q l cation"). These requirements were established during the review of the  !

Palo Verde application and are to be implemented in the design. 1 i

On the basis of the above commitment and satisfactory adherence to the requirements set forth by the staff, this issue is resolved.

(2) Safety-Related RCS Gas Vent System l As mentioned earlier in this safety evaluation, the staff requires that CESSAR provide an interface requirement to specify a safety-related RCS gas vent system that meets the single-failure criterion for CE System 80 l plants. To meet this requirement, CE, in a letter dated March 22, 1985, i provided the new interface requirement that the reactor coolant gas vent  !

system (RCGVS) shall be designed as safety grade and shall meet the  !

single-failure criterion of IEEE Standard 279-1971.

The staff concluded that this interface requirement is acceptable.

(3) Demonstration of the Natural-Circulation Cooldown Process at Palo Verde Unit 1 The Palo Verde Unit 1 boron-mixing and natural circulation cooldown test program and procedures have been reviewed and approved by the staff. On January 24, 1986, a boron mixing and natural-circulation test was performed at Palo Verde Unit 1, the leading plant of CESSAR System 80 d2 sign. By a letter dated February 9,1987, Arizona Nuclear Power Project (ANPP) sub-mitted the post-test report for the staff's r.eview and approval. The staff's evaluation of this test report will be reported in a future CESSAR SER supplement.

On September 12, 1985, an unexpected event occurred at Palo Verde Unit 1 that caused the loss of all three charging pumps. Auxiliary pressurizer spray (APS) capability was lost. The event and the staff evaluation are described in detail in the Palo Verde SER (NUREG-0857, Supplement No. 9, December 1985).

In the staff investigation of this event, design inadequacies were identified in the water supply system to charging pumps. The major deficiencies are:

CESSAR SSER 3 5-4

(1) The level instrumentation of the volume control tank (VCT) is not designed to safety grade standards. Its failure could cause hydrogen binding in all three charging pumps.

(2) Valves CH-501 and CH-503 are not designed to meet the single-failure cri-terion. A postulated failure of valve CH-501 at its open position or a failure of valve CH-536 at its closed position could cause loss of water supplies to the charging pumps, until numerous operator actions ar e per-formed to vent hydrogen gas from the charging pumps and restore their operation.

The staff does not consider that the APS system design, with the above t.tated deficiencies in its supporting systems, meets the safety grade standards. As a result, the staff in a letter to CE dated October 29, 1985, expressed its concerns and requested additional information as follows:

(1) The staff is concerned about the APS system and its supporting systems for the following reasons:

(a) The natural circulation cooldown analysis submitted for CESSAR relies on the APS system and its supporting systems for reactor coolant system (RCS) depressurization during cooldown.

(b) The APS system and its supporting systems are relied on to mitigate

f. the steam generator tube rupture (SGTR) accident and possibly other accidents requiring depressurization of the RCS when the main pres-surizer spray is not available.

, (c) The staff relied heavily on the reliability of the APS system and its supporting systems in its evaluation reported in NVREG-1044,

" Evaluation of the Need for a Rapid Depressurization Capability for Combustion Engineering Plants."

(2) The staff asked CE to respond to the following request for additional information:

(a) Address need for a safety grade APS system and supporting systems (i.e., from its water source through the spray nozzles) considering the systems' functions as discussed above.

(b) Propose any changes in the CESSAR design and/or interface require-ments as a result of item 2a above.

In response to the staff request, CE, in a letter dated February 6, 1987, pro-a posed certain design improvements for the APS system; but CE has not provided sufficient information for determining whether the APS system, as modified, is a safety grade system. In a letter to CE, dated March 24, 1987, the staff repeated its request that CE should provide sufficient information to establish that the APS system and its associated water supply system are designed to safety grade standards. The staff takes the position that the APS system should be designed to safety grade standards in order to take credit for its availability in the SGTR accident analysis. The safety grade AP3 system is also needed to fulfill the requirements to BTP RSB 5-1 for Class 1 plants.

1 i

CESSAR SSER 3 5-5

h l

d Thus, in order for future applications to be able to reference CESSAR with respect to BTP RSB 5-1, the APS system must meet safety grade requirements.

CE has not responded to these' staff concerns. The staff considers this issue confirmatory and will report the resolution of this issue in a future supplement to the CESSAR SER. '

The staff concluded that Confirmatory Issue No.1 can be closed after (1) the staff approves the post-test report of the boron-mixing and natural circulation test at Palo Verde Unit 1 and (2) the staff concerns about the design of the APS system addressed above are resolved.

5.4.3.2 Reactor Coolant Piping In a letter dated March 22, 1985, CE revised Section 5.4.3 of CESSAR to reflect design changes that were decided after hot functional testing of the first System 80 plant, Palo Verde Unit 1. These changes were made to correct defi-ciencies revealed during this testing: One thermal sleeve on a safety injection nozzle became loose and another sleeve fell from the nozzle.

A stress analysis of all plant transients indicated that the thermal sleeves do i

not protect the crucial locations in the nozzle and that their removal does not affect the operability of the nozzle. The removal of the thermal sleeves in Palo Verde Unit I was approved in Supplement No. 7 to the Palo Verde SER ,

(NUREG-0857). The proposed revisions of Sections 5.4.3.1 and 5.2.4.3 of CESSAR reflect this removal. The staff concludes that they are adequate and acceptable.

5.4.3.3 Decay Heat Removal Without Power-Operated Relief Valves  ;

Supplement No. 1 to the CESSAR SER states that the CESSAR RCS is designed without power operated relief valves (PORVs) on the pressurizer. Decay heat l removal capability ultimately relies on heat removal via steam generators using '

emergency feedwater and atmospheric steam dump valves. Supplement No. 1 also discusses the applicability of PORVs to System 80 plants and concludes with the statement: "Should the NRC decide that design or procedural changes are necessary, CE will be required to implement them for CESSAR."

Since Supplement No. I was issued in March 1983, the staff's detailed sys-tematic study of the need for a depressurization capability in current CE plants without PORVs has been completed. An NRC policy paper (SECY-84-134) was issued on March 23, 1984, which concluded that the decision regarding PORVs for these CE plants (designed without PORVs) should be deferred and incorporated into any requirements associated with the technical resolution of USI-A-45,

" Shutdown Decay Heat Removal Requirements." Because part of the benefit of the PORVs was based on their ability to provide an alternate decay heat removal path (" feed and bleed"), any improvements in decay heat removal capability that might be promulgated as a result of the USI A-45 assessment could reduce the net benefit of PORVs. The report attached to the policy paper, containing the staff's evaluation, was issued in its final form as NUREG-1044 in December 1984.

In a letter dated A.ugust 3, 1984, CE requested that the CESSAR SER be modified to reflect the above conclusions. In a letter dated November 15, 1984, the CESSAR SSER 3 5-6

staff informed CE that the requested change is app.ropriate and stated that the SER would be modified ~accordingly. The above update documents the current NRC position.

5.4.5 Pressurizer l '

In a letter dated October 22, 1984, CE submitted changes to CESSAR that were found necessary as a result of a review of the technical specifications for Palo Verde Unit 1. In Section 5.4.10.2 of CESSAR, Figure 5.4.10-2 was modified to reflect the actual installed pressurizer level program for Palo Verde Unit 1.

The staff concludes that this new figure is bounded by the initial conditions specified in Table 15.0.5 of CESSAR and is, therefore, acceptable.

In a Isolation Valve Closure Time|letter dated December 5, 1984]], CE proposed a change in Section 5.4.10.3 as a result of startup testing conducted at Palo Verde Unit 1. This change reflects the fact that pressurization rate testing is not performed during hot

' functional testing; the pressure control setpoints are determined analytically and checked during power ascension testing.

The staff has reviewed this change, agrees that it reflects the approved hot functional testing program, and hence finds it acceptable.

5.4.6 Safety and Relief Valves ,

In a letter dated August 11, 1903, CE informed the staff that Electric Power I Research Institute tests discussed in a CE Owners Group Report (19^2) previously transmitted to the staff in a letter dated December 20, 1982, showed that safety valve blowdown settings for the System 80 design would be i necessarily higher than those listed in CESSAR Section 5.4.13 and Appendix 5A. l CE proposed revisions to CESSAR that involved changes in primary and secondary system safety valve parameters and analyses to justify these changes. The re-vised Section 5.4.13 and Appendix SA were later resubmitted to the staff as part of CESSAR Amendment 9. The changes to the safety valve parameters in CESSAR Tables 5.4.13-1 and 5.4.13-2 reflect the 18.5% maximum blowdown through the pressurizer safety valves and approximately an 11% maximum blowdown through the main steam safety valves. Both of the proposed changes to the maximum blowdown values exceed the values assumed in the CESSAR Chapter 15 safety 1 analyses. CE referenced its own Topical Report (CEN-227, " Summary Report on the Operability of Pressurizer Safety Relief Valve in CE Designed Plants"), to justify the adequacy of the changes made to the pressurizer safety valves. On the basis of its review of the information contained in CEN-227, the staff requested additional information from CE on the following concerns: (1) iden-tification of the limiting event, (2) the potential moisture carryover into the .

I pressurizer safety valve, and (3) the effect of these changes on the results of safety analyses.

In response to the staff's request, CE submitted a letter dated June 4, 1985;  !

in this letter, CE stated that the results of an analysis demonstrate the adequacy of the proposed changes.

CESSAR SSER 3 5-7

The results of the CE analysis show that the feedwater line break (FWLB) with loss of offsite power (LOOP) is the most adverse event in the decreased heat removal category which results in the greatest increase in pressurizer water level and RCS temperature. The blowdown of the pressurizer safety valves is assumed to be 18.5% (closure at pressures below 2040 psia). The results of the CE analysis have shown a predicted maximum pressurizer water volume of 1213 ft3 To conservatively account for subcooled insurges and ph&se separation during the transient, CE made a conservative adjustment to the limiting FWLB and LOOP to increase the predicted maximum level response by 306 f t3 to reach a two phase pressurizer level of 1519 ft3 Even for this limiting case, the pressurizer level (at 87%) is 12% below the elevation of the bottom of the pressurizer safety valve nozzle (at 99.4%). This corresponds to more than 4 feet below the nozzle. Because of this large margin, moisture carryover through the pressurizer safety valves is believed to be minimal. The results of the analysis have also shown that sufficient margin exists on RCS subcooling

(>20 F) and that the acceptance criteria with respect to maximum RCS pressure and minimum departure from nucleate boiling ratio (MDNBR) are : net for the event analyzed. The results of the CE analysis also indicated that the design changes to the pressurizer safety valve will not affect the results of other CESSAR Chapter 15 safety analyses.

CE has also reviewed the increased blowdown of the main steam safety valves (MSSVs). Because increasing MSSV blowdown changes neither the valve's capacity nor the opening setpoint, the maximum secondary pressure of any CESSAR analyzed event will remain unchanged. Increasing the MSSV blowdown can, however, affect secondary steam releases and the resulting radiological doses. For the safety analyses in CESSAR, the effect of increased MSSV blowdown on radiological doses was evaluated by CE and it was found that only the SGTR events would have mea-surable changes in the consequences. For the SGTR with loss of power and with a stuck-open atmospheric dump valve, the increased MSSV blowdown setting was used for the analysis already in CESSAR with acceptable radiological consequences.

The staff has evaluated the analysis submitted by CE and agrees with CE's con-clusions. Therefore, the staff finds that the CE proposed design changes to the primary and secondary safety valve minimum blowdown set pressure are a_ eptable.

CESSAR SSER 3 5-8

l 6 ENGINEERED SAFETY FEATURES 6.2 Containment Systems 6.2.1 Containment Functional Design ,

6.2.1.2 Mass and Energy Release Amlysis for Postulated Secondary System Pipe Ruptures Inside Containment By a Isolation Valve Closure Time|letter dated December 5, 1984]], CE submitted changes to CESSAR that were found necessary for consistency as a result of the technical specification review performed for Palo Verde Unit 1, the first System 80 plant. Specifi-cally, Tables 6.2.1-11 through 6.2.1-20 (" Sequence of Events") were modified to avoid any inconsistency with technical specifications and interface requirements.

The staff has reviewed these changes and finds that they are minor and have little or no effect on the containment-related safety analyses. Therefore, these changes are acceptable.

6.2.2 Containment Spray System In a letter dated October 22, 1984, CE identified an apparent typographical error in the CESSAR SER. In Section 6.2.2, page 6-8, the safety evaluation of the containment spray system mentions that the elapsed time from receipt of the containment spray actuation signal to the delivery of spray flow is 50 seconds.

CE pointed out that this time is, in fact, 58 seconds, as documented in Appen-dix 6A of the CESSAR FSAR. The staff agrees with CE. The correct value is i 58 seconds.

6.2.4 Containment Isolation System l

CE's October 22, 1984 submittal also included a change in CESSAR Table 6.2.4-1, which gives the positions of the containment isolation system valves under various plant conditions. The change replaced the postaccident position of

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valve CH-524 (penetration number 41), which was "open," by "open'or close." q The staff has reviewed this change, finds that it better reflects the post-accident use of the corresponding line, as determined by the technical specifi- l cations, and concludes that it is acceptable.

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6.3 Emergency Core Cooling System 6.3.1 System Design 6.3.1.3 Interface Requirements By a letter dated May 31, 1985, CE submitted additional CESSAR clarifications ,

intended to be incorporated into CESSAR Amendment 10. These include modifica-tions of the text describing the safety injection system delay time. In Sec-tion 6.3.1.3 to CESSAR, page 6.3-2, the sentence that reads "Each emergency CESSAR SSER 3 6-1 i

generator and the' automatic sequencers necessary for generator loading shall be designed such that flow to the core is attained within a maximum of 30 seconds" is modified and becomes "Each emergency generator and the automatic sequencers necessary for generator loading shall be designed'such that flow is delivered to the RCS [ reactor coolant system] within a maximum of 29 seconds af ter Safety Injection Actuation Signal.(SIAS) is generated." ~~-

The staff agrees that the modified text is'more accurate and finds it acceptable.

To ensure consistency with this change,.other sections of CESSAR have to be similarly modified: Section 6.3.3, " Performance Evaluation," pages 6.3-24, 6.3-27, and 6.3-30, and Section 8.3, "0nsite Power System," Table 8.3.1-4,

" Required Standby Generator Loads."

The staff has reviewed these changes and finds them acceptable.

6.3.3 Testing 6.3.3.1 Small-Break Analysis By a Isolation Valve Closure Time|letter dated December 5, 1984]], CE had submitted modified Table 6.3.3.3-6,

" Times of Interest for Small Breaks," and Table 6.3.3.5-1, Sequence of Events for Representative large and Small Break LOCAs," to avoid the appearance of any inconsistency with Technical Specifications and interface requirements.

The staff has reviewed these changes, agrees that they are justified, and finds them acceptable.

6.3.3.2 Post-Loss of-Coolant-Accident Long-Term Cooling Amendment 9 to CESSAR introduces two changes to Section 6.3.3.4:

(1) The criterion for entering the shutdown cooling mode after a loss-of-coolant accident (this criterion ensures that the break is small enough for successful operation in the shutdown cooling mode) is modified as a result of CE's latest long-term. cooling analysis.

(2) The maximum instrument errors are raised to provide consistency with the actual narrew-range pressurizer pressure channel uncertainty.

To initiate long-term cooling after a loss-of coolant accident (LOCA), either one of two procedures is used, depending on the break size. Shutdown cooling is initiated if the break is sufficiently small. For large-break LOCAs, simultaneous acid flushing. hot and cold-leg injection will maintain core. cooling and boric The plant operator initiates the ap basis of the reactor coolant system (RCS) pressure.propriate procedure In Amendment on the 9 to CESSAR, CE defines the criteria for selecting operating procedures for long-term cooling.

If the RCS pressure is above 538 psia between 8 and 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> after the LOCA, the RCS is filled and there is assurance that all conditions-for e-ntering the shut-down cooling mode can be established.

CESSAR SSER 3 I 6-2

Cooling of the RCS with the steam generators continues until the RCS temperature and pressure are lower than the maximum design shutdown cooling system entry conditions. If the indicated RCS pressure has fallen below 538 psia at 8 to 9 I hours, the break is assumed to be too large for proper operation of the shutdown 1 cooling system. However, in this case simultaneous hot-leg and cold-leg injec-tion will cool the core and flush boric acid inside the core. The staff finds this long-term-cooling plan acceptable. The changes proposed in CESSAR Sec-tion 6.3.3.4 are, therefore, acceptable.

6.5 Containment spray as a Fission Product Cleanup System i l

In Supplement No. 2 to the CESSAR SER, the staff evaluated the iodine-removal i function of the containment spray system. The CESSAR design permits long-term iodine control by restarting hydrazine addition to the containment spray as needed. This implies that provisions exist for refilling the hydrazine tank (called spray chemical storage tank (SCST) in CESSAR) under postaccident con-ditions.

By a Isolation Valve Closure Time|letter dated December 5, 1984]], CE proposed a revision to Section 7.16 of CESSAR Appendix 6B, " Iodine Removal System Licensing Report," so as to remove unwarranted restrictions on the transfer of hydrazine to the SCST and to clarify the arrangement usod for this transfer.

After discussions with CE, the staff hcs determined that the transfer line between the hydrazine backup tank and the SCST need not be safety grade be-cause damage to thi, line could be repaired during the activities associated with the SCST replenishment. Therefore, the staff finds the proposed change acceptable.

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7 INSTRUMENTATION AND CONTROLS 7.1 Introduction By a Isolation Valve Closure Time|letter dated December 5, 1984]], CE identified changes to CESSAR as a result of startup testing conducted at Palo Verde Unit 1, the first System 80 plant.

One change was made in Chapter 7 (Section 7.1.2.10) to clarify conformance to IEEE Standard 384-1971. The modified text indicates that the plant protection system (PPS) cabinet contiol and instrumentation circuits do not meet the 6-inch separation criterion of 'EEE Standard 384 and that tests and analyses have been completed to justify the nonconformance.

l This issue was resolved during the Palo Vede operating license review, as docu-mented in the Palo Verde SER (HUREG-0857), Supplement No. 7. On the basis of its review of this issue for Palo Verde, the staff finds the CESSAR change '

acceptable.

7.2 Reactor Coolant System Reactor Protective System In a letter dated October 22, 1984, CE submitted changes to CESSAR that were found necessary to ensure consistency with the Palo Verde Unit 1 Technical Specifications. One of these changes was made in Table 7.2-1, " Reactor Pro-tective System Bypasses," to indicate that the removal of the departure from nucleate boiling ratio and local power density trip bypasses occurs when the reactor power exceeds 1% of rated thermal power instead of 10 4%. During the Palo Verde technical specification review, CE informed the staff that 10 4% was incorrect and that 1% was the proper value to use. The staff reviewed this change and approved it. On the basis of this review, the staff finds the CESSAR change acceptable.

By a Isolation Valve Closure Time|letter dated December 5, 1984]], CE described other changes to Section 7.2.1 l to eliminate any inconsistency between the Palo Verde technical specifications  !

and CESSAR interface requirements. These changes clarify the inclusion of reactor protection system sensor response times in the safety analyses. The staff has verified that they do not affect the results and conclusions of the safety l analyses and they do not result from any design modification. On the basis of these findings, the staff concludes that the changes are acceptable.

7.3 Engineered Safety Features Actuation System In the letter dated October 22, 1984, CE changed CESSAR Table 7.3-3 to indicate that the steam generator differential pressure is monitored for emergency feed-water actuation. This revision is consistent with the Palo Verde Technical Specifications (Section 3/4.3.2) and with the system actuations basis of CESSAR (Section 7.3.2.2.6), which have been reviewed and approved by the staff. It is, therefore, acceptable.

CESSAR SSER 3 7-1

7. 5 Safety-Related Display Instrumentation By a letter dated May 31, 1985, CE submitted additional clarifications to CESSAR. A change was made to Table 7.5-3, "Postaccident Monitoring Instru-mentation," to reflect the fact that the reactor coolant system pressure instrumentation is now Class 1E instead of non-Class 1E, as provided previously.

The staff had previously approved this modification and finds this change to 4 CESSAR acceptable.

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CESSAR SSER 3 7-2 l

l 9 AUXILIARY SYSTEMS 9.1 Fuel Storage Facility 9.1.4 Fuel Handling System In a letter dated October 22, 1984, CE submitted changes to CESSAR that were 3 identified as a result of the technical specification review for Palo Verde Unit 1, the first System 80 plant. Section 9.1.4.1.2 of CESSAR, " Fuel Handling Equipment," was revised to delete an outdated procurement specification (Hoist Manufacturing Institute HMI-100-74). The staff finds this change acceptable.

Subsequently, in a letter dated May 31, 1985, CE submitted an additional clari-fication to Section 9.1.4.3.4 of CESSAR, " Safety Evaluation, Fuel Handling."

The purpose of this revision is to distinguish between the radiation levels at the refueling pool work station (inside the reactor building) and at the spent fuel pool work station (in the spent fuel pool building) during fuel handling.

The staff has reviewed this revision and finds it acceptable.

9.2 Water Systems 9.2.3 Demineralized Water Makeup System In a Isolation Valve Closure Time|letter dated December 5, 1984]], CE submitted changes to CESSAR that were identified as a result of startup testing at Pr Verde Unit 1. Table 9.2-1,

" Demineralized Water Makeup System Effluent Lim ,, " was revised. At a meeting on April 3, 1985, the staff obtained additional information on the reasons for the revisions as documented in a letter transmitted by CE on April 25, 1985.

As a result of these discussions, the staff finds the changes acceptable.

9.3 Process Auxiliaries 9.3.4 Chemical and Volume Control System In the letter dated October 22, 1984, CE submitted changes to CESSAR Table 9.3-1,

" Operating Limits." In a subsequent Isolation Valve Closure Time|letter dated December 5, 1984]], CE submitted additional changes to Section 9.3.4 and Table 9.3-1 to reflect the latest guidance on chemistry controls. Discussions documented in CE's April 25, 1985 letter enabled the staff to conclude that these changes are acceptable. Upon implemen-l tation, the chemistry controls should be reviewed against the current industry l guidance (September 1986, PWR Primary Water Chemistry Guidelines).

Finally. in the letter dated May 31, 1985, CE corrected a typographical error in Table 9.3-6, " Chemical and Volume Control System Process Flow Data." The staff finds this change acceptable.

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CESSAR SSER 3 9-1 j

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i 10 STEAH AND POWER CONVERSION SYSTEM 10.3 Circulating Water System 10.3.1 Secondary Water Chemistry In a Isolation Valve Closure Time|letter dated December 5, 1984]], CE submitted revisions to CESSAR Section 10.3.4 to reflect the latest guidance on water chemistry controls. The staff requested justifications for these changes by a letter dated March 25, 1985. In a reply dated April 25, 1985, CE provided these justifications.

The revised condensate, feedwater, and steam generator secondary water chem-istry limits are consistent with the recommendations in "PWR Secondary Water Chemistry Guidelines "(Electric Power Research Institute (EPRI) NP-2704-SR, October 1982), which the staff has endorsed, and, therefore, are acceptable.

The changes also include monitoring additional' impurities, such as sulfate, which are not currently in the guidelines. The staff finds the changes acceptable. Upon implementation, the limits should be reviewed against current EPRI Guidelines (i.e., March 1987, Revision 1).

The staff's proposed resolutions of Generic Issue A-4 include a recommendation that licensees implement a condenser inservice inspection program. The pro-gram should be defined in plant-specific, safety related procedures and should ir.clude procedures to implement a condenser inservice inspection program that will be initiated if condenser leakage is of such magnitude that a power reduc-tion corrective action is required more than once every 3-month period. This will be reviewed under the generic resolution of Generic Issue A-4.

On the basis of the above evaluation, the staff concludes that the proposed changes in the water chemistry limits for the steam generator secondary water, feedwater, and condensate water are acceptable.

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CESSAR SSER 3 '10-1 s

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'i 14 INITIAL TEST PROGRAM 14.1 Introduction CESSAR does not describe a complete initial test program. A description of the initial test program for the balance of plant must be provided in the applica-tions referencing CESSAR. The CESSAR SER describes the staff's review of the CESSAR initial test program and lists some of the changes made by CE to this program as a result of the review. The SER concluded that the initial test program describcd in CESSAR complies with the acceptance criteria of Section 14.2 of the Standard Review Plan (SRP, NUREG-0800). It also concluded that the requirements of GDC 1 and Section IV of 10 CFR 50, Appendix B, will be met for that portion of a facility's initial test program that is described in CESSAR.

However, at the time this review was perforned, no facility referencing CESSAR had yet undergone initial tests. The construction of Palo Verde Unit 1 prompt-ed a continuous reevaluation of CESSAR by CE. CE sent the staff several submit-tals concerning changes to Chapter 14, some tefore Palo Ver b Unit 1 started its initial test program and some that were found necessary 13 a result of the experience at this first System 80 plant. These changes are discussed below.

Initial testing at Palo Verde Unit 1 also necessitated changes to other CESSAR chapters. They will be briefly mentioned for completeness; the most signif-icant changes have already been discussed in the appropriate sections of this supplement.

14.2 Specific Information To Be Included in the FSAR 14.2.1 Conformance of Initial Test Programs With Regulatory Guides and Industry Standards By a letter dated August 30, 1983, CE deleted the requirement for hot, nc-flow control rod drop testing and stated that testing of rod drop times under full-flow corditions is more limiting than under no-flow conditions. The staff's review included examination of data from recent startup testing at a CE facility. These data showed that, in every case, the rods scram faster under hot, no-flow conditions than under full-flow conditions. By a letter dated March 26, 1984, the staff advised CE of its conclusion that the deletion of hot, no-flow rod drop testing from the CESSAR initial test program is accep-table and that the previous SER conclusions are still valid. Subsequently, Section 14.2.7.1 of CESSAR was modified to reflect this change, as part of Amendment 4 to CESSAR. The staff finds this amended Section 14.2.7.1 l acceptable. '

By a letter dated October 16, 1934, CE made other changes to CESSAR Section ,

14.2.7.1, removing the rod drop tests under cold (260 F), partial-flow condi- '.

l tions, because experience from previous startups showed that testing under hot, I full-flow conditions was also more limiting. The staff has reviewed this change, agrees with CE's justification, and finds it acceptatile.

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I CESSAR SSER 3 14-1

14.2.2 Initial Fuel Loading and Initial Criticality CE's October 16, 1984 letter also changed CESSAR Section 14.2.10.1, " Initial ,

Fuel Loading," to delete the description of the containment evacuation alarm because this equipment is not provided in the CESSAR scope and, according to CE, there is no requirement for such a device. Palo Verde Unit 1 is not equip-ped with such a device. The staff finds this change acceptable.

14.2.3 Individual Test Programs Amendment 9 to CESSAR made several changes in CESSAR Se M in 14.2.12 that were qualified by CE as clarifications and updating. These changes are:

1 (1) The prerequisites of the natural circulation test, one of the power ascen-sion tests, were revised to clarify the minimum level of decay heat neces-sary for the performance of this test.

(2) The test procedure requirements for the unit load tesnsient test and turbine trip test were updated.

(3) The method for monitoring noise and vibration caused by waterhammer and the corresponding acceptance criteria were revised in the main and emergency feedwater system test program.

The staff reviewed these changes, found the first two acceptable, and sent a request for additional information about the third to CE by a letter dated November 26, 1984. The staff took the position that the new wording of the waterhammer monitoring part of the main and emergency feedwater cystem test resulted in less precise acceptance criteria than in the approv$d CESSAR ver-sion, making evaluation of successful performance more difficult. After sev- )

eral discussions with the staff, CE finally submitted new acceptance criteria i for waterhammer monitoring in a letter dated April 30, 1985. The stafi has i reviewed the criteria and finds them acceptable.

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CE's letter of October 16, 1984, also provided numerous changes in CESSAR Sec-tion 14.2.12 that resulted from a review of this material in preparation for

, st wtup of Palo Verde Unit 1. These changes were submitted to facilitate an l improved testing procedure. The staff has reviewed these changes, finds that they are improvements or clarifications of existing procedures, and concludes that they are acceptable.

Subsequently, in a Isolation Valve Closure Time|letter dated December 5, 1984]], CE submitted another change to CESSAR Section 14.2.12 to correct an editorial error. The staff finds that this change is indeed editoriai,-does not impact its previous evaluation, and is acceptable. -

14.3 Impact of Palo Verde 1 Hot Functional Testing on Other CESSAR Chapters By a letter dated September 9, 1983, CE advised the staff that during hot func-tional testing at Palo Verde Unit 1, several problems arose on CE-supplied equip-ment. CE then reviewed the potential effect on the content of CESSAR of the CESSAR SSER 3 14-2

modifications decided on or contemplated to solve these problems. T_he modifi-cations involved the reactor coolant pumps, the resistance temperature detector (RTD) thermowells, the safety injection nozzle thermal sleeves, and the control element assembly (CEA) shroud assembly. In this submittal, CE stated that the {

design changes to the reactor coolant pumps probably would not affect the cor-  !

responding CESSAR text, centered on functional performance rather than detailed I design description. Similarly, because specific details of the thermowell design are not discussed in CESSAR, CE felt that the. changes made at Palo Verde Unit 1 would not entail any revision of the appropriate CESSAR section.. No definite statement on possible CESSAR changes could be made at that time by CE regarding the other two design modifications, which were not yet fully defined.

In view of the above, the Final Design Approval for CESSAR System 80 was issued in December 1983, with the requirement that any design changes resulting from i the pre-core hot functional testing (HFT) at Palo Verde Unit 1 that affect CESSAR be submitted to the staff for review and approval. In 1984, the four '

design changes found necessary on the basis of the HTT results from Palo Verde were completely defined. Their potential effect on the CESSAR System 80 de-sign was discussed in a letter from CE dated March 22, 1985, which proposed two revisions to CESSAR to reflect these design changes.

These four design changes have been mentioned above. CE submitted the follow- ,

ing topical reports, which addressed the changes:

(1) CEN-265(v)-P, Rev. 1-P, " Final Report on Palo Verde Unit 1 Resistance Temperature Detector Thermowell," August 1984.

(2) CEN-264(v)-P, " Report on Palo Verde Unit 1 Safety Injection Nozzle Thermal i Liner," January 1984.

(3) CEN-267(v)-P, Rev. 1-P, " Final Report on The Performance Evaluation of the Palo Verde Control Element Assembly Shroud," August 1984.

(4) CEN-271(v)-P, Rev. 1-P, " Final Report on Palo Verde Nuclear Generating Station Reactor Coolant Pumps," August 1984.

The staff's evaluation of these design changes is presented in Section 14 of the Palo Verde SER, Supplement No. 7, dated December 1984. The staff's evaluation of CE's March 22, 1985, submittal, addressing the effects of these design changes on CESSAR, follows.

14.3.1 Reactor Coolant Pump Diffuser and Impeller Repairs CE's letter of March 22, 1985, reaffirmed its previous position that the details of the pump and impeller design affected by the repairs exceed those in CESSAR Section 5.4.1, and, therefore, revisions to CESSAR Section 5.4.1 are not needed.

The staff has reviewed this submittal, finds that the corrective actions taken, which include (1) additional bolting and pretensioning of the bolts fastening i the diffuser to the pump casing, (2) recontouring the diffuser vane tips, and l (3) backfilling and welding repairs of the pump impeller, correspond to a much finer level of detail than the description in CESSAR Section 5.4.1, and agrees that revisions to Section 5.4.3 are unwarranted.

CESSAR SSER 3 14-3 1

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14.3.2 Resistance Temperature Detector (RTD) Thermowell Design Revision '

Failures of the original resistance temperature detector (RTD),thermowell at.

Palo Verde Unit I were caused by wear and high cycle. fatigue, ihich were the result of the resonance of vortex shedding frequencies and the natural fre-quency'of thermowells. In eupplement No. 7 to the Palo Verde SER, the staff concluded that the adequacy of the new" design was demonstrated by analytical results and data collected from several CE loop tests and the full-scale demon-stration test. Because CESSAR does not describe the' detailed design of thermo- ,

wells, CE does not propose a revision to CESSAR. The staff agrees 3 that no re- )

vision is required.

14.3.3 Safety Injection Nozzle Thermal Sleeve Removal This design change entails revisions of.CESSAR Sections 5.4.3.1 and,5.4.3.2.

The staff's evaluation of these revisions is provided in Section 5.4.3.2 of i this supplement.

14.3.4 Control Element Assembly Guide Removal This design change involved revisions in CESSAR Sections 3.7 and 3.9.5. The i staff's evaluation of these revisions is provided in Section 3.9.5 of this i supplement.

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1 15 ACCIDENT AND TRANSIENT ANALYSIS I 15.1 Introduction and Analytical Techniques-1 15.1.1 CESSAR Consistency Review Changes As a result of a review of the technice.1 specifications for Palo Verde Unit 1, CE determined that changes were necessary in CESSAR Chapter 15. By a Isolation Valve Closure Time|letter dated December 5, 1984]], CE submitted changes in the Chapter 15 sequence-of-events tables and supporting text to avoid any inconsistency with the Palo Verde Unit 1 Technical Specification's and CESSAR interface requirements. Since Palo Verde Unit 1 references CESSAR, these changes were first reviewed by the staff as part of this plant's application for an operating license. The staff's safety evaluation appears in Appendix H of Supplement No. 7 to the Palo Verde Units 1, 2, and 3 SEE (NUREG-0857). The changes submitted by CE's letter of December 5, 1984, are acceptable for Palo Verde Units 1, 2, and 3. Subse-quently, the staff reviewed the acceptability of these changes for CESSAR. It concludes that the previous Palo Verde Unit 1 evaluation is applicable to CESSAR and finds these changes acceptable for CESSAR as well.

15.3 Limiting Accidents 15.3.1 Steam Piping Failures Inside and Outside Containment l

Supplement No. 2 to the CESSAR SER contains a detailed safety evaluation of CE's accident analyses involving steam piping failures inside and outside con-Lainment as transmitted by letters dated April 26 and May 10, 1983. These sub-mittals were intended to meet the staff's request for confirmatory small-break analyses performed in accordance with the criteria of Standard Review Plan (SRP) Section 15.1.5 (NUREG-0800). As stated in Supplement No. 2, these analy-ses were found acceptable. In Amendment 9 to CESSAR, Section 15.1.5 is modi-fied to incorporate these analyses, previously transmitted and found acceptable by the staff. The staff has reviewed this submittal, finds it accurately re-flects the approved analyses, and concludes that it is acceptable.

Supplement No. 2 to the CESSAR SER also gives the status of the staff's review and audits of the CE methodology for analyzing steamline breaks. This method-ology is described in Appendix C to CESSAR Chapter 15. The staff has approved the application of this methodology to CESSAR System 80, so that the issue is closed for CESSAR. However, the staff felt that a final approval of CE's gen-eric steamline break methodology would require significant upgrading of Appen-dix C to CESSAR Chapter 15, which would include a quantification of the conser-vatisms inherent in the methodology. To meet this concern, CE submitted, as part of Amendment 9, a revised Appendix C to Chapter 15. The review of this submittal is outside the CESSAR scope. The staff will present the results of its evaluation in a future generic repurt.

CESSAR SSER 3 15-1

15.3.7 Steam Generator Tube Rupture Supplement No. 2 to the CESSAR SER states that CE was required to reanalyze the steam generator tube rupture (SGTR) event assuming the most limiting single failure following the SGTR. In response, in a letter dated May 10, 1983, CE stated that it would perform the SGTR analysis, assuming offsite power is un-available, after a reactor trip. In accordance with the SRP (NUREG-0800),

operator actions would be considered to ensure that the most severe case had been taken into account. The reanalysis would take into account the CE emer-gency procedure guidelines. It would be assumed that the operator used the i atmospheric dump valves (ADVs) an appropriate time after the reactor trip, to )

facilitate the primary system depressurization and cooldown to a set of condi-  !

tions that would not challenge the main steam safety valves, and then the '

affected steam generator would be isolated. CE would assume that an ADV on the affected steam generator was stuck open at the operating position, which would

result in the fastest cooldown rate allowed. Credit would be taken for opera-l tor action 10 minutes after it became apparent that an ADV on the affected steam generator was stuck open. It would be assumed that the operator closed a block valve (safety-related) upstream from the stuck-open ADV and thus limited '

the release. Cycling of the remaining ADVs during the analysis would be allow-ed should it become necessary to prevent overfilling of the steam generators.

The calculated doses shall be within the guidelines of 10 CFR 100.

By a letter dated July 22, 1983, CE submitted a reanalysis of the SGTR event, based on the assumptions described above. The operator actions assumed in this analysis are consistent with the CE emergency procedure guidelines documented in CEN-152. For a double-ended rupture of a steam generator tube, the primary to secondary leak rate exceeds the capacity of the charging pumps. As a re-sult, the pressurizer pressure and level will drop. At 47 seconds, a reactor trip signal is generated because the hot-leg saturation temperature range limit of the core protection calculator is exceeded. The pressurizer empties at approximately 346 seconds. After the pressurizer empties, the upper head of the reactor vessel begins to control primary system pressure. Following reac-l tor trip and with turbine bypass unavailable, a maximum main steam system pres-sure of 1330 psia occurs at 56 seconds. The main steam safety valves cycle three times until the operator takes control of the plant. Following reactor trip and loss of offsite power, the level in the steam generators decreases because of loss of main feedwater. The auxiliary feedwater (AFW) flow initiates to feed the steam generators. At 460 seconds, the operator takes control of the plant and opens one ADV on each steam Generator for plant cooldown. At 2100 seconds, the reactor coolant system (RCS) has been cooled to 550 F. The operator isolates the AFW flow to the affected steam generator, closes the main steam isolation valves (MSIVs) of both steam generators, and attempts to close the ADV on the affected steam generator. The ADV is assumed to stick open and 10 minutes later it is assumed that the operator has closed the block valve upstream of the stuck-open ADV, stopping the blowdown. The operator then ini-tiates an orderly cooldown by using ADVs and AFW flow to the unaffected steam generator. CE's analysis indicates that for the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> following the initiation of the SGTR event, approximately 484,000 lbm of steam are released to the environment through the ADVs, and for the 2- to 8-hour cooldown period, an additional 973,000 lbm of steam are released via the ADVs from both steam generators. The maximum RCS and secondary pressures do not exceed 110% of design pressure throughout the transient.

CESSAR SSER 3 15-2

The auxiliary pressurizer spray (APS) system was assumed operable for plant depressurization following an SGTR event.  !

The staff notes that the CE System 80 SGTR analysis takes credit for the backup pressurizer heaters after 4500 seconds to regain pressurizer pressure control.

These heaters are not safety related but are powered from Class 1E power sup-plies. The heaters are not safety related only in that they are not reismical-ly qualified and they are not environmentally qualified to a post-LCCA contain-ment environment. However, the SGTR should not create a harsh environment in-side containment. Also, the staff does not postulate SGTRs concurrent with the safe shutdown earthquake. Thus, although the backup heaters are not fully i safety related, the use of the heaters in the recovery following an SGTR is  ;

acceptable. However, because the backup heaters are used to mitigate an SGTR accident, the staff requires that CESSAR provide an interface requirement to specify the need of technical specifications for surveillance and operability testing of the backup heaters for CE System 80 plants.

The staff has reviewed CE's SGTR analysis and concludes that the assumptions are reasonably conservative and the results of the analysis are acceptable with respect to peak system pressure during the transient. The analysis also proper-ly reflects the emergency operating guidelines. The results of the staff's evaluation of the radiological consequences of this accident are given in Sec-tion 15.4.5 of this supplement.

The staff requested that CE specify an interface requirement in CESSAR System 80 on the need for safety-related block valves upstream of the ADVs con-sistent with the assumptions in the SGTR analysis. Amendment 9 tc CESSAR in-cludes Appendix 150, which incorporates the analysis initially submitted by CE's letter of July 22, 1983. The staff has reviewed it and finds it consis-  :

I tent with the earlier submittal.

Final resolution of the loop charging valve issue is given in Section 5.4.3 of this supplement. The interface requirement for safety-related ADV block valves was provided by CE's letter dated February 24, 1984. However, the staff needed-additional information. Af ter several discussions with the staff, CE submitted the following revised interface requirements in CESSAR Section 5.1.4 by a letter i dated March 15, 1985:

Item B.2 The steam piping and associated supports from the steam generator up to and including the main steam isolation valves (MSIVs), the ADVs and their associated isolation valves, and any auxiliary steam supply systems up to the isolation valves which connect upstream of the MSIVs shall be seismic Category I and designed to ASME Boiler and Pressure Vessel Code,Section III, Class 2 requirements.

Item I.14 If the isolation valves upstreaL of the ADVs are electrically con-trolled or operated, the valve operator and control systems shall be designed to the same IEEE standards as applied to the ADVs.

CESSAR SSER 3 15-3 u

The staff has reviewed the revised CESSAR interface requirements and has con-cluded that they are acceptable. This interface requirement was incorporated in Amendment 10 to CESSAR. The ADV block valves do not have to be remotely operable because the SGTR analysis in CESSAR assumes manual actions to close the block valve. However, the staff will review the accessibility of the ADV block valves on each referencing applicant's docket if the local manual opera-tor ADV block valves are used in the plant design.

In a letter dated August 3, 1984, CE pointed out that, in the above staff.evalu-ation, the description of the SGTR reanalysis states that credit will be taken for operator action 10 minutes after it becomes apparent that an ADV on the affected steam generator is stuck open, whereas CE's reanalysis takes credit for this operator action 30 minutes after the operator realizes that the ADV is stuck open. The staff addressed this question by a letter dated November 15, 1984. Because the assumption of operator action affects the radiological consequences of the SGTR, this issue is discussed further in Section 15.4.5 of this supplement.

The SGTR analysis assumes that the APS system is used for primary system de-pressurization during the accident mitigation. The acceptability of the results of the SGTR analysis relied on the fact that the systems and components used.

tor. accident mitigation are designed to safety related standards. TM staff has recently identified potential design deficiencies in the APS system water supply system. The detailed staff concerns on this issue are reported in Sec-tion 5.4.3 of this supplement. The staff concluded that staff acceptance of the SGTR analysis will depend on the final resolution of the staff concerns re-garding APS system design addressed in Section 5.4.3 of this supplement. The staff will report the resolution of this issue in a future supplement.

With the exception of the above staff concern, the staff finds that the SGTR reanalysis in Appendix 150 to CESSAR (Amendment 9), together with the amendment to changes involving the ADV block valve interface requirements and the modifi-cations to the loop charging valves discussed in Section 5.4.3 of this supple- 4 ment, is acceptable, provided the analysis of the radiological consequences is also acceptable. This subject is discussed in Section 15.4.5 of this supplement.

The staff concluded that Confirmatory Issue No. 2 can be closed after the final resolution of the staff concerns regarding the design of the APS system and its water supply system addressed in Section 5.4.3 of this supplement.

15.4 Radiological Consequences of Design-Basis Acciden_ts Supplement No. 2 to the CESSAR SER lists four site-related interface require-ments for CESSAR reference plants established by the staff on the basis of the analyses described in the SER and its supplements. One of these interface re-quirements given in Section 15.4 of Supplement No. 2, steam generator tube leakage, is established as 0.1 gpm primary to secondary. By a letter dated August 3, 1984, CE pointed out that the analyses evaluated in Supplement No. 2 show acceptable consequences for steam generator tube leakage of 1.0 gpm and requested that the value listed in the SER be changed to 1.0 gpm to provide consistency with analyses for the worst-case assumptions used in design-basis accidents. The staff has reviewed this request, finds that the radiological l

CESSAR SSER 3 15-4 I

1 consequences of applicable design-basis accidents were evaluated assuming a primary-to-secondary leak rate of 1.0 gpm, and concludes that the change is acceptable. l 15.4.5 Steam Generator Tube Rupture By letter dated July 22, 1983, CE submitted the reanalysis of the SGTR event based on the assumptions described in Section 15.3.7. However, CE used a more conservative analysis than required and assumed that the operator action to close the block valve would not occur for 30 minutes instead of the 10 minutes previously stated. The staff has reviewed the CE analysis and concludes that the radiological assumptions summarized below are consistent with the current licensing guidance in SRP Section 15.6.3 and are acceptable. Using these as-sumptions, CE has estimated the offsite radiological consequences at the ex-clusion area and the low population zone boundaries to be 60 rems (thyroid) and  ;

15 rems (thyroid), respectively.

(1) The primary coolant activity was assumed at the technical specification value of 1.0 pCi/gm dose equivalent iodine-131 (CEI-131).

(2) The secondary coolant activity was assumed at the technical specification limit of 0.1 pCi/gm DEI-131.

(3) A spiking factor of 500 times the normal release rate from the fuel was assumed.

(4) The technical specification leak rate of 1 gpm was assumed for the unaffected steam generator for the duration of the accident.

(5) During the period when the ADV was stuck open, the tubes remained covered I with water; therefore, only the tube leakage that flashes was assumed to be released directly to the atmosphere without any iodine scrubbing (decon-tamination factor = 1).

(6) A partition factor of 100 was assumed between the steam generator water and steam phases to estimate the additional releast: of activity from the steam generator water.

(7) The atmospheric dispersion factor used for the exclusion area boundary in i the staff's analysis was 2,0 x 10 3 sec/m3.

(8) Dilution of primary and secondary system activities caused by the high-pressure safety injection flow and auxiliary feedwater is accounted for in the calculation.

The staff has also performed an independent evaluaticn of the offsite radiologi-cal consequences of the postulated event at the exclusion area boundary. Using assumptions consistent with SRP Section 15.6.3, as summarized above, the staff estimated the potential radiological consequences at the exclusion area boundary to be 76 rems (thyroid) and 1 rem (whole body). Considering the level of detail of the CE calculation with respect to the staff's estimate, the staff believes that the CE and staff values are in reasonable agreement. It, therefore, con-cludes that the CE analysis of an SGTR with a failure of an ADV and isolation l

CESSAR SSER 3 15-5

1 sithin 30 minutes shoulo result in offsite radiological consequences less'than the guideline va,oes of 10 CFR 100. It should be noted that this analysis is limited exclusively to the CESSAR System 80 design and that any change to the  ;

System 80 design (such as size of the ADVs or no block valves for the ADVs) '

could make a significant difference in the estimated offsite radiological con-sequences from the SGTR scenarios described above, and a plant-specific calcu- )

lation may be required for those plants that deviate from the System 80 design.

The staff, therefore, concludes that the radiological consequences of the steam generator tube rupture accident, as given in CE's letter of July 22, 1983, and in CESSAR Amendment 9, Appendix 15D, are acceptable. Confirmatory Issue No. 2 can be closed after the final resolution of the staff concerns regarding the design of the APS system and its water supply system addressed in Section 5.4.3 of this supplement.

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l CESSAR SSER 3 15-6

l 16 TECHNICAL SPECIFICATIONS Supplement No. 2 to the CESSAR SER indicates that Chapter 16, " Technical Specifi-cations," was not explicitly reviewed by the staff. Since then, Amendment 9 to i CESSAR was submitted by CE by a letter dated March 16, 1984. Amendment 9 includes a completely rewritten Chapter 16, which constitutes a set of nuclear swam supply system standard technical specifications applicable to System 80 piants.

The approved standard technical specifications for Combustion Engineering (CE) pressurized-water reactors (NUREG-0212, Revision 2) were developed before System 80 was approved. It was, therefore, clear that to promote the greatest possible homogeneity among the technical specifications of different plants referencing CESSAR, it would be advantageous to have a set of approved standard System 80 technical specifications, even though such specifications would be limited to the CESSAR scope, the nuclear steam supply system. In a letter dated April 26, 1985, CE requested that CESSAR Chapter 16 be reviewed by the staff within the same time frame as the other CESSAR changes. To facilitate the staff's review, CE did a word-by-word comparison of CESSAR Chapter 16 with the approved ,

Palo Verde (the first System 80 plant) technical specifications that pertained to the CESSAR scope. The differences were identified. This effort led to a final version of Chapter 16, which was formally submitted to the staff by a letter dated May 16, 1985.

In a letter dated June 4, 1985, the staff agreed to review the CESSAR technical-specifications, with the goal of issuing a proof-and-review copy of these spec-ifications by the first week of July 1985. As part of its review, the staff considered it necessary to separately compare CE's submittal with the approved Palo Verde technical specifications, as well as to identify the differences be-tween this submittal and NUREG-0212, Revision 3. Furthermore, in the course of the Palo Verde technical specifications review, the differences between these specifications and NUREG-0212, Revision 3, had been identified.

In the course of its review, the staff asked CE a set of questions, which were answered and resulted in the proof-and-review copy of the CESSAR technical specifications that was sent to CE by a letter dated July 3, 1985. To docu-ment this process, however, the staff formally sent these questions to CE by a letter dated July 10, 1985.

CE commented on the proof-and-review copy of the technical specifications by a letter dated July 18, 1985, and formally answered the staff's questions in a letter dated July 31, 1985.

Further discussions with CE resulted in a final draft of the technical specifi-cations being sent to CE on August 14, 1985. CE then certified them and sub-mitted them as Amendment 11 to CESSAR on September 3, 1985.

The staff has reviewed this submittal, finds it identical to the version it had previously approved, and concludes that it is acceptable.

CESSAR SSER 3 16-1

Chapter 16 of CESSAR Amendment 11 constitutes only the nuclear steam supply

. system portion of the standard technical specifications (STS) for CESSAR Sys-tem 80 up to Amendment 11. These technical specifications are labeled CESSAR 80-NSSS-STS. The balance-of plant (80P) portion of the STS has only a reference to either "(B0P)" or "See Applicant's SAR." The BOP technical spec-ifications will be developed using the guidance contained in the Combustion Engineering Standard Technical Specifications. CESSAR 80 NSSS-STS will be re-vised to reflect applicable changes in regulatory requirements at the time they are approved for use in the licensing process.

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CESSAR SSER 3 16-2

07 QUALITY ASSURANCE E 7. 3 Quality Assurance Program 07.3.1 CESSAR Amendments 8 and 9 he staff has reviewed Amendments 8 and 9 to CESSAR to determine whether their ontents affect the conclusions of the evaluation in the CESAR SER regarding E's Quality Assurance Program. The staff concludes that these amendments are

cceptable from the quality assurance standpoint in that they do not degrade he previously accepted Quality Assurance Program.

$7.3.2 CESSAR Inconsistencies he review of the Palo Verde Unit 1 Technical Specifications revealed apparent inconsistencies between these specifications and the Chapter 15 safety analy-es in CESSAR, as well as between CESSAR and CE safety analyses that are the asis for CESSAR. During a' meeting on October 4,1984, CE described to the ptaff the results of its detailed review and proposed actions to correct the inconsistencies. Further, CE stated its intent to conduct an independent au-

$it program to ensure that CESSAR analyses are consistent with design specifi-hations, interface requirements, reasonable engineering judgment, and proposed technical specifications. In a letter dated October 22,19M,. the staff re-guested that CE formally document the corrective measures described at the meet-Ung and its intentions regarding the proposed independent audit program. In a Detter dated November 6, 1984, CE addressed the staff's requests as discussed below:

[1) To correct the inconsistencies, CE was revising Chapter 15 sequence of events tables and other CESSAR chapters as appropriate.

l2) The independent audit of the process used to generate the CESSAR and Palo Verde Unit 1 Technical Specifications was performed by senior engineers not involved in the preparation of the material they were to audit. The  !

results of this audit were to be transmitted to the staff on completion of the program; CE was to immediately inform the staff of any significant finding.

rhe changes to CESSAR that were found to be needed were submitted by a letter lated December 5, 1984. These changes were evaluated and found acceptable as Described in the appropriate sections of this supplement.

iy a letter dated January 11, 1985, CE submitted the report that resulted from

his independent audit, which found, according to CE, no serious technical trrors, no safety concerns, and no systematic patterns of error. Moreover,

,he CE design process was found to be " adequate and conservative."

he staff has reviewed this report, finds that it adequately meets the concerns npressed in its October 22, 1984 letter, and concludes that this issue is losed.

ESSAR SSER 3 17-1

i I

22 TMI-2 REQUIREMENTS j 22.1 Introduction Supplement No. 2 to the CESSAA 'f8 states that Amendment 6 to CESSAR added two related NUREG-0737 items to the ctSSAR scope: (1) I.D.2, " Plant Safety Param-eter Display Console," and (2) III.A.1.2, " Upgrade Emergency Support Facili-ties." When Supplement No. 2 was issued, the staff was still reviewing these two items; hence, these items were not approved for addition to the CESSAR scope. Item I.D.2 is specifically excluded from the CESSAR scope in the Final i Design Approval issued in December 1983, pending completion of the staff re-view. Item I.D.2 is discussed in the next section. No action has been taken on Item III.A.I.2, which is still unreviewed on the CESSAR docket.

By a letter dated March 16, 1984, CE submitted Amendment 9 to CESSAR, which contains changes to Chapter 22, Items II.F.2, " Evaluation of Inadequate Core Cooling Detection Instrumentation"; II.K.2.13, "Effect of High Pressure Injec-tion on Vessel Integrity for Small Break LOCAs With Auxiliary Feedwater Un-available"; II.K.2.17, " Potential for Voiding in the Reactor Coolant System During Transients"; and II.K.3.30, " Revised Small Break LOCA Methods To Show Compliance With 10 CFR 50, Appendix K."

These changes will be discussed in the next section (which also discusses Item II.D.1, " Performance Testing of Boiling Water Reactor and Pressurized Water Reactor Relief and Safety Valves"), which supersedes the evaluation in the CESSAR SER.

22.2 Evaluation i

I.D.2--Plant Safety Parameter Display Console The staff reviewed the description of the safety parameter display system (SPDS) in Amendment 6 to CESSAR, Appendix B, and found the description insuf-ficient to meet its evaluation needs. CE then suggested tnat the staff use the SPDS safety analysis to be provided by San Onofre Nuclear Generating Station, Unit 2 (SONGS 2) as additional information for its review, since SONGS 2 was equipped with a CE-designed SPDS. The safety analysis of the SPDS for San Onofre Units 2 and 3 was submitted to the staff on the San Onofre docket by a letter from M. O. Bedford to G. Knighton, dated November 30, 1984, but was never submitted on the CESSAR docket.

The staff took the position that if the SPDS is included in the CESSAR scope, a generic safety evaluation must be performed for this part of a standard design. '

This safety evaluation thould include several site visits to enable the review-ers to judge the actual installation of this equipment, particularly from a human factors engineering standpoint. Several discussions with the staff con-firmed to CE that the generic SPDS review would involve a significant number of plant-specific aspects; the validity of the conclusions related to the plant- i specific aspects would have to be checked when licensing the next CESSAR plant,  !

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CESSAR SSER 3 22-1 1

.I l

if the SPDS is part of the CESSAR scope. CE then suggested that although it I wishes the SPDS to be part of the CESSAR scope, information to support plant- l specific implementation could only be provided by an applicant referencing CESSAR. The staff agreed that it would be more cost effective to review the CESSAR SPDS once it is implemented in an actual plant. However, for the SPDS to remain in the CESSAR scope, CE was required to provide an interface require-ment clearly stating that providing the additional information needed to per-form the SPDS review would be the responsibility of the applicant referencing CESSAR. By a letter dated August 15, 1985, CE submitted this interface require-ment.

i The staff finds this interface requirement acceptable and concludes that the '

SPDS can be included in the CESSAR scope, but that a complete review of this equipment, because it involves plant specific information, must be deferred until a license application is reviewed for a plant referencing CESSAR.

II.D.1--Performance Testing of Boiling Water Reactor and Pressurized Water Reactor Relief and Safety Valves When performing its generic review of the need for rapid depressurization capa-bility at CE-designed plants, the staff found that additional information was needed as a result of tests conducted by the Electric Power Research Institute (EPRI) on full-sized PWR primary system safety valves, in response to NUREG-0737, Item II.D.1. This request for additional information was sent to CE in a letter dated April 24, 1984. Specifically, the EPRI tests indicated f that the manufacturers cannot obtain a complete understanding of valve perfor-I mance capability without at least some full scale test or operational experi-ence. The staff's concern was related to-the CESSAR shutdown cooling system (SDCS) relief valves, which provide low-temperature overpressure protection for the reactor vessel and overprespJre protection of the SDCS itself. These valves are larger by at least an Weer of magnitude than those in other PWR plants. At the time these valves were manufactured, the ASME Code allowed such valves to be capacity certified solaly on the basis of calculations performed by the manufacturer.

The staff, in its April 24, 1984, letter, required confirmation that these large SDCS relief valves can provide the required pressure relief capacity and subsequently reclose under all fluid conditions to which they could be exposed.

This confirmation could consist of test or operational data on valves in non-nuclear service that substantiate valve performance for sizes comparable to that of the SDCS relief valve.

CE provided additional informati.on in a letter dated September 7, 1984. This information, despite a meeting with the staff en September 18, 1984, failed to assure the staff that the valves would perform as designed. Another meeting on November 9, 1984, which also included representatives of Arizona Public Service (operator Sf Palo Verde Unit 1) and of the valve manufacturers, provided adequate information, which was formally submitted by a letter from CE dated December 13, 1984.

The staff's safety evaluation for Palo Verde Unit 1 is docuniented in Supple-ment No. 7 to the Palo Ver % SER (NUREG-0857). This evaluation concluded that CESSiR 55ER 3 22-2

l I

the information provided at the November 9,1984 meeting gives acceptable as-surance that the Palo Verde Units 1, 2, and 3 SDCS relief valves can ade-quately perform their intended safety function to relieve the system overpres-sure and subsequently reclose. It has, therefore, been shown to the staff's satisfaction that valves meeting the CESSAR interface requirements can j adequately perform their intended safety function. This issue is, therefore, closed on the CESSAR docket. However, because the specifics of valve operation i are outside the CESSAR scope, the operability of the valves selected for future plants referencing CESSAR will have to be justified on a case-bycase basis.

II.F.2--Instrumentation for Detection of Inadequate Core Cooling Amendment 9 to CESSAR includes a revised Table 3-1, "Evaltation of Inadequate Core Cooling (ICC) Detection Instrumentation to Documentation Requirements of NUREG-0737 Item II.F.2," in Appendix B. The staff reviewed this table and found it acceptable, but suggested that the sentence "The CE Owners Group has developed generic emergency procedure guidelines for addressing ICC" be moved from Item 8 to Item 7, where it would be more appropriate. For a discussion of the acceptability of these emergency procedure guidelines, see Section 4.4.11 of this supplement.

By a letter dated May 31, 1985, CE submitted a modified Table 3-1, Appendix B, that follows the staff's suggestion. This modified table is acceptable.  ;

II.K.2.13 Effect of High-Pressure Injection on Vessel Integrity for Small-Break Loss-of-Coolant Accident With No Auxiliary Feedwater l Supplement No. 2 to the CESSAR SER indicates that, in a letter dated March 27, 1983, CE stated that Topical Report CEN-189, " Evaluation of Pressurized Thermal Shock Effects Due to Small-Break LOCAs With Loss of feedwater for the Combus-tion Engineering NSSS," was applicable to CESSAR and fulfilled CE's commitment for this item. Amendment 9 to CESSAR includes a modified response to Item II.K.2.13 to incorporate the above information into CESSAR. The staff finds this change acceptable.

Supplement No. 2 to the CESSAR SER also states that the staff's review of CEN-189 will be covered in the review of Unresolved Safety Issue A-49. Since i Supplement No. 2 was issued, SECY 85-60, " Final Pressurized Thermal Shock '

Rule," was published on February 20, 1985, and adopted by the Commission on July 23,1985. The corresponding amendments to 10 CFR 50 became effective on that date. The staff's review of CEN-189 can be found by consulting SECY 85-60 and the references included.

II.K.2.17--Potential for Voiding in the Reactor Coolent System During Transients Supplement No. 2 to the CESSAR SER indicates that, in a letter dated March 27, 1983, CE stated that Topical Report CEN-199, " Effects of Vessel Head Voiding During Transients and Accidents in CE NSSS," was applicable to CESSAR and ful-filled CE's commitment for this item. Amendment 9 to CESSAR includes a modi-fied response to Item II.K.2.17 to incorporate this position into CESSAR.

CESSAR SSER 3 22-3

The staff has reviewed this change and finds it acceptable. Moreover, the staff has completed its review of CEN-199 and found the results of that study acceptable.

II.K.3.30--Revised Small-Break LOCA Methods To Show Compliance With 10 CFR Part 50, Appendix K Supplement No. 2 to the CESSAR SER indicates that, by a letter dated March 27, 1982, CE stated that Topical Report CEN-203, Revision 1, " Response to NRC Action Plan Item II.K.3.30--Justification of Small-Break LOCA Methods," was applicable to CESSAR and fulfilled CE's commitment for this item. The staff ,

1

.has finished its review of this topical report and finds it meets the require- I ments of II.K.3.30. The staff concludes that this item has been resolved.

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CESSAR SSER 3 22-4

APPENDIX A CONTINUATION OF CHRONOLOGY OF CESSAR REVIEW December 20, 1987 Letter from Northeast Utilities (CE Owner Group repre-sentative) transmitting CEN-227, " Summary Report on Operability of Pressurizer Safety Valves in CE Design Plants."

May 10, 1983 Letter from CE requesting approval of CENPD-255-A, Revision 3 for referencing.

Hay 25, 1983 Meeting with CE concerning reactor shutdown cooling i analysis. '

June 30, 1983 Letter to CE approving requests for withholding dated j April 11 and April 26, 1983. ]

l June 30, 1983 Letter to CE approving requests for withholding dated February 28, April 26, and May 10, 1983.

July 19, 1983 Letter to CE forwarding request for additional informa-tion on core protection calculator and statistical com-bination of uncertainty issues.

July 22, 1983 Letter from CE forwarding reanalysis of steam generator tube rupture event.

July 29, 1983 Letter to CE requesting formal justification for reduction of core bypass flow valve by Amendment 5.

August 3, 1983 Letter from CE forwarding information on Systen 80 bypass flow.

August 11, 1983 Letter from CE forwarding proposed revised pressurizer relief valve blowdown analysis.

August 12, 1983 Letter from CE forwarding " Computer Simulation of l Natural Circulation Cooldown."

l l August 19, 1983 Letter from CE forwarding enclosures to " Statistical Combination of Uncertainties, Part III," Revision 1.

August 30, 1983 Letter from CE forwarding proposed text changes concerning hot no-flow scram insertion testing.

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CESSAR SSER 3 1 Appendix A

August 31, 1983 Letter from CE forwarding additional.information on application of statistical combination-of-uncertainties methodology.

August 31, 1983 Letter to CE requesting detailed assessment of relationship between CESSAR and any potential design changes resulting from CE equipment that wt found damaged during hot functional . testing at Palo Verde.

September 8, 1983 Letter to CE regarding the acceptance of CENPD-255-A, Revision 3 for referencing.  !

l l September 9, 1983 Letter from CE discussing the effect of problems with i reactor coolant pumps, resistance temperature detector thermowells, safety injection nozzle thermal sleeves, and control element assembly identified during hot functional testing at Palo Verde.

September 14, 1983 Letter to CE regarding applicable fees for review of Final Design Approval application.

September 20, 1983 Letter frcm CE forwarding revision to technical speci-fications for. nuclear steam supply system.  ;

September 30, 1983 Letter from CE discussing referenceability of CESSAR Final Design Approval.

September 30, 1983 Letter to CE forwarding Supplement No. 2 to CESSAR SER. .

l l October 26, 1983 Meeting with CE to discuss " Basis of Design of Plant l Without Pipe Whip R0straints for Reactor Coolant System Main Leop Piping."

November 1, 1983 Letter to CE stating that information dated August 1983 on CESSAR core bypass flow will be withheld from public disclosure.

November 1, 1983 Letter to CE approving August 19 and 31, 1983 requests for withholding from public disclosure information on review of use of statistical-combination-of-uncertainties methodology.

1 November 4, 1983 Meeting with CE to discuss plant design without pipe whip restraints.

November 9, 1983 Letter from CE commenting on NRC evaluation of need for rapid depressurization capability for CE plants.

l December 20, 1983 Letter from CE stating that reactor vessel lower key horizontal support load-limiting device test results in support of Palo Verde operating license schedule will not be available until first quarter of 1984.

CESSAR SSER 3 2 Appendix A

December 21, 1983 Letter to CE forwarding Final Design Approval (FDA-2) for CE System 80 nuclear steam supply system.

December 23, 1983 Letter from CE forwarding responses to NRC comments on

" Leak Before-Break Evaluation of Main Loop Piping of C-E Reactor Coolant System."

February 16, 1984 Letter from CE forwarding " Comprehensive Vibration Assessment Program for Palo Verde, Unit 1 (System 80 Prototype)."

February 24, 1984 Letter from CE forwarding interface requirements regarding block valves on atmospheric dump valves.

March 16, 1984 Letter from CE forwarding Amendment 9 to CESSAR for review.

March 22, 1984 Letter from CE forwarding interface requirement regarding shutdown cooling analysis of reactor coolant gas vent system.

March 22, 1984 Letter to CE forwarding safety evaluation of statistical combination of uncertainties.

March 26, 1984 Letter to CE forwarding comments on rod drop data from recent startup testing. ,

i March 27, 1984 Letter to CE forwarding request for additional information  !

regarding single-failure vulnerabilities in auxiliary pressure spray design.

April 3, 1984 Letter to CE forwarding SER supporting use of CESEC-III in calculating Final Safety Analysis Report (FSAR)

Chapter 15 events (excluding anticipated transients I

without scram).

April 24, 1984 Letter to CE requesting additional information to address concerns of Electric Power Research Institute tests on full-size PWR primary system safety valves.

May 7, 1984 Letter to CE advising of review of costs incurred in processing Final Design Approval.

May 25, 1984 Letter from CE stating that information on operability of shutdown cooling system relief valves will be provided by August 1984.

August 3, 1984 Letter from CE updating items on steam generator tube rupture, radiological consequences of design-basis accident, and shutdown cooling.

August 3, 1984 Letter from CE requesting information on potential single failure in auxiliary pressurizer spray system.

CESSAR SSER 3 3 Appendix A

September 7, 1984 Letter from CE forwarding information on operability of shutdown cooling system relief valves in response to request for additional information.

September 18, 1984 Meeting with CE to discuss resolution of remaining confirmatory issues.

September 18, 1984 Letter from CE regarding possible failure of auxiliary pressurizer spray system because of sticking valve. l j

September 28, 1984 Letter from CE stating that requested information on i inadequate core cooling instrumentation will be submitted by December 1984.

October 10, 1984 Meeting with CE to discuss resolution of remaining confirmatory issues.

October 11, 1984 Letter to CE forwarding safety evaluation regarding removal of pipe whip restraints.

October 16, 1984 Letter from CE forwarding minor modifications to CESSAR Chapter 14 to facilitate testing procedure.

October 22, 1984 Letter from CE forwarding consistency review changes to  ;

CESSAR and affected marked-up pages of FSAR.

i October 22, 1984 Letter to CE requesting description of corrective '

measures to CESSAR System 80 design within 10 days.

November 6, 1984 Letter from CE responding to request for description of  ;

measures taken to address inconsistencies between i Palo Verde Technical Specifications and CESSAR.

November 9, 1984 Meeting with CE to discuss operability of shutdown cooling system relief valves of System 80 design.

November 15, 1984 Letter to CE regarding changes requested in August 3, 1984 letter.

November 26, 1984 Letter to CE requesting additional information on initial plant test program description.

November 27, 1984 Letter to CE recommending modifications to severe-accident policy statement.

November 28, 1984 Meeting with CE to discuss shutdown cooling system relief valves for System 80.

December 5, 1984 Letter to CE forwarding changes regarding reduced high pressure safety injection pump flow.

December 5, 1984 Letter to CE forwarding for review proposed changes to startup testing.

CESSAR SSER 3 4 Appendix A

December 5, 1984 Letter to CE forwarding marked-up FSAR pages reflecting i review of technical specifications for first System 80 plant.

December 5, 1984 Letter from CE forwarding CESSAR changes regarding reduced HPSI flow.

December 13, 1984 Letter from CE forwarding " Summary Report on Design Basis of Shutdown Cooling System Relief Valves for CESSAR System 80" and " Summary Report on Operability of Shutdown Looling System Relief Valves for Palo Verde Units 1, 2, and 3."

December 28, 1984 Letter to CE regarding recommended modifications to severe-accident policy statement.

January 11, 1985 Letter from CE forwarding "Palo Verde Technical Audit" l of safety analysis and applicable CESSAR analysts.

January 15, 1985 Meeting with CE to discuss proposed CESSAR changes.

January 31, 1985 Letter to CE transmitting Generic Letter 85-01, " Fire Protection Policy Steering Committee Report."

February 7, 1985 Meeting with CE to discuss changes to be included in amendment to Final Design Approval 2.

February 8, 1985 Letter to CE forwarding request for additional informa-tion on single-failure vulnerabilities of auxiliary pressurizer spray system.

February 22, 1985 Letter from CE forwarding list of submittals to close out Final Design Approval on CESSAR issues.

March 7, 1985 Letter from CE forwarding comprehensive vibration assessment program for Palo Verde. 4 I

March 15, 1985 Letter from CE forwarding revised interface  !

requirements on atmospheric dump block valves.

March 22, 1985 Letter from CE forwarding revisions to 1984 reports on removal of safety injection nozzle thermal sleeve and control element assembly guide tube brackets. Lists corrective actions for reactor coolant pumps and temperature detector thermowell.

4 March 22, 1985 Letter from CE discussing single-failure vulnerability of auxiliary pressurizer spray. i l

March 25, 1985 Letter to CE requesting information on water chemistry l controls. ]

March 26, 1985 Letter to CE providing status of review of confirmatory issues.

]

1 CESSAR SSER 3 5 Appendix A

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April 2, 1985 Letter from CE informing staff of transportation error j in review of nozzle elevation of pressurizer safety:

valve.

-April 25, 1985 Letter from CE forwarding comparison of NRC secondary' water chemistry guidance with Palo Verde /CESSAR proposed water chemistry.

April 26, 1985 Letter from CE withdrawing information on proposed-changes to reduce minimum required low pressure safety injection and high pressure safety injection flows.

April 26, 1985 Letter.from CE requesting finalization of. Chapter 16, I

" Technical Specifications" and meeting to discuss scope of review.

April 30, 1985 Letter from CE responding to request for additional ~

i.nformation on modified. acceptance criteria on main and emergency-feedwater. system test.

May 16, 1985 Letter from CE forwarding final version of Chapter 16, )

" Technical Specifications."

May 21, 1985 Meeting with CE to discuss review of Standard Technical 1 Specifications for System 80 design.

May 31, 1985 Letter from CE. forwarding CESSAR clarifications on safety injection system delay time.

June 4, 1985 Letter from CE forwarding results of analysis of increased blowdown for pressurizer safety valves. j June 4, 1985 Letter to CE forwarding review'and safety evaluation of proposed changes and issues identified in CESSAR SER.

July 3, 1985 Letter to CE forwarding proof-and-review CESSAR.techni-cal specifications for comment.

July 10,1985 Letter to CE forwarding questions on proof-and-review CESSAR technical specifications.

July 18,1985 Letter from CE forwarding comments on proof-and-review technical specifications.

July 31,1985 Letter from CE forwarding responses to questions on technical specifications covering reactor protective systems trip setpoint limits.

August 2, 1985 Letter from CE requesting amendment to FDA-2 to close out remaining issues.

August 14, 1985 Letter to CE forwarding final draft of Standard Technical Specifications for review.

CESSAR SSER 3 6 Appendix A

i l

August 15, 1985 Letter from CE informing that requirement will be added for safety parameter display system.

August 30, 1985 Letter from CE regarding application for amendment to CESSAR Final Design Approval.

l September 3, 1985 Letter from CE submitting Amendment 11 to CESSAR.

October 29, 1985 Letter to CE regarding auxiliary pressurizer spray system and supporting systems.

November 21, 1985 Letter from CE forwarding Amendment 10 to CESSAR.

December 20, 1985 Letter from CE transmitting corrected final draft of Standard Technical Specifications.

January 17, 1986 Letter to CE listing types of applications under 6-month j billing process.

February 4, 1986 Letter from CE transmitting Amendment 11 to CESSAR.

February 19, 1986 Letter from CE regarding closeout of TMI Items II.K.3.30 J and II.K.3.31. j l

February 6, 1987 Letter from CE concerning auxiliary pressurizer spray i system.

February 27, 1987 Letter from CE regarding ATWS, 10 CFR 50.62.

March 24, 1987 Letter to CE regarding auxiliary pressurizer spray i system.

March 27, 1987 Letter to CE regarding steam generator turbine degra-dation problem.

April 10, 1987 Letter from CE regarding steam generator tube vibration i problems.  !

l July 13, 1987 Letter from CE regarding steam generator tube vibration l problems.

l l

l CESSAR SSER 3 7 Appendix A

]

APPENDIX B BIBLIOGRAPHY Bedford, M. O., Southern California Edison Co., letter to G. Knighton, NRC,

" Safety Analysis of SPDS for San Onofre Units 2 and 3," November 30, 1984. ,

CE Owners Group, " Summary Report on Operability of Pressurizer Safety Valves in CE-Designed Plants," 1982. ,

CE Owners Group, letter dated November 26, 1984,

Subject:

"CEN-152 Combustion Engineering Emergency Procedures Guidelines."

Combustion Engineering, Inc., Topical Report CN-152, Rev. 2, " Combustion Engineering Emergency Procedures Guidelines," May 1984.

-- , Topical Report CEN-189, " Evaluation of Pressurized Thermal Shock Effects Due to Small Break LOCAs With Loss of Feedwater for the Combustion Engineering j l

NSS," January ]982.

-- , Topical Report CEN-199, " Effects of Vessel Head Voiding During Transients 1 and Accidents in CE NSSS," March 1982. 1

-- , Topical Report CEN-203, Rev. 1, " Response to NRC Action Plan Item II.K.3.30 -

Justification of Small Break LOCA Methods," April 1982.

-- , Topical Report CEN-227, " Summary Report on the Operability of Pressurizer  !

Safety Relief Valves in CE Designed Plants," December 1932.

-- , Topical Report CEN-264(v)-P, " Report on Palo Verde Unit 1 Safety Injection  ;

Nozzle Thermal Liner," January 1984.

-- , Topical Report CEN-265(v)-P, Rev.1-P, " Final Report on Palo Verde Unit 1 Resistance Temperature Detector Thermowell," January 1984.

-- , Topical Report CEN-267(v)-P, Rev. 1-P, " Final Report on the Performance Evaluation of the Palo Verde Control Element Assembly Shroud," August 1984.

-- , Topical Report CEN-271(v)-P, Rev. 1-P, " Final Report on Palo Verde Nuclear I Generating Station Reactor Coolant Pumps," August 1984.

-- , Topical Report CENPD-255-A, Revision 3, " Environmental Qualification of 4 Class 1E Electrical Equipment," October 1985.

Electric Power Research Institute, EPRI NP-20704-SR, "PWR Secondary Water Chemis-try Guidelines," October 1982.

CESSAR SSER 3 1 Appendix B l

l Institute of Electrical and Electronics Engineers, Standard 279, " Criteria for Protection Systems for Nuclear Power Generating Stations," 1971.

-- , Standard 323, " Qualifying Class 1E Equipment for Nuclear Power Generating Stations," 1974.

]

-- , Standard 384, " Standard Criteria for Independence of Class 1E Equipment I and Circuits," 1971.

]

Lawrence Livermore National Laboratory, UCRL-86249, " Failure Probability of PWR Reactor Coolant Loop Piping," by T. Lo, H. H. Woo, G. S. Holman, and C. K.

Chou, February 1984 (preprint of paper intended .for publication).

U.S. Nuclear Regulatory Commission, Generic Letter 84-04 to all PWR Licensees, Construction Permit Holders, and Applicants, " Safety Evaluation of Westinghouse Topical Reports Dealing With Elimination of Postulated Breaks in PWR Primary Main Loops," February 1,1934.

-- , NUREG-0212, Rev. 2, " Standard Technical Specifications for Combustion En-gineering Pressurized Water Reactors," December 1980.

-- , NUREG-0737, " Clarification of TMI Action Plan Requirements," November 1980.

-- , NUREG-0800, " Standard Review Plan for the Review of Safety Analysis Re-ports for Nuclear Power Plants-LWR Edition," July 1981 (contains branch techni-cal positions).

-- , NUREG-0857, " Safety Evaluation Report Related to the Operation of Palo Verde Nuclear Generating Station, Units 1, 2, and 3," Supplement No. 6, October 1984; Supplement No. 7, December 1984; Supplement No. 9, December 1985.

-- , NUREG-1044, " Evaluation of the Need for a Rapid Depressurization Capabil-ity for Combustion Engineering Plant," December 1984.

-- , SECY-84-134, " Power Operation Relief Valves for Combustion Engineering Plants," March 23, 1984.  ;

-- , SECY 85-60, " Final Pressurized Thermal Shock Rule," February 20, 1985.

Westinghouse Electric Corp., Letter Report NS-EPR-2519, from E. P. Rahe to D. G. Eisenhut, NRC, " Westinghouse Response to Questions and Comments Raised by Members of ACRS Subcommittee on Metal Components During the Westir,ghouse Pre-sentation on September 25,1981 (Westinghouse Class 2 proprietary).

-- , Topical Report WCAP-9558, Rev. 2, " Mechanistic Fracture Evaluation of Re-actor Coolant Piping Containing a Postulated Circumferential Throughwall Crack,"

May 1981 (Westinghouse Class 2 proprietary).

-- , Topical Report WCAP-9787, " Tensile and Toughness Properties of Primary Pi-ping Weld Metal for Use in Mechanistic Fracture Evaluation," May 1981 (Westing-house Class 2 proprietary).

CESSAR SSER 3 2 Appendix B

APPENDIX D ABBREVIATIONS ADV atmospheric dump valve AFW auxiliary feedwater ANPP Arizona Nuclear Power Project APS auxiliary pressurizer spray ASME American Society of Mechanical Engineers -

ATWS anticipated transient without scram B0P balance of plant BTP Branch Technical Position CE Combustion Engineering CEA control element assembly CESSAR Combustion Engineering Standard Safety Analysis Report CFR Code of Federal Regulations COLSS core operating limits supervisory system CP construction permit CPC core protection calculator i

DEGB double-ended guillotine break DEI dose-equivalent iodine DNBR departure from nucleate boiling ratio  !

EPG Emergency Procedure Guideline EPRI Electric Power Research Institute FDA Fina' Design Approval FMLB feedwater line break GDC General Design Criteri(on)(a)

HFT hot functional testing IEEE Institute of Electrical and Electronics Engineers LLNL Lawrence Livermore National Laboratory LOOP loss of offsite power MDNBR minimum departure from nucleate boiling ratio MSIV main steam isolation valve MSSV main steam safety valve NRC U.S. Nuclear Regulatory Commission NSSS nuclear steam supply system CESSAR SSER 3 1 Appendix 0 l

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l PwR pressurized-water reactor RCGVS reactor coolant gas vent system a RTD resistance temperature detector 3 >

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RVUH reactor vessel upper head i 'g 4 i i SCST spray chemical storage tank  ! <

%RU statistical combination of uncertainties '

shutdown cooling system  !

SCCS ,

Sa Safety Evaluation Report c xi SGTR stem ,4enerator tube rupture >

SIAS nWt.y inje.: tion actuation signal SPDS (Muty parameter display system '

SRP Standard Review Plan SSE safe shutdown eart,hquake ,

USI borb olved Safety Issua '\

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a APPENDIX E PRINCIPAL CONTRIBUTORS s

Name Area G. Vissing Project Management H. Balukjian Core Pe forn,ance R. Becker Test Irogram J. Donohew Technical Specifications Y. Hsii Core Perforn;ance J. Huang Containment Systems C. Y. Liang Reactor Systems W. Regan Tesi. Program H. Shaw Equipment. Qualification P. Shemanski Equipment Qualification J. Spraul Quality Assurance R. Stevens Instrumentation and Control J. Wing Chemical Engineering K. Wichman Materials Engineering l

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CESSAR SSER 3 1 Appendix E e

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APPEM)IX J i

SAFETY EVALUATION REPORT ON COM8USTION ENGINEERING EQUIPMENT QllALIFICATION DOCUMENTATION CENPD-255-A, REVISION 3, AND AMEN 0 MENT 9 TO CESSAE CHAPTER 3 ENVIRONMENTAL QUALIFICATION OF CLASS 1E ELECTRICAL EQUIPMENT 1

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1 INTR 000CTION........................................................ 1 2 BACKGROUND.......................................... ............... 1 2.1 Purpose........................................................ 1 2.2 Scope.......................................................... 2 2.3 General Summary of CENPD-255-A, Revision 3..................... 2 2.4 Summary....................................................... 2 3 EVALUATION OF METHODOLOGY DOCUMENTS................................. 4 3.1 Class IE Safety-Related Equipment Covered...................... 4 3.2 Service Conditions............................................. 4 3.3 Temperature, Pressure, and Humidity Conditions Inside Containment.................................................... 4 3.4 Temperature, Pressure, and Humidity Conditions Outside Containment............................. ........... .......... 5 3.5 Submergence.................................................... 5 3.6 Chemical Spray................................................. 5 3.7 Aging--Equipment in Harsh and Non-Harsh Environments........... 5 3.8 Radiation--Equipment in Harsh and Non-Harsh Environments....... 6 a 3.9 Test Sequence.................................................. 7 3.10 Margin......................................................... 7 4 CONCLUSIONS......................................................... 8 s

I s

s CESSAR SSER 3 iii Appendix J 1

1

1 INTRODUCTION Equipment that performs a necessary safety function must be demonstrated to be capable of maintaining functional operability under all service conditions postulated to occur during its installed life for the time it is required to operate. This requirement, which is embodied in General Design Criteria (GDC) 1 and 4 of Appendix A to 10 CFR 50 and Sections III, XI, and XVI.I of Appendix B to 10 CFR 50, are applicable to equipment located inside as well as outside containment. More detailed requirements and guidance relating to the methods and procedures for demonstrating this capability have been set forth in 20 CFR 50.49, " Environmental Qualification of Electric Equipment Important to Safety for Nuclear Power Plants," and NUREG-0588, " Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment," which sup-plements IEEE Standard 323-1974 and various NRC regulatory guides and industry standards.

2 BACKGh0VND NUREG-0588 was issued in December 1979 to promote a more orderly and systematic implementation of equipment qualification programs by industry and to provide guidance to the NRC staff for its use in ongoing licensing reviews. The posi-tions in this report provide guidance on (1) how to establish environmental service conditions, (2) how to select methods that are considered appropriate for qualifying equipment in different areas of the plant, and (3) other areas such as margins, aging, and documentation.

In February 1980, the NRC requested certain near-term operating license (OL) applicants to review and evaluate the environmental qualification documentation for each item of safety-related electrical equipment and to identify the degree to which their qualification programs comply with the staff positions in NUREG-0588. IE Bulletin 79-01B, " Environmental Qualification of Class 1E Equipment,"

issued on January 14, 1980, and its supplements dated February 29, September 30, and October 24, 1980, established environmental qualification requirements for operating reactors. This bulletin and its supplements were provided to OL ap-plicants for consideration in their reviews.

A final rule on environmental qualification of electric equipment important to safety for nuclear power plants became effective on February 22, 1983. This rule, 10 CFR 50.49, specifies the requirements to be met for demonstrating the environmental qualification of electrical equipment important to safety located in a harsh environment.

The qualification requirements for mechanical equipment are principally con-tained in Appendices A and B to 10 CFR 50. The qualification methods defined in NUREG-0588 can also be applied to mechanical equipment.

2.1 Purpose This appendix evaluates the adequacy of the CE Equipment Qualification Program, as described in CE Topical Report CENPD-255-A (Revision 3), " Environmental CESSAR SSER 3 1 Appendix J

Qualification of Class 1E Electrical Equipment," October 1985, and Amend-ment 9 to Chapter 3 of CESSAR (Section 3.11).

This evaluation will ensure that a general methodology exists and can be applied to specific CE-supplied safety-related electrical equipment so that the require-ments identified previously are met. The staff position on any open items identi-fied in Appendix I of Supplement No. 2 is provided.in this appendix.

2.2 Scope l

)

The scope of this appendix is limited to an evaluation of environmental qualifi- l cation methodologies and their potential application to specific equipment items.

No plant-specific information is presented, so margin adequacy is not evaluated.

2.3 General Summary of CENPD-255-A, Revision 3 l

CENPD-255-A (Revision 3) and Amendment 9 to Chapter 3 of CESSAR (Section 3.11) 1 deteribe the methods used to qualify Class 1E CE-supplied electrical equipment and include equipment in harsh and non-harsh environments. These documents cover the following general topics.

The scope of supply of equipment for a generic plant design is listed. This equipment is then assigned a typical environment depending on its ' plant loca-tion. Qualification procedures for equipment in a harsh environment are out-lined, with testing according to type (type testing) of age-conditioned equip- l ment as the method selected. The test sequence is that recommended by IEEE i Standard 323-1974.

1 Qualification of equipment in a non-harsh environment addresses normal and abnormal environments as well as seismic events. An aging analysis, which may result in the performance of accelerated aging, is conducted before testing.

As with items in a harsh environment, typical environments are postulated.

Other material related to qualification is presented, including treatment of radiation degradation, Arrhenius methodology assumptions and conservatism, thermal analysis methods, and operating time derivation techniques.

I 2.4 Summary This appendix gives the results of the staff's review of the CE Equipment Quali-fication Program as outlined and reported in CENPD-255-A (Revision 3) and Amend-ment 9 to Chapter 3 of CESSAR (Section 3.11). The conformance of the CE quali-fication program to current NRC requirements determines its usefulness to licensees or applicants in establishing qualification adequacy on an individual plant basis. The review process consisted of formal and informal rounds of questions, meetings with CE and NRC representatives during which industry and NRC positions were discussed, report modifications as a result of these dis-cussions, and a final approval of specific subtopics.

The staff concludes that CENPD-255-A (Revision 3) and Amendment 9 to Chapter 3 of CESSAR (Section 3.11) describe an adequate qualification program that con-forms to present NRC requirements.

CESSAR SSER 3 2 Appendix J

The staff concludes that the generic approach to environmental qualification of safety-related electrical equipment in CENPD-255-A (Revision 3) and Amendment 9 to Chapter 3 of CESSAR (Section 3.11) is acceptable and meets the requirements of 10 CFR 50.49 as well as NUREG-0800 and its associated standards.

3 EVALUATION OF METHODOLOGY DOCUMENTS The evaluation of safety-related electrical equipment was based on the following criteria.

Harsh Environment The criteria in NUREG-0588, Category 1, as defined in 10 CFR 50.49, formed the basis for the staff's evaluation of the acceptability of electrical equipment in a harsh environment. NUREG-0588 is supplemented by IEEE Standard 323-1974, augmented by Regulatory Guide 1.89. NUREG-0800 was also used as a review guide. 5 Non-Harsh Environment The criteria for environmental qualification of all electrical equipment in a non-harsh environment are the following.

The " design / purchase" specifications for each equipment item shall give the functional requirements for its specific environmental zone during normal and abnormal environmental conditions. A well-supported maintenance / surveillance program shall be established to ensure that equipment that meets the design /

purchase specifications is qualified for the designed life.

The maintenance / surveillance program data and records shall be reviewed by the licensees periodically (18 months or less) to ensure that the equipment has not suf fered thermal and cyclic degradation triggered by abnormal environmental conditions and normal wear because of its service condition. Engineering judg-ment shall be used to modify the replacement program and/or replace the equip-ment as deemed necessary.

These criteria are expanded in NUREG-0800, Section 3.11.

3.1 Class 1E Safety-Related Equipment Covered CENPD-255-A (Revision 3) lists typical safety-related electrical equipment that will be supplied by CE. Every plant referencing CENPD-265-A (Revision 3) will have a specific list of CE-supplied equipment. This list can be found in the applicant's Safety Analysis Report (SAR).

Appendix A of CENPD-255-A (Revision 3) lists and categorizes CE-supplied equip-ment required to mitigate a design-basis event or to obtain a safe shutdown. A general listing, covering several variations of supply scope and building de-sign, cannot be complete. As a result, each applicant must ascertain, through the application of appropriate interface criteria, that the complete listing of nuclear steam supply system (NSSS) scope of supply and balance-of plant items identifies all CE equipment associated with systems that are important to safety.

The individual applicant must also verify that the environmental and seismic conditions and locations of equipment are correct on a plant-specific basis.

CESSAR SSER 3 3 Appendix J

3.2 Service Conditions NUREG-0588 defines the methods for determining the environmental conditions associated with loss-of-coolont accidents (LOCAs) or high-energy line breaks (HELBs) inside and outside containment. This document provides the option of establishing a bounding pressure and temperature condition tased on plant-specific analysis or on general profiles. Since CENPD-255-A (Revision 3) is a i generic document, typical profiles have been developed that will envelope most I

of the specific plant's expected environmental conditions. Appendix B of CENPD-255-A (Revision 3) contains these typical profiles for inside as well as outside containment. These profiles, subject to the ccnditions specified in the conclusions of this appendix (Appendix J), are regarded as adequate for the intended purpose. In addition, assumed normal and abnormal operating condi-tions are supplied for all areas where Class 1E equipment is used. However, no judgment is intended on the validity of the typical profiles. Each specific plant calculates its normal, abnormal, and accident environmental profile, and must compare those profiles with equipment-specific tested profiles to ascer-tain qualification.

3.3 Temperature, Pressure, and Humidity Conditions Inside Containment Environmental conditions for normal and abnormal operation inside containment are specified.

Typical LOCA and main steam line break (MSLB) profiles are furnished, with a peak LOCA temperature of 350 F and a peak MSLB temperature of 370 F. Peak pres-sure in both cases is 60 psig, with superheat conditions postulated for a maxi-mum of 10 minutes followed by saturated cor.ditions. These postulated accident condition profiles are combined in Figure B-7 of Appendix B to CENPD-255-A (Revision 3) and provide a typical combined LOCA/MSLB test envelope. This test profile incorporates the recommended IEEE Standard 323-1974 double peak at 385 F and 66 psig. Temperature and pressure margins are adequate. A tempera-ture profile utilizing the saturation temperature associated with peak project-ed containment pressure is included. The staf f's conclusion is that Figuro B-7, in conjunction with an approved thermal analysis technique entitled "ther-mal equivalence," is acceptable with the following restrictions.

The general method of temperature qualification by thermal equivalence is acceptable with the understanding that the containment temperature transients will be reviewed on a plant-specific basis. CE will also provide a detailed description of the analytical modeling techniques and application of the thermal equivalence approach, including the basis for selecting the parameters.

The following restrictions are part of the general application of the method-ology, and conformance shall be documented in each test report.

(1) Application of the thermal equivalence approach shall be justified for each piece of equipment, including any judgments regarding the surviva-bility limits of the equipment.

(2) The specific heat transfer modeling shall be described, and that the critical surface or surfaces are limiting both in time and location shall be justified.

CESSAR SSER 3 4 Appendix J

(3) Multiple temperature measurements of the critical suiface for testing shall envelope the calculated critical surface (s) temperature transient, including the initial temperature ramp. Soaking will not be permitted.

(4) The margin between the minimum measured surface temperature and the cal-culated surface temperature shall exceed the uncertainties associated with design, production, testing, and operability time, or a temperature margin of 15 by reference to the guidelines of IEEE Standard 323-1974 shall be applied. .

(5) Application of the thermal equivalence approach shall be restricted to the limiting superheated steam harsh environment. For the limiting erviron-ment and operability time, a spectrum of breaks shall be considered.,

The open item on thermal equivalence methodology identified in Appendix I of Supplement No. 2 has been described in CENPD-255-A (Revision 3) and is con-sidered acceptable by the staff.

The typical curves specified must be verified by the individual applicant to satisfactorily envelope plant-specific postulated conditions.

3.4 Temperature, Pressure, and Humidity Conditions Outside Containment CENPD-255-A (Revision 3) has developed a series of environmental condition cate-gories, most of which are for locations outside containment. These locations include areas affected by HELB conditions. Equipment in those areas will be qualified to these typical conditions. Other categories include conditions in a non-harsh environment, with normal and abnormal extremes specified. These categorical conditions must be reviewed by the specific applicant to determine applicability; however, this general approach is considered adequate.

3. 5 Submergence When submergence qualification is necessary because of flood-level and equipment-location analyses, CENPD-255-A (Revision 3) states that qualification will be performed by test according to type of equipment or partial tests and analyses. The effect of containment pressure during submergence is considered.

This is an adequate approach.

3.6 Chemical Spray Typical values for chemical spray art stated as 4400 ppm boron and 50 to 100 ppm hydrazine with a pH of 4 to 10. Actual values will be determined through single-failure analysi:; of the spray system to determine the most severe spray composition. Chemical spray simulation during in-containment testing starts 10 minutes after the start of the second transient. Each applicant must ensure that the value used envelopes the conditions postulated at the specific plant. This is an adequate general approach.

3.7 Aging--Equipment in Harsh and Non-Harsh Environments As stated in CENPD 255-A (Revision 3) and Amendment 9 to Chapter 3 of CESSAR (Section 3.11), the aging portion of the qualification program is defined on CESSAR SSER 3 5 Appendix J

the basis of whether or not equipment is located in a harsh or non-harsh

, environment. Equipment located in a harsh environment will undergo an aging analysis and an accelerated age conditioning program. Equipment located in a non-harsh environment will undergo an aging analysis that focuses on the identification of known aging mechanisms that significantly increase the equipment's susceptibility to its design-basis event (seismic event only for non-harsh environments). If no known significant aging mechanisms are found, a surveillance / preventive maintenance program will be developed to monitor for degradation trends that suggest increasing seismic susceptibility. If an aging mechanism is found that is known to significantly increase the equipment's seismic susceptibility with time, that mechanism will be analyzed to determine whether an accelerated aging program or a periodic part-replacement program is appropriate.

These general aging methodologies conform to the requirements of NUREG-0588 and IEEE Standard 323-1974 and, coupled with sound engineering judgment, provide an acceptable treatment of aging.

\

3.8 Radiation--Equipment in Harsh and Non-Harsh Environments Equipment will be designed for the tepes and levels of radiation associated with normal operation plus the radir. ion associated with the limiting design-basis accident (DBA). These levels are defined in CESSAR Appendix 3.11A.

Equipment that is exposed to *adiation above 104 rads will be irradiated to its anticipated total integrated Jose (TID) before type testing unless it is deter-mined by analysis that radiation does affect its ability to perform its required function. Where the application of the accident dose is planned during DBA testing, it need not be included during the aging process.

Equipment that will be exposed to radiation levels of 104 rads or below will be analyzed to determine whether low-level radiation could affect its ability to perform its required function. Where analysis supported by partial type test data cannot demonstrate proper operation at the required radiation levels, type testing will be performed. Additionally, electronic equipment exposed to low-level radiation will be addressed by an aging analysis that focuses un the identification of semiconductor (organic material) components that are con-sidered to be age sensitive in 40 years. For electronic components that are age sensitive, a surveillance / preventive maintenance program will be developed.

CENPD-255-A (Revision 3) outlines this methodology.

Mechanical / electrical equipment will be qualified to the typical radiation environments defined in CESSAR Appendix 3.11A. If more than one type of radia-tion is significant, each type may be applied separately.

Gamma Cobalt-60 is considered an acceptable gamma radiation source. Other sources may be found acceptable and will be justified. Electrical equipment will be tested to typical gamma radiation levels defined in CESSAR Appendix 3.11A.

CESSAR SSER 3 6 Appendix J

Beta Equipment exposed to beta radiation will be identified, and an analysis will be performed to determine if the operability of the equipment is affected by beta radiation ionization and heating effects. Qualification will be performed by test unless analysis demonstrates that the safety function will not be degraded by exposure to beta radiation. Equipment will be tested and/or analyzed to the beta radiation levels defined in Appendix 3.11A. Where testing is recommended, an equivalent gamma radiation source will be used.

Neutron Equipment exposed to neutron radiation will be identified and neutron radiation levels defined. When actual neutron dose qualification testing is not per-formed, an equivalent gamma radiation dose will be used for qualification testing to simulate neutron exposure. The basis for establishing an equivalent gamma radiation dose will be provided. '

l Pj:.its and Radiation Effects An analysis will be performed addressing exposure of paint to beta and gamma radiation. Qualification of painted equipment will be by test if analysis indicates that the safety function of the equipment could be impaired by fail-ure caesed by radiation.

The radiation aging methodology described conforms to the positions of IEEE Standard 323-1974 and NUREG-0588, which provide an acceptable treatment of radiation. Eaci; applicant must ensure that, for each equipment item and loca-tion, the plant specific postulated radiation dose is enveloped by the dose specified for that item.

3.9 Test Sequence The test sequence employed for equipment subject to operation in an HELB envi-ronment is the sequence outlined in IEEE Standard 323-1974. As allowed by this standard, performance extremes testing may be done on separate equipment items.

l The test sequence in CENPD-255-A (Revision 3) is acceptable.

l l 3.10 Margin i

CENPD-255-A (Revision 3) approaches margin in several ways, depending on the parameter.

Derivation of the in-containment temperature and pressure envelopes for testing purposes have 15 Fahrenheit degrees and 10 psi added to ensure adequate margin.

The margin available in specific plant applications will vary.

A commitment to a radiation margin of 10% of the accident dose is included, unless the radiation dose is calculated on the basis of NUREG-0588, Appendix D.

In this case, additional margin will not be added.

The above applications of margin are adequate, keeping in mind that each ap-plicant must verify acceptable margin on the basis of plant-specific information.

CESSAR SSER 3 7 Appendix J

CENPD-255-A (Revision 3) approaches time margin in two ways. The first, to be employed in a majority of cases, utilizes the standard margin "10% or 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, whichever is greater" mandated by NUREG-0588.

The second approach, entitled " unusual time margin," deviates from this stan-dard. Discussions with the staff resulted in an acceptance of CE's methodology subject to the following.

If it is necessary to use time margin evaluation techniques, the following information, as a minimum, shall be documented.

(1) Application of time margins less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> shall be justified for each piece of equipment, including any judgments regarding the survivability limits of the equipment.

(2) The maximum operability time shall be. justified with consideration of a spectrum of breaks and the potential need for the equipment later in an event or during recovery operations.

(3) It shall be demonstrated that failure of the equipment after the maximum operability time will neither mislead the operator to take an improper action nor further degrade the event by causing a failure in systems necessary for mitigation of the event.

(4) The margin applied to the minimum operability time when combined with other test margins shall account for the uncertainties associated with the design, production tolerances, testing techniques, and number of units tested.

The open item on " unusual time margin" identified in Appendix I of CESSAR SER Supplement No. 2 has been described in CENPD-255-A (Revision 3) and is now considered acceptable by the staff.

4 CONCLUSIONS The methodology used by CE to qualify NSSS safety-related electrical equipment is outlined in CENPD-255-A (Revision 3). It is expected that this report may be referenced by license applicants for the scope and methods employed for qualification of individual equipment items.

The qualification program in CENPD-255-A (Revision 3) is generic, with param-eters developed to envelope a range of different plant and containment designs.

Accordingly, each applicant must ensure that the time-dependent environmental parameters used in the testing of each equipment item, as well as assumed normal and abnormal conditions used in analysis, envelop the corresponding values developed by plant-specific analysis.

The staff concludes that, with the caution noted above, the general approach to environmental qualification of safety-related electrical equipment is acceptable and meets the requirements of 10 CFR 50.49 and its associated standards.

CESSAR SSER 3 8 Appendix J

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3 LE AVE BLANE Safety Evlluation Report related to the final design of the Standard Nhqlear Steam Supply Reference System CESSAR Syste 80

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