ML20248H850

Nonproprietary Amend 3 to RESAR-SP/90 Pda Module 2, Regulatory Conformance

ML20248H850
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Site:	05000601
Issue date:	08/31/1989
From:	WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
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ML19307A421	List: ML20248H846 ML20248H850 NRC-89-3461, Forwards Proprietary & Nonproprietary Amend 3 to RESAR-SP/90 Pda Module 2, Regulatory Conformance
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..,.i WESTINGHOUSE CLASS 3- -

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RESAR-SP/90 PDA-Amendment 3-to Module 2.

f Regulatory Conformance 20 y.

Westinghouse Electric Corporation P.O. Box 355 Pittsburgh, PA 15230 0 HAPWR-RC 2

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AUGUST 1989 0006D:1D rgio #d,1 AMENDMENT 3 PDB A

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Amendment-3 to RESAR-SP/90 PDA Module 2 Regulatory Conformance 1

v Instruction' Sheet 40 u Replace current page 11 with revised page 11.

Replace current page 1.0-1/1.0-2 with revised page 1.0-1/1.0-2.

Replace current page 3.0-1 with revised page 3.0-1/3.0-2.

Replace current pages 3.1-1 through 3.1-55 with revised pages 3.1-1 through 1

3.1-53.

Replace current pages 3.2-1 through 3.2-17 with revised pages 3.2-1 through 3.2-6.

O Replace current pages 3.3-1 through 3.3-13 with revised pages 3.3-1 through 3.3-36.

• Replace current pages 4.0-1 through 4.0-55 with revised ,pages 4.0-1 through j 4.0-53.

Replace current page 5.0-1/5.0-2 with revised pages 5.0-1 through 5.0-3.

Replace current pages 5.1-1 through 5.1-37 with revised pages 5.1-1 through O- 5.1-21.

Replace current pages 5.2-1 through 5.2-59 with revised pages 5.2-1 through 5.2-42.

Replace current pages 5.1-1 through 5.3-13 with revised pages 5.3-1 through 5.3-13.

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0006D:1D AMENDMENT 3 I

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-Replace current page 5.4-1/5.4-2 with revised page 5.4-1/5.4-2.

l Replace current pages 5.5-1 through 5.5-43 with revised pages 5.5-1 through 5.5-111.

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. Replace current page 6. 5-1/6.'5-2' with revised page 6.5-1/6.5-2.

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-Replace current page 6.5-7/6.5-8 with revised page 6.5-7/6.5-8.

Place pages 260-1 through 260-16 behind Amendment 2 in the Question / Answer section of Module 2.

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WAPWR-RC AUGUST 1989 D006D:1D AMENDMENT 3

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TABLE OF CONTENTS t.

Section Title Pggg

1.0 INTRODUCTION

1.0-1 2.0 LICENSING REQUIREMENT SOURCES 2. 0-1 3.0 3.0-1 f POST-TMI REQUIREMENTS AND RECOMMENDATIONS 3.1 10CFR 50.34(f). Additional TMI-Related Requirements 3.1-1 3.2 Severe Accident Review and Related 3.2-1 Considerations 3.3 Other Post-TM1 Issues 3.3-1 3 3.3.1 NUREG-0737 3.3-2 3.3.2 NUREG-0660 3.3-4 3.3.3' NUREG-0985, Human Factors Program Plan Issues 3.3-25 4.0 UNRESOLVED SAFETY ISSUES 4. 0-1 5.0 GENERIC SAFETY ISSUES - TASK ACTION PLAN CATEGORY A, B, 5.0-1 3 C and D ISSUES 5.1 Category A Issues 5.1 -1 5.2 Category B Issues 5.2-1 5.3 Category C Issues 5.3-1 5.4 Category D Issues 5.4-1

. 5.5 New Generic Issues 5.5-1 3 6.0 REGULATIONS AND REGULATORY GUIDANCE 6.1 -1 6.1 NRC Regulations 6.1-1 6.1.1 New Rules 6.1 -1 6.1.2 Proposed Rules /Rulemakings 6.1-9 1

O 6.1.2.1 Proposed Rules 6.1 -9 6.1.2.2 Advance Notices of 6.1-15 Proposed Rulemaking 6.1.2.3 Unpublished Rules 6.1-18 6.2 Regulatory Guides 6-2-1 l 6.2.1 Division 1 Regulatory Guides - 6.2-2 Power Reactors 6.2.2 Division 4 Regulatory Guides -- 6.2-2 Environmental and Siting O  !!APWR-RC ii AMENDMENT 3 0127e:10/092789 AUGUST, 1989 l

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1.0 INTRODUCTION

Westinghouse is in the process of developing an Advanced Pressurized Water i

( Reactor design (SP/90) for domestic as well as international application in O] the 1990's time f rame. This total plant design has been developed through a l 3 i

major cooperative effort with a Japanese vendor and is directed toward the establishment of final design detail and completion of an extensive test program. -l 3 f]

V Westinghouse believes that ensuring compliance of the SP/99 design with current licensing requirements is best satisfied through consideration of regulatory requirements and guidance as an integral part of the initial design process. Westinghouse also has a unique opportunity to optimally address new and potential future licensing requirements in the SP/90 design. Westinghouse has attempted to take full advantage of this design optimization opportunity 3 with the goal of achieving assurance that the fundamental SP/90 design will remain relatively unchanged as a result of future regulatory changes.

This report presents the bases for the SP/90 design in response to licensing requirements consistent with the Nuclear Regulatory Commission Policy

, Statement on Severe Reactor Accidents Regarding Future Designs and Existing 3 Plants (50FR32138), August 1985 and the Policy Statement on Safety Goals for the Operation of Nuclear Power Plants (50FR30028), August 21, 1986. For the purpose of this report, licensing requirements have been categorized as follows:

o Historical requirements o Recent new or revised requirements o Potential future requirements v .

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WAPWR-RC 1.0-1 AMENDMENT 3 AUGUST, 1989 0127e:10/092789

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The major sections of this report address the latter two categories of licensing requirements (i.e., recent new or revised requirements and potential future requirements) including a description of the planned course of action {

for addressing the requirements ia the SP/90 design.

l The sources of licensing req'.'rements considered by Westinghouse as comprising i

these two categories are discussed in Section 2.0.

The appendices to this report, which presen't design bases for specific SP/90 structures, systems, components, features, etc., address all three categories of licensing requirements and are intended to be used directly in the licensing process for the SP/90 design.

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WAPWR-RC 1.0-2 AMENDMENT 3 D127e:10/021789 MARCH, 1989

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3.0 POST-TMI REQUIREMENTS AND RECOMMENDATIONS

' Shortly after the initial recovery phases following the March 28, 1979 i

incident at TMI-2, various task forces and investigating groups were set-up

l. ( (both inside and outside'of the NRC) to make recommendations for plant design and operating changes to ensure that a TMI-2 type event or similar event does-not happen again. The requirewMs and recommendations from these task forces and investigating groups were consolidated and documented in NUREG-0660, "NRC Action Plan Developed as a Result of the TMI-2 Accident."

l NUREG-0737, " Clarification of TMI Action Plan Requirements," was issued in l November 1980 to provide clarification of those action items which were approved for implementation at the date of issue.

l 3 l The initial regulatory efforts did not specifically address requirements for l new plant designs, since at that time the NRC directed its technical review resources to assuring the safety of operating power reactors rather than the l issuance of new licenses or permits.

In mid-1980 the NRC staff initiated a program for Commission approval of a course of action that would lead to the establishment of TMI related requirements for pending construction permit applications. This program led to the issuance of a revision to 10CFR50.34, " Contents of Applications; Technical Information," that incorporates applicable post-TMI requirements into the NRC regulations for pending CP applications. NUREG-0718, Revision 2

" Licensing Requirements for Pending Applications for Construction Permits and Manufacturing License," issued in January 1982, provides guidance for compliance with those NRC Action Plan (NUREG-0660) items required to be implemented or committed to by a pending applicant prior to receiving a construction permit or a license to manufacture. A discussion of applicable 10CFR50.34 (f) requirements is presented below.

• Prior to the TMI event, USNRC regulations required general consideration of beyond-design-basis accidents. However, no explicit analysis of consequence was required. Subsequent to TMI, the industry began a focused consideration O

WAPWR-RC 3.0-1 AMENDMENT 3 D060e:1d AUGUST 1989

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of degraded core events. The Severe Accident Policy Statement describes the l policy that the commission intends to use to resolve safety issues relating to reactor accidents more severe than design basis accidents. Design l considerations relating to severe accident rulemaking are pravided in the )

3 following sections.

Certain TMI action plan items require further development before an adequate generic resolution can be established. These are further discussed in the

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following sections.

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f,s 3.1 1JCFR50.34(f), ADDITIONAL THI-RELATED REQUIREMENTS b

The revision to 10CFR 50.34 is written such that it is applicable to construction permit applications pending at the effective date of the rule

,3 (i.e., February 16, 1982). liowever, the NRC " Policy Statement on Severe U Reactor Accidents Regarding Future Designs and Existing Plants" (50FR32138 3 August 8, 1988) indicates that the requirements of10CFR50.34(f)arealso applicable to new construction permit applications or reactivation.

Therefore, applicable post-TMI requirements of 10CFR 50.34 are being addressed (j in the SP/90 design as indicated in the following sections.

i The following are the licensing requirements and SP/90 design responses for  !

each 10CFR 50.34(f) item that impacts or potentially impacts the SP/90 design. I

1. Plant / Site Specific Probabilistic Risk Assessment 3 l 10CFR 50.34(f)(1)(i) (II.B.8) l

• Perform a plant / site specific probabilistic risk assessment, the aim of which is to seek such improvements in the reliability of core and containment heat removal systems as are significant and practical and do not impact excessively on the plant."

SP/90 Response Refer to Section 3.2 which has been devoted to the interrelated issues of probabilistic risk asressment, safety goal, and severe accidents.

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10CFR 50.34(f)(1)(ii) (II.E.1.1) p.

C/ " Perform an evaluation of the proposed auxiliary feedwater system (AFWS),

to include (applicable to PWR's only): (A) a simplified AFWS reliability b,/m WAPWR-RC 3.1-1 AMENDMENT 3 UO60e:1d AUGUST 1989

i analysis using event-tree and fault-tree logic techniques, (B) a design ]

review of AFWS, and (C) an evaluation of AFWS flow design bases and

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criteria." I l

4 Discussion 1 A conventional AFWS functions, in conjunction with a seismic Category I )

water source, as an emergency system for the removal of heat from the i primary system when the main feedwater system is not available. It also plays an important role in mitigating the effects of some design basis events (e.g., main feedwater line breaks and some small break  !

loss-of-coolant accidents). Existing AFWS designs hold the plant at hot standby, or cool down the primary system to temperature and pressure levels at which the low pressure residual heat removal system can operate. The AFWS can also be used during ncrmal plant startup and shutdown conditions. AFWS designs usually consist of a combination of steam turbine-driven and electric motor-driven pumps.

SP/90 Response The purpose of requirement (A) above is to: (1) assess the reliability of the AFWS design under various loss of feedwater transient conditions, with particular emphasis being given to determining potential failures that could result from human errors, common causes, single point vulnerabili-ties, and test and maintenance outages, and (2) incorporate design 3 provisions and/or procedural actions as necessary to improve the AFWS reliability relative to the NRC generic AFWS reliabilities published in NUREG-0611, " Generic Evaluation of Feedwater Transients and Small Break Loss-of-Coolant Accidents in Westinghouse-Designed Operating Plants."

The purpose of requirement (B) above is to: (1) assess the level of compliance of the AFWS design to the NRC acceptance criteria c'ocumented in l Standard Review Plan 10.4.9, " Auxiliary Feedwater Systems," and (2) where deviations are identified, modify the AFWS design as necessary to comply with the NRC acceptance criteria or justify the deviatunr.

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4 The p'urpose of requirement (C)' above is to assure that the design bases and criteria for establishing AFWS requirements for flow to the steam

. generator (s) to assure adequate removal of reactor decay heat are defined and documented. J The SP/90 design is somewhat different than a conventional two electric motor-driven and one steam turbine-driven AFWS design.

The SP/90 design includes an emergency feedwater system (EFWS)anda startup feedwater system (SFWS). The EFWS is a safety system utilizing four pumps; two electric motor-driven and two steam turbine-driven. -The EFWS functions similarly to a conventional AFWS except that during normal plant startup/ shutdown and hot standby the SFWS is utilized. The EFWS is designed for such events as main steam line breaks, main feedwater line breaks, steam generator tube ruptures, loss-of-coolant accidents, loss of all AC power, and any other event in which the main and startup feedwater 3

systems are not available. The SFWS is a control grade system utilizing one motor-driven pump and provides feedwater during normal plant startup/

shutdown and hot standby. The SFWS is also started automatically during reactor trips and other anticipated transients.  !

. In regard to requirement (A), a detailed reliability analysis of the combined SFWS/EFWS has been performed. The results are contained in Section 3.7 of RESAR-SP/90 PDA Module 16, "Probabilistic Safety Study." j In regard to requirement (B), Westinghouse has performed a Preliminary Design Review of the EFWS to satisfy internal Quality Assurance requirements. Additional design reviews will be conducted during the Final Design Application phase.

In regard to requirement (C), the Preliminary Design Review referred to above included evaluations of SFWS/EFWS design bases and criteria.

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3. Reactor Coolant Pump Seals 3 10CFR 50.34(f)(1)(iii) (II.K.2.16 and II.K 3.25)

" Perform an evaluation of the potential for and impact of reactor coolant pump seal damage following small-break LOCA with loss of offsite power.

If damage cannot be precluded, provide an analysis of the limiting small-break LOCA with subsequent reactor coolant pump seal damage."

SP/90 Response The potential for RCP seal damage is minimal due to the redundancy of the seal cooling design for the SP/90. During normal operation, seal injection flow is provided by the chemical and volume control system.

Should normal seal injection flow be interrupted, the seals are cooled by reactor coolant flowing upward from the pump bowl through the thermal barrier, which has been cooled by the component cooling water passing through the thermal barrier heat exchanger. In addition to the above normal seal cooling provided in conventional designs, the SP/90 design includes an AC-independent backup seal injection capability which provides g

3 an alternate source of seal injection water to the reactor coolant pumps during situations involving the loss of both normal seal injection and thermal barrier cooling. Operational experience with seals incorporating redundant cooling features shows an excellent history (WCAP 10541).

Specifically, for the limited number of events in which all seal cooling was lost, the data indicated that the seals were not severely damaged and were reused in the majority of instances.

In the event of a design basis small-break LOCA coincident with loss of offsite power, the charging pumps, which are the normal source of seal injection, are not credited to start automatically since they do not have Class IE motors. Upon loss of normal charging, the AC-independent backup seal injection pump will start automatically; however, no credit is taken for this pump in accident analyses since it is not safety grade.

Therefore, RCP seal injection is assumed to be unavailable. With the WAPWR-RC 3.1-4 AMENDMENT 3 D060e:1d AUGUST 1989

additional small-break LOCA limiting single failure assumption that only l

one of the two diesel generators starts, component cooling water will be available to only two of the four RCP seals. For this highly improbable scenario, the loss of seal cooling will result in a temperature increase

.C\ in the seal area and may result in an increase in the No. I seal leak rate for the two affected pumps. Based upon the analysis and test results in WCAP 10541, the increase in the leakage is self-limiting with a conserva-tive estimate of the stable leakage rate being 21 gpm/ pump, which briefly peaked at 61 gpm/ pump at a pressure of 2250 psig.

To address the concern of increased seal leakage coincident with small-break LOCA, analyses were performed using the Westinghouse NOTRUMP Evaluation Model (WCAP 10054). For conservatism, the seal leakage rate was assumed to be 500 gpm at 2300 psia pressure drop, modeled as an 0.56 inch diameter break at the top of the pump. Results of these studies showed that the effect of the RCP seal leakage was minor and that no 3

significant effects on the transient behavior was observed. In fact, the peak clad temperatures were slightly lower for the cases with the increased real leakage than for the limiting break case without seal leakage modeled. Thus, it was concluded that the seal leakage effects are covered by performing break spectrum studies.

The small-break LOCA analysis, reported in Section 15.6.4 of RESAR-SP/90 PDA Module 1, " Primary Side Safeguards System," covered a range of break sizes to establish the limiting small break. No core uncovery occurred for the 3 inch, 4.313 inch or 6 inch diameter breaks which were analyzed.

i Indeed, the minimum vessel level for the limiting 4.313 inch break remained more than three feet above the top of the core. Consideration of the additional seal leakage would have the effect of slightly increasing the equivalent break size. As in the studies in WCAP 10054, it is concluded that the seal leakage effects are covered by the break spectrum performed.

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l 4. Automatic PORV Isolation System 3 10CFR 50.34(f)(1)(iv) (II.K.3.2) l l

" Perform an analysis of the probability of a small-break LOCA caused by a g l stuck-open PORV. If this probability is a significant contributor to the W.

I probability of small-break LOCA's from all causes, provide a description and evaluation of the effect on small-break LOCA probability of an automatic PORV isolation system that would operate when the reactor coolant system pressure falls after the PORV has opened."

Discussion General Design Criterion 14, " Reactor Coolant Pressure Boundary," of Appendix A to 10CFR Part 50 requires that the reactor coolant pressure boundary be designed, fabricated, erected, and tested to have an extremely low probability of abnormal leakage, rapidly propagating failure, and gross rupture. Historically, the application of this criterion has emphasized the integrity of passive components in the reactor coolant system, such as the reactor vessel and the piping, however, this criterion also applies to the valves that provide isolation for the system.

The primary purpose of pressurizer relief and safety valves is that they operate in conjunction with the reactivity control system to limit system overpressure during anticipated operational transients or accidents. The pressurizer relief valves are not part of ASME Code requirements for overpressure protection and, therefore, they can be and are isolatable with remote-operated block valves. The consequence of the failure of the pressurizer relief valves to close is the loss of coolant and depressuri-zation of the reactor system. This consequence can be mitigated if the remote-operated block valves are closed either automatically or by operator action.

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' 1 The purpose of this requirement is to evaluate (using probabilistic O techniques) the benefit of incorporating an automatic pressurizer PORV isolation system. I SP/90 Response The SP/90 plant contains several design features which will minimize challenges to the PORVs. Firstly, the charging pumps, whose shutoff head is well. above the PORV opening setpoint, do not perform a safety injection 9 function. This function is performed by the High Head Safety Injection (HHSI) pumps, whose shutoff head has been selected to be well below the PORV opening setpoint. Secondly, the pressurizer and the steam dump 3 system have been sized to prevent PORV opening for all anticipated transients up to and including a full load rejection. Nevertheless, automatic closure of the PORV block valves is included'in the SP/90 design to further reduce the probability of a stuck open PORV.

9 NOTE: [(f)(1)(v) through (f)(1)(xi) is BWR only]

5. Hydrogen Control Systems Evaluation 10CFR 50.34(f)(1)(xii)

" Perform an evaluation of alternative hydrogen control systems that would satisfy the requirements of paragraph (f)(2)(ix) of this section (50.34).

As a minimum include consideration of a hydrogen ignition and postaccident inerting system. The evaluation shall include: (A) a comparison of costs and benefits of the alternative systems considered, (B) for the selected 9 system, analyses and test data to verify compliance with the requirements of (f)(2)(ix) of this section (50.34), and (C) for the selected system, preliminary design descriptions of equipment, function, and layout."

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SP/90 Response Refer to item 14 of this section for a discussion of this requirement in conjunction with the requirements of 10CFR 50.34(f)(2)(ix).

6. Simulator Capability 3 . 10CFR50.34(f)(2)(i)(I.A.4.2)

" Provide simulator capability that correctly models the control room and includes the capability to simulate small-break LOCA's."

Discussion Beyond the above regulation, 10CFR Part 55, Appendix A. "Requalification Programs for Licensed Operators of Production and Utilization Facilities,"

permits and encourages the use of simulators for operator training. This is due to. the undesirability of imposing additional challenges to the plants protective features that would result if the actual plant is used for training operators to respond to accidents.

The purpose of this NRC requirement is to: (A) require simulator capability,and(B) ensure that the proposed simulator capability for training of operators is performed on a simulator that correctly models the actual plant specific control room design and has the capability to accurately simulate a small-break LOCA.

In addition, the NRC has issued Regulatory Guide 1.149, " Nuclear Power Plant Simulators for Use in Operator Training," which basically endorses ANSI /ANS 3.5-1981, " Nuclear Power Plant Simulators for Use in Operator Training," and describes a method acceptable to the NRC staff for specifying the functional requirements of a nuclear power plant simulator i to be used for operator training.

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This requirement does not impact the SP/90 design. Simulator capability is the responsibility of each utility utilizing the SP/90 design.

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.7. Plant Procedures i

3 10CFR 50.34(f)(2)(ii) (I.C.9) J

" Establish a program to begin during construction and follow into operation, for integrating and expanding current efforts to improve plant i procedures. The scope of the program shall include emergency procedures, reliability analyses, human factors engineering, crisis management, operator training, and coordination with INPO and other industry efforts."

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Discussion p The area of operating procedures has received great attention as a result of the THI-2 event. This attention stems from certain opinions that the severity of the TMI-2 event might have been significantly reduced if the operating procedures were better written (human engineered and supported 3

. by appropriate analyses) and if the operators were better trained in the use of the procedures.

Since the TMI-2 event there have been extensive industry efforts undertaken to improve emergency operating procedures and their use. As one would expect, these efforts to date have been focused on current-day operating and near-term operating plants. The NRC concern that resulted in the above requirement is that programs for the continued improvement of

)I plant operating procedures should be pursued and coordinated with other 3 industry efforts (e.g., INPO) and other post-TMI related improvements I (e.g., safety parameter display systems) in relation to new applications.

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s SP/90 Response Emergency Response Guideline development for the W SP/90 design will be ]

based on the Westinghouse Owner's Group Emergency Response Guidelines {

(ERG). These guidelin'es reflect symptom based procedures that include l event-oriented and function-oriented guidance for accident recovery.

Revision 1 of the ERGS has been reviewed and approved for use on current W PWRs by the NRC staff. In general, these guidelines are applicable to the W SP/90 design; however, it is expected that some modifications may be l required to address W SP/90 specific features. An outline of the program ,

to identify and make the appropriate modifications is presented here.

The process of emergency procedure development for the W SP/90 will be similar to the current process of developing plant specific procedures for current W plants.

3 A review of the W SP/90 design features will be performed to identify differences between the W SP/90 design and the ERG refervace plant design. The impact on the ERG guidance of the SP/90 design features will be assessed and documented. It is expected that the magnitude of these differences will not be greater than those identified in current plant-specific applications of the ERGS.

The identification of design differences will be used as input for an assessment of the applicability to the W SP/90 design of the reference plant analyses which have been performed to verify the actions specified in the ERGS. Any new analyses required to support actions for the W SP/90 not already covered by the reference plant analyses will be identified and performed, as necessary. It is not expected that significant new analysis will be required.

The final step will be development of W SP/90 specific technical guidelines that will specify the identified modifications to the ERGS to develop emergency procedures for each application of the W SP/90 design.

The guidelines will specify differences and recommended modifications. It WAPWR-RC 3.1-10 AMENDMENT 3 O

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is not intended to create a new complete set of ERGS specific to the W SP/90, but the existing WOG ERGS will be maintained as the basis document for the W SP/90 emergency procedures.

l The~ advantage of this process is that the W SP/90 emergency guidance will

[ be based to the largest extent possible on the existing guidance in use at all other W plants. The existing ERGS are maintained by the WOG and W and 3 the W SP/90 guidance will benefit from any improvements identified in the future on existing plants.

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Prior to issuance of a final design approval (and in a timely manner that permits verification, possible NRC review, and possible operator training)

Westinghouse will develop the actual SP/90 Emergency Response Guidelines.  !

8. Control Room Design (I.D.1) 3 10CFR 50.34(f)(2)(iii)

(G) " Provide, for Commission review, a control room design that reflects state-of-the-art human factor principles prior to committing to fabrication or revision of fabricated control room panels and layouts."

Discussion General Design Criterion 19, " Control Room," of Appendix A to 10CFR Part 50 requires that a control room be provided from which actions can be '

taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under accident conditions.

For current-day operating plant licensees and applicants, this item is being implemented as a detailed review of their control room designs with the purpose of correcting weaknesses to improve the ability of control room operators to prevent accidents or cope with accidents if they occur.

The NRC has issued guidanca for performing control roon design reviews in i the form of NUREC-0700, " Guidelines for Control Room Design Reviews."

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NUREG-0700 is written specifically for existing control room designs and new guidance or criteria may be issued in the future for new control room designs.

Again for existing control room designs, the NRC has issued draft acceptance criteria for control room design reviews which is documented in NURES-0801, " Evaluation Criteria for Detailed Control Room Design Review." NUREG-0801 includes guidelines for the organizational structure and personnel qualifications for performing control room design reviews as gl well as guidelines for the actual review process and results documentation. T l

This requirement, as written, simply states that the control room design must be submitted to the NRC for review prior to fabrication. Inherent with this requirement is that it must be demonstrated to the NRC that the I control room design meets applicable licensing criteria (e.g., human factors engineering, new instrumentation requirements, etc.).

Key to the overall issue of ensuring a good control room design is the i fact that there are numerous current-day licensing issues that impact and may even conflict with good control room design. The NRC has indicated through NUREG-0737, Supplement 1, " Requirements for Emergency Response Capability," that these requirements (including the safety parameter display system, Regulatory Guide 1.97 instrumentation, emergency operating procedures, etc.) should be integrated with respect to the overall enhancement of the operators ability to comprehend plant conditions and cope with potential emergencies.

SP/90 Response RESAR-SP/90 PDA Module 15, " Control Room / Human Factors Engineering,"

describes the control room design and its conformance to applicable 3

criteria.

The overall control room design for the SP/90 is intended to integrate the requirements of this regulation concerning human factors principles with WAPWR-RC 3.1-12 AMENDMENT 3 9

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Lthe. requirements of . .the other regulations' concerning control room

, W .- instrumentation and their installation (e.g.,-the instrumentation required by items: 9, 10, 21, 22,.and 23 below, and the long standing requirements 3 for PAMS (Regulatory Guide 1.97) installation--and for Class 1E controls

-(IEEE-279,etc.).

9. Safety Parameter Display System 3

10CFR 50.34(f)(2)(iv) (I.D.2)

" Provide a plant safety parameter display console:that will-display to operators a minimum set of parameters defining _ the safety status of the plant, capable of ~ displaying a full range of important plant parameters and data. trends on demand, and capable of indicating when . process . limits are'being approached or exceeded."'

Discussion The purpose' of the plant safety . parameter display console (or safety

. parameter display system).is to provide a concise display of critical plant variables to control room personnel in order to assist them in rapidly and reliably determining the safety status of the plant. Although not specifically mentioned in the above regulation, the NRC is recom-mending that the licensee consider duplication of the safety parameter display console displays in the onsite technical support center and the near-site emergency operations facility to improve the exchange .of information between these facilities and the control room and assist corporate and plant management in the decision-making process.

In general, this requirement is no different from that currently being implemented by operating plant ~ licensees and applicants in response to NUREG-0737, Supplement 1, " Requirements for Emergency Response Capability." -

The NRC has also issued NUREG-0696, " Functional Criteria for Emergency <

Responso facilities," which provides certain guidance information for the O WAPWR-RC 3.1-13 AMENDMENT 3 1

5060e:1d AUGUST 1989 i

implementation of a safety parameter display system. In addition, the NRC has issued draft human factors acceptance criteria for safety parameter display systems which are documented in NUREG-0835, " Human Factors Evaluation Criteria for Safety Parameter Display Systems."

SP/90 Response Subsection 18.3.2 of RESAR-SP/90 PDA Module 15, " Control Room / Human l Factors Engineering" describes the SP/90 alarm system and its integration, along with other design aspects, into the complete control room design.

3 The alarm system for the SP/90 control room will include the attributes of a safety parameter display system (SPDS), and will meet the intent of the requirements of NUREG-0696, " Functional Criteria for Emergency Response Facilities," and NUREG-0835, " Human Factors Evaluation Criteria for Safety Parameter Display Systems."

10. Safety System Status Indication 3 10CFR 50.34(f)(2)(v) (I.D.3) g

" Provide for automatic indication of the bypassed and operable status of safety systems."

Discussion 10CFR 50.55a(h) requires that protection systems meet the requirements set forth in IEEE Standard 279-1971, " Criteria for Protection Systems for Nuclear Power Generating Stations." Section 4.13 of IEEE Standard 279-1971 requires that, if the protective action of some part of the protection system has been bypassed or deliberately rendered inoperative for any purpose, this fact shall be continuously indicated in the control room.

The intent of this requirement is to provide the operator with an automatic indication of the bypassed or inoperable status of systems and I WAPWR-RC 3.1-14 AMENDMENT 3 UO60e:1d AUGUST 1989 1

s components that perform a function important to safety in accordance with Regulatory Guide 1.47, " Bypassed and Inoperable Status Indication for Nuclear Power Plant Safety Systems." It should be noted that Regulatory Guide 1.47 does. permit certain limited use of manual activation of system-level indicators.

The NRC has also issued Branch Technical Position ICSB 21, " Guidance for

. Application of Regulatory Guide 1.47," which provides (as its title suggests) additional NRC guidance for implementation of Regulatory Guide 1.47.

SP/90 Response The Bypassed and Inoperable Status Indication (BISI) System hardware has been incorporated in the integrated I&C architecture described in Chapter 7. The bypassed and inoperable information necessary to meet the intent of Regulatory Guide 1.47, " Bypassed and Inoperable Status Indica-tion for Nuclear Power Plant Safety Systems," has been integrated into the plant display system. A physically independent system is not being supplied. Detailed design drawings of the complete plant display system will not be available until the final design phase, however, samples are 3

, provided in RESAR-SP/90 PDA Module 15, " Control Room / Human Factors Engineering." The drawings will not be specific to BISI.

All protection systems, and systems actuated or controlled by the protection system, will be automatically indicated if inoperability was intentionally induced or the system bypassed. Also included will be those systems which directly support automatically initiated systems but which themselves may not be automatically initiated because they are normally in the operating mode. Related support systems may have their own subsystem bypass and inoperability indication as well as input to the primary system

[3 indication, i.e., status information for support systems will be pyramided )

up to all related primary safety systems.

WAPWR-RC 3.1-15 AMENDMENT 3 UO60e:1d AUGUST 1989 .

l i

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ J

s The information available to the operating staff with respect to inopera-bility and bypass is at least as comprehensive as that of an independent system. Both overall system state (i.e., the ability to perform) and actuation status (i.e., failure to perform) will be available. These capabilities for each of these purposes are likely to take the form of binary lights for the ISS. The alarm system will drive these lights based on its results in recognizing the completion or lack of completion of the patterns for the " readiness" and for the " active" system status. The status of the systems not having these lights will be available by accessing system mimics of the display system.

The design philosophy of that portion of the display system that fulfills the requirements for a BISI System is the same as for the remainder of the display system. This approach to display bypassed and inoperable status is part of the overall functional decomposition performed to determine what information is provided in the control room and the manner in which it is displayed. tiigh level goals are identified and then decomposed into system level indications, component level summaries, and support system status. With the BISI requirements incorporated at the conceptual level of the display system in the integrated system, the same operator mental model and display rules are used throughout the system, minimizing the potential for operator confusion and errors.

That equipment rendered inoperable for maintenance less frequently than once per year will not necessarily be automatically indicated. The display system will have the capability to indicate manual initiation of bypass of safety features on a system level. Under administrative control, manual bypass indication can be input or removed. The automatic indication feature cannot be overridden by operator action.

The BISI System, while not a Class 1E system, will not degrade the Class 1E systems with which it interfaces if subjected to a credible event. The isolation provisions it uses to satisfy this requirement will meet Class 1E criteria.

J WAPWR-RC 3.1-16 AMENDMENT 3 e

UO60e:1d AUGUST 1989

s

(~N 11. Reactor Coolant System High Point Vents c)  ;

10CFR 50.34(f)(2)(vi) (II.B.1) 3

(] " Provide the capability'of high point venting of noncondensible gases from V the reactor coolant system, and other systems that may be required to maintain adequate core cooling. Systems to achieve this capability shall be capable of being operated from the control room and their operation

,/ shall not lead to an unacceptable increase in the probability of k loss-of-coolant accident or an unacceptable challenge to containment i

integrity."

Discussion 10CFR 50.46(b)(5) requires that after any calculated successful initial operation of the emergency core cooling system, the calculated core temperature shall be maintained at an acceptably low value and decay heat shall be removed for the extended period of time required by long-lived I radioactivity remaining in the core. Additionally, General Design Criterion 35, " Emergency Core Cooling," of Appendix A to 10CFR Part 50 requires that a system to provide abundant emergency core cooling shall be

, provided. The system safety function shall be to transfer heat from the reactor core following any loss of reactor coolant at a rate such that:

(A) fuel and clad damage that could interfere with continued effective core cooling is prevented, and (B) clad metal-water reaction is limited to I

negligible amounts.

During the THI-2 accident, a condition of low water level in the reactor vessel and inadequate core cooling existed and was not rectified for a l long period of time. The resultant high core temperatures produced a l metal-water reaction with the subsequent production of significant amounts of hydrogen. The collection of noncondensible gases impaired natural l]

b circulation cooling capability. Additionally, the cc11ection of l

noncondensible gases limited reactor coolant pump operational capability l oecause of coolant voids in the system occupied by the gases. Even when 1

v -

WAPWR-RC 3.1-17 AMEN 0 MENT 3 UO60e:1d AUGUST 1989

s reactor coolant pump operation was possible, the installed plant venting system was capable of removing the noncondensible gases only through an extremely slow process.

The purpose of this requirement is to provide for the capability of reactor coolant system high point venting of noncondensible gases collected in the system in order to allow satisfactory long-term core cooling.

The above 10CFR 50.34 regulation must be considered in conjunction with the recent requirements of 10CFR 50.44(c)(3)(iii). This regulation, which is part of the NRC interim requirements related to hydrogen control, also mandates the installation of high point vents.

"To provide improved operational cacability to maintain adequate core cooling following an accident, . . . each light-water nuclear power reactor shall be provided with high point vents for the reactor coolant systein, for the reactor vessel head, and for other systems required to maintain adequate core cooling if the accumulation of noncondensible gases would cause the loss of function of these systems. (High point vents are not required, however, for the tubes in U-tube steam generators.) The '

high point vents must be remotely operated from the control room. Since the vents form a part of the reactor coolant pressure boundary, the design of the vents and associated controls, instruments and power sources must conform to the requirements of Appendix A and Appendix B of this part (10CFR Part 50). In particular, the vent system shall be designed to ensure a low probability that: (A) the vents will not perform their ,

safety functions, and (B) there would be inadvertent or irreversible I actuation of a vent. Furthermore, the use of these vents during and following an accident must not aggravate the challenge to the containment or the course of the accident."

O 3 l WAPWR-RC 3.1-18 AMENDMENT 3 9

UO60e:1d AUGUST 1989

l L .

3 SP/90 Response

(

The SP/90 plant includes venting of both the reactor vessel head and the pressurizer. The vents are designed to safety class standards and are operable from the Main Control Room. The RESAR-SP/90 Probabilistic Safety 3 Study-(PDA Module 16) demonstrates that the inclusion of these vents has not led to an unacceptable' increase in the probability of loes-of-coolant-accident, nor to unacceptable challenges to containment integrity.

12. Plant Shielding 3

10CFR 50.34(f)(2)(vii) (II.B.2)

" Perform radiation and shielding design reviews of spaces around systems  !

that may, as a result of an accident, contain TID-14844 source term radioactive materials, and design as necessary to permit adequate access to important areas and to protect safety equipmer,t from the radiation environment."

Discussion 10CFR Part 20 and General Design Criteria 19, 60, and 64 of Appendix A to 10CFR Part 50 require the control of radiation exposure associated with '

plant operations. General Design Criterion 4, " Environmental and Missile l Design Bases," requires that systems and components important to safety be designed to accommodate the environmental conditions associated with accidents.

O After an accident in which significant core damage occurs, the radiation source terms may approximate those of Regulatory Guide 1.4, " Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss-of-Coolant Accident for Pressurized Water Reactors." In addition, systems that were not designed to contain large radiation sources may become highly radioactive. The resulting radiation fields may make it difficult to effectively perform accident recovery operations or may WAPWR-RC 3.1-19 AMENDMENT 3 U060e:1d AUGUST 1989

(

l i

impair safety equipment. Currently, a combined NRC/ nuclear industry effort is underway to more accurately define the source terms based upon l the information obtained as a result of the TMI incident. l The purpose of this requirement is to facilitate post-accident operations l using systems that may contain abnormally high levels of radioactivity and I I

to ensure that safety equipment in proximity to the resulting radiation

- fields is not unduly degraded.

Current NRC guidance for performing radiation and shielding design reviews is detailed in Item II.B.2 of NUREG-0737, " Clarification of TMI Action Plan Requirements." Basic in this guidance is that the reviews should identify the location of vital areas and equipment (such as the control room, onsite technical support center, sampling station and sample analysis area, containment isolation reset control area, security center, radwaste control stations, emergency power supplies, motor control centers, and instrument areas) in which personnel occupancy may be unduly limited or safety equipment may be unduly degraded by the radiation fields during post-accident operations of these systems. These design reviews are intended to identify corrective actions (e.g., design changes) g necessary to provide for adequate access to vital areas and protection of safety equipment.

SP/90 RESPONSE The only systems outside containment that may be required to operate post-accident with highly radioactive fluid are the Integrated Safeguards 3

Systems (ISS), which includes the emergency core cooling, residual heat removal, and containment spray functions, and the Post Accident Sampling System (PASS).

The SP/90 ISS consists of four independent and separate subsystems which are located in four physically separate and independent safeguard component areas. Each of these four areas can be individually accessed.

WAPWR-RC 3.1-20 AMENDMENT 3 O

D060e:1d AUGUST 1989 l

[

s

.. N This system configurationsand related layout /HVAC arrangement facilitates b post-accident recirculation . operation and ; maintenance and provides a significant improvement relative to current plants.

3 The area containing the PASS is designed to' allow operation of this system O' even in the presence of highly radioactive fluid. l

. 13. Post-Accident Sampling 3

10CFR 50.34(f)(2)(v8%) (II.B.3)

" Provide a ' capability to promptly obtain and analyze samples from the reactor coolant' system and containment that may contain TID-14844 ' source term radioactive materials' without radiation exposures to any individual exceeding 5 rem to.the whole-body or 75 rem to the extremities. Materials to be analyzed and quantified include. certain radionuclides that are.

indicators of.the degree of core damage (e.g, noble gases, iodines and cesiums, and non-volatile isotopes), hydrogen in the containment atmosphere, dissolved gases, chloride, and boron concentrations."

Discussion Prompt sampling and analysis of reactor coolant and the containment atmosphere can provide information important to the efforts to assess and control the course of an accident. Chemical and radiological analysis of reactor coolant liquid and gas samples can provide substantial information regarding core damage and coolant characteristics. Analysis of containment atmosphere (air) samples can determine if there is any prospect of hydrogen reaction in containment, as well as provide core damage information.

Beyond .the above requirement, no definitive regulations exist for obtaining and analyzing reactor coolant or containment samples following an accident. . NRC guidance and acceptance criteria (e.g, Standard Review Plan 9.3.2 and Regulatory Guide 1.97) have, however, been revised since the TMI-2 event to require this capability.

WAPWR-RC 3.1-21 AMENDMENT 3 D060e:1d AUGUST 1989

(

Most recent plant designs include (or are being modified to include) post-accident sampling capability through the use of in-line monitoring systems. These systems usually have the capability to obtain samples from I the reactor coolant system hot legs, the containment recirculation sump, and the containment a'tmosphere. The time required for taking and analyzing samples (in an onsite radiological and chemical analysis facility) must be 3-hours or less from the time a decision is made to sample, except for chloride which is 24-hours or less.

In addition to in-line monitoring systems, backup sampling is require to be available through grab samples. Capability of analyzing these samples must be demonstrated (established planning for analysis at offsite facilities is acceptable to the NRC).

The above regulation also requires that radiation exposures to those individuals performing sampling and analyses be limited to acceptable values. Facility, system, and shielding design must be such that personnel exposure is minimized.

SP/90 Response O

The SP/90 post-accident sampling capability is described in RESAR-SP/90 3

PDA Module 13, " Auxiliary Systems," Subsection 9.3.2.2.2; it is designed to meet applicable requirements such as those contained in NUREG-0737 and Regulatory Guide 1.97.

14. Hydrogen Control .

3 10CFR 50.34(f)(2)(ix) (II.B.8) l

" Provide a system for hydrogen control that can safely accommodate hydrogen generated by the equivalent of a 100 percent fuel-clad t metal-water reaction. Preliminary design information on the tentatively preferred system option of those being evaluated in paragraph (1)(xii) of O

WAPWR-RC 3.1-22 AMENDMENT 3 D050e:1d AUGUST 1989 i

s

[N s /

this section (50.34) is sufficient at the construction permit stage. The hydrogen control system and associated systems shall provide, with reasonable assurance, that:

{')

(A)' Uniformly distribut'ed hydrogen concentrations in the containment do not exceed 10 percent during and following an accident that releases an equivalent amount of hydrogen as would be generated from a 100 percent fuel-clad metal-water reaction, or that the post-accident atmosphere will not support hydrogen combustion.

b(3 (B) Combustible concentrations of hydrogen will not collect in areas where unintended combustion or detonation could cause loss of containment integrity or loss of appropriate mitigating features. ,

(C) Equipment necessary for achieving and maintaining safe shutdown of the plant and maintaining containment integrity will perform its safety function during and after being exposed to the environmental

, conditions attendant with the release of hydrogen generated by the (s) equivalent of a 100 percent fuel-clad metal-water reaction including the e1vironmental conditions created by activation of the hydrogen control system.

(D) If the method chosen for hydrogen control is a post-accident inerting system, inadvertent actuation of the system can be safely accommodated during plant operation."

Discussion U The accident at TMI-2 resulted in a severely damaged or degraded reactor core with a concomitant release of radioactive material to the primary coolant system and a large amount of fuel cladding metal-water reaction in

[' the core with hydrogen generation well in excess of the amounts required to be considered for design purposes by historical Commission regulations. The accident revealed design and operational limitations that existed relative to mitigating the consequences of the accident and j (' determining the status of the facility during and folicwing the accident.

l l l l WAPWR-RC 3.1-23 AMENDMENT 3  !

D060e:1d AUGUST 1989 l 1

s 10CFR 50.44(c)(1) requires that it be shown that during the time period following a LOCA but prior to effective operation of the combustible gas control system either: (A) an uncontrolled hydrogen-oxygen recombination would not take place in the containment, or (B) the plant could withstand the consequences of uncontrolled hydrogen-oxygen recombination without g loss of safety function. If these conditions cannet be shown, the W containment is required to be provided with an inerted or an oxygen deficient atmosphere in order to provide protection against hydrogen i I burning and explosions.

4 For operating plant licensees and applicants prior to the TMI-2 event, the NRC is proposing regulations similar to those contained in 10CFR 50.34(f)  !

(1)(xii) and 10CFR 50.34(f)(2)(ix). The major differences between the proposed rules for existing plants and the effective rules for new plants l are that:

o The uniform hydrogen concentration in the containment must not exceed 10 percent by volume during and following a degraded core accident for new plants.

do not impose such a limit on hydrogen concentration.

The proposed rules for existing plants gi o The amount of hydrogen to be considered for new plants is equivalent to that generated from the reaction of 100 percent (versus 75 percent for existing plants) of the fuel cladding surrounding the actual fuel region.

For new plant designs a suitable hydrogen control system will be required to meet this regulation, whereas no hydrogen control system is needed for large dry containments to meet the proposed regulations of the interim rule for existing plants.

Among the various hydrogen control systems evaluated by the industry thus far, a hydrogen ignition system appears to be the best choice. A hydrogen ignition system is relatively inexpensive, easy to test, and inadvertent actuation of the system during normal plant operation will not result in any adverse effects.

• WAPWR-RC 3.1-24 AMENDMENT 3 UO60e:1d AUGUST 1989

SP/90 Response SP/90 analyses as provided in RESAR-SP/90 PDA Module 16, "Probabilistic Safety Study" demonstrate that for all accident sequences considered, the hydrogen concentration inside containment remains below values that would 9 cause flammable conditions. Nevertheless, a hydrogen ignition system is included in the design of the SP/90 plant in order to accommodate an amount of hydrogen equivalent to that generated from the reaction of 100% 3 J

of the fuel cladding surrounding the active fuel. In addition, the I G containment is designed to prevent hydrogen from concentrating in localized areas. Finally, equipment necessary to achieve or maintain safe shutdown is located such that it is protected to the extent practical from 3

the environmental effects of the operation of the hydrogen ignition system.

15. Reactor Coolant System Valve Testing 10CFR 50.34(f)(2)(x) (II.D.1) 3

" Provide a test program and associated model development and conduct tests to qualify reactor coolant system relief and safety valves and, for PWR's, PORV block valves, for all fluid conditions expected under operating conditions, transients and accidents. Consideration of anticipated transients without scram (ATWS) conditions shall be included in the test program. Actual testing under ATWS conditions need not be carried out until subsequent phases of the test program are developed."

Discussion O General Design Criteria 14, 15, and 30 of Appendix A to 10CFR Part 50 require that the reactor coolant pressure boundary be designed, fabricated, and erected to the highest quality standards and be tested to ensure an extremely low probability of abnormal leakage, rapidly propagating failure, and gross rupture. These criteria also require that the design conditions of the reactor coolant pressure boundary not be exceeded during any condition of normal operation, including anticipated operational occurrences.

WAPWR-RC 3.1-25 AMENDMENT 3 UO60e:1d AUGUST 1989

s Proper operation of the reactor coolant system relief, safety, and block valves is necessary for conformance to these design criteria. The inability of these valves to open or close could lead to a violation of the integrity of the reactor coolant pressure boundary.

When the reactor coolant system relief and safety valves open, the flow through these valves is normally saturated steam. Some reactor coolant system transients and accidents as well as alternate core-cooling methods can result in solid-water or two phase steam-water flow through these valves. Historical qualification requirements for these valves included only flow under saturated steam conditions.

The purpose of this regulation is to require qualification of reactor coolant system relief, safety, and block valves under expected operating conditions (including solid-water and two phase flow conditions) and ATWS conditions.

SP/90 RESPONSE Generic reactor coolant system valve testing (sponsored by EPRI) has been O

conducted in support of operating plant licensees and applicants. The EPRI program included representative testing of Westinghouse reactor coolant system valve types at representative fluid conditions including solid-water and two phase flow conditions. The EPRI program did not, however, include specific consideration of ATWS conditions.

3 The generic EPRI test results discussed above are expected to be directly applicable to the SP/90 design, since the latest Westinghouse pressurizer power-operated relief valves and safety valves were included in the test program.

Westinghouse will document the applicability of the generic EPRI test results to the SP/90 design (including valve designs, piping and support designs, and fluid conditions) at the FDA stage. If the generic EPRI test l

results do not envelope the specific SP/90 design, Westinghouse will WAPWR-RC 3.1-26 AMENDMENT 3 O

UO60e:1d AUGUST 1989

m i

either:..(A) perform additionalJtesting, or (B) demonstrate justification

'for not performing additional testing possibly through additional analyses and/or evaluations.-

16. Valve Position' Indication

[u -

10CFR 50.34(f)(2)(xi) (II.D.3) 3

' "" ' '"' " ' ' '"' ' ""* " ' " ( "'" '

'O closed) in the control room."

Discussion This regulation is written in very general terms. A review of the NRC background: material in, relation to this regulation (i.e.,NUREG-0578, NUREG-0660, NUREG-0718, and NUREG-0737) indicates that this requirement for . valve position indication applies.to reactor coolant system relief and safety valves.

V

-General Design Criterien 14, " Reactor Coolant Pressure Boundary," of Appendix A to 10CFR Part 50 requires that the reactor coolant pressure

, boundary be' designed, fabricated, erected, and tested to have an. extremely low probability of abnormal leakage, rapidly propagating failure, and gross rupture. Historically, the application of this criterion has emphasized the integrity of passive components in the reactor coolant system, such as the reactor vessel and the piping, however, this criterion also applies to the valves that provide isolation for the system. Failure of relief and safety valves to close can cause events that result in small-break loss-of-coolant accidents. Unambiguous indication of the position of the valves can aid the operator to detect a failure and take proper corrective action.

O The purpose of this requirement is to orovide the control room operator a positive indication of valve position and, therefore, provide additional L assurance that the integrity of the reactor coolant pressure boundary can be maintained or a loss of integrity directly diagnosed.

WAPWR-RC 3.1-27 AMENDMENT 3 0060e:1d AUGUST 1989

s SP/90 Response  !

The SP/90 design includes positive control room position indication for the pressurizer power-operated relief valves and safety valves. 3 6

17. Auxiliary Feedwater System Initiation and Indication 3 10CFR 50.34(f)(2)(xii) (II.E.1.2)

" Provide automatic and manuti auxiliary feedwater system initiation, and provide auxiliary feedwater system flow indication in the control room."

Discussion In conventional PWR designs the auxiliary feedwater system (AFWS) is used 3 to remove heat from the reactor system when the main feedwater system is not available.

The need to automatically initiate the operation of the AFWS was not considered by all vendors to be essential to safety in the past, and in some plants dependence was placed on the operator to put the system in service when required.

3 General Design Criterion 13, " Instrumentation and Control," sets forth the requirements for instrumentation to monitor the variables and systems, over their anticipated ranges of operation, that can affect reactor safety. Auxiliary feedwater flow indication to the s' team generators is considered an important adjunct to the manual regulation of auxiliary feedwater flow to maintain the required steam generator level.

S 0 Response 3

l In the SP/90 plant, the function of the AFWS is performed by the combined startup feedwater system (SFWS) and emergency feedwater system (EFWS).

WAPWR-RC 3.1-28 AMENDMENT 3 O'

D060e:1d AUGUST 1989 l

1 l

L  !

w Both systems are normally automatically actuated, but also include the 1, capability for manual.. actuation. Both startup and emergency foedwater j flow .as .well as steam generator . level will be displayed in the Main Control Room by Class 1E indicators in .accordance with. the above 3

regulation and the . guidance of Regulatory Guide 1.97, Revision 3, 1

" Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident"'(refer to item 23 below).

O 18. Pressurizer Heater Power Supplies 3

10CFR 50.34(f)(2)(xiii) (II.E.3.1)

" Provide pressurizer heater power supply and associated motive and control power interfaces sufficient to establish ~and maintain natural circulation in hot standby conditions with only onsite power available."

Discussion Pursuant to NRC regulations in 10CFR Part 50, Appendix A, " General Design Criteria for Nuclear Power Plants," the loss of offsite power is

, considered to be an anticipated operational occurrence, since it is expected to occur one or more times during the life of a nuclear plant.

Following a loss of offsite power, stored and decay heat frem the reactor would normally be removed by natural circulation using the steam generators as.the heat sink. Natural circulation cooling of the primary system requires the use of the pressuMzer to maintain a suitable overpressure on the reactor coolant system. Consistent with satisfying the basic requirements in General Design Criteria 10, 14, 15, 17, and 20 a selected number of pressurizer heaters should be supplied from the emergency power buses.

O 3 O

WAPWR-RC 3.1-29 AMENDMENT 3

.U060s:Id AUGUST 1989

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SP/90 Response l For the SP/90 design one group of pressurizer backup heaters (manually loaded within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />) is sufficient to maintain natural circulation i following a loss of offsite power. To ensure availability of at least one group of backup heaters upon loss of offsite power, emergency power is  !

3 provided from each of the two emergency diesel generators to redundant )

groups of backup heaters.

19. Containment Isolation System 3 10CFR 50.34(f)(2)(xiv) (II.E.4.2)

" Provide containment isolation systems that: (A) ensure all nonessential systems are isolated automatically by the containment isolation system, (B) for each non-essential penetration (except instrument lines) have two isolation barriers in series, (C) do not result in reopening of the containment isolation valves on resetting of the isolation signal, (D) utilize a containment set point pressure for initiating containment isolation as low as is compatible with normal operation, and (E) include automatic closing on a high radiation signal for all systems that provide a path to the environs."

s Discussion -

General Design Criterion 54, " Piping Systems Penetrating Containment," of Appendix A to 10CFR Part 50 requires that piping systems penetrating primary reactor containment be provided with leak detection, isolation, and containment capabilities having redundancy, reliability, and performance capability which reflect the importance to safety of isolating the piping systems. Standard Review Plan 6.2.4, " Containment Isolation System," requires that there be diversity in the parameters sensed for the initiation of containment isolation.

WAPWR-RC 3.1-30 AMENDMENT 3 O .

U060e:1d AUGUST 1989 l l

Some early plants (including' TMI-2) provided automatic containment isolation demand on high containment pressure only. For small rates. of loss of coolant, there would be little pressure increase in the containment, and automatic isolation could be delayed or possibly not occur. The loss of coolant at TMI-2, which produced a small pressure rise in the containment, was' accompanied by substantial core damage and a large release of radionuclides into the containment building. Containment

' isolation was not achieved for some hours after the start of the event.

O The purpose of this requirement is to ensure that effective containment isolation is accomplished and maintained.

SP/90 Response The SP/90 containment isolation system is described in Subsection 6.2.4 of 3 RESAR-SP/90 PDA Module 10, " Containment Systems," It is designed to meet the above requirements.

.f

\. 20. Containment Purging / Venting 10CFR50.34(f)(2)(xv)(II.E.4.4) 3

" Provide a capability for containment purging / venting designed to minimize the purging time consistent with ALARA principles for occupational exposure. Provide and demonstrate high assurance that the purge system will reliably isolate under accident conditions."

t Discussion While the containment purge and vent systems provide plant operational flexibility, their designs must consider the importance of minimizing the release of containment atmosphere to the environs following a postulated loss-of-coolant accident. Therefore, the NRC position is that plant designs must not rely on their use on a routine basis.

O WAPWR-RC 3.1-31 AMENDMENT 3 D060e:Id AUGUST 1989

1 3

The need for purging has not always been anticipated in the design of l 1

plants, and therefore, design criteria for the containment purge system have not been fully developed. The purging experience at operating plants varies considerably from plant to plant. Some plants do not purge during  !

reactor operation, some purge intermittently for short periods, and some s purge continuously. There is similar disparity in the need for, and use of, containment vent systems at operating plants.

Containment purge systems have been used in a variety of ways; for example, to alleviate certain operational problems, such as excess air leakage into the containment from pneumatic controllers, for reducing the airborne activity within the containment to facilitate personnel access during reactor power operation, and for controlling the containment pressure, temperature, and relative humidity. Containment vent systems are typically used to relieve the initial containment pressure buildup caused by the heat load imposed on the containment atmosphere during reactor power ascension, or to periodically relieve the pressure buildup due to the operation of pneumatic controllers.

The sizing of the purge lines in most plants have been based on the need to control the containment atmosphere during refueling operations. This need has resulted in very large lines penetrating the containment (some on the order of 42 inches in diameter). Since these lines are normally the only ones provided that will permit some degree of control over the containment atmosphere to facilitate personnel access, some plants have used them for containment purging during normal plant operation. The NRC is concerned with this situation during a postulated loss-of-coolant accident, since the lines provide an open path from the containment to the environs and calculated accident doses could be significant.

I Therefore, the NRC is currently requiring compliance with Branch Technical Position CSB 6-4, " Containment Purging During Normal Plant Operations."

The following are included as requirements in Branch Technical Position CSB 6-4:

)i WAPWR-RC 3.1-32 AMENDMENT 3 O

UO60e:1d AUGUST 1989 l

j L

[ o The use of large containment purge lines is restricted to cold shutdown and refueling operations-(the lines must be sealed closed in all other operational modes). 3 1

1 O

o Additional smaller purge lines (about 8 inches in diameter or /

smaller) can be provided for continuous purging (lines larger than B inches in diameter must be justified to the NRC).

. 1 A SP/90 Response U

Containment purginJ prior to and during cold shutdown and refueling operations is performed by the SP/90 containment purge system which is described in Subsection 9.4.6 of RESAR-SP/90 PDA Module 13, " Auxiliary Systems." The capacity of this system is equivalent to about two 3 1 containment volumes per hour, which is sufficient to meet ALARA 3 requirements. Control of containment atmosphere during power operation is performed by the containment operating purge system, which includes six-inch supply and exhaust penetrations. Both the containment purge and q

V containment operating purge system meet the requirements of Item 19,

" Containment Isolation System."

21. Specific Accident Monitoring Instrumentation 3

10CFR 50.34(f)(2)(xvii) (II.F.1)

" Provide instrumentation to measure, record and readout in the control room: (A) containment pressure, (B) containment water level, (C) containment hydrogen concentration, (D) containment radiation intensity (high level), and (E) noble gas effluents at all potential accident release points. Provide for continuous sampling of radioactive iodines and particulate in gaseous effluents from all potential accident release points, and for onsite capability to analyze and measure these samples."

i WAPWR-RC 3.1-33 AMENDMENT 3 UOS0e:1d AUGUST 1989

_ _ _ - - _ _ _ _ - - - - - - - - - - a

s Discussion General Design Criterion 13, " Instrumentation and Control," of Appendix A to 10CFR Part 50 requires instrumentation to monitor variables and systems over their anticipate'd ranges for normal operation, for anticipated operational occurrences, and for accident conditions as appropriate to assure adequate safety, including those variables and systems that can effect the containment and its associated systems.

In the past, General Design Criterion 13 had been implemented based on design basis accidents analyzed in Chapter 15.0 of safety analysis reports. Based on conditions experienced at TMI-2, situations can arise which produce containment conditions beyond those postulated for the Chapter 15.0 events.

The purpose of this requirement is to ensure that capability is provided in the control room to ascertain containment conditions during the course of an accident.

SP/90 Response ,

O

, The above required instrumentation is included in Regulatory Guide 1.97, Revision 3, " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident,"

3 and, as such, has been included in the overall SP/90 post-arcident monitoring design that addresses this regulatory guide. Refer to the j discussion of item 23 below.

Control room instrumentation for the SP/90 design as a result of this 3

regulation has been integrated with other control room instrumentation requirements as discussed in item 8 above.

9 l l l I WAPWR-RC 3.1-34 AMENDMENT 3 O I U060e:1d AUGUST 1989 b-- ----

s

22. Inadequate Core Cooling Instrumentation 3

10CFR 50.34(f)(2)(xviii) (II.F.2)

" Provide instruments that provide in the control room an unambiguous l

O~ indication of inadequate core cool kg, such as primary coolant saturation meters in PWR's, and a suitable combination of signals from indicators of coolant level in the reactor vessel and in core thermocouple in PWR's and l BWR's."

Discussion l

l General Design Criterion 13 " Instrumentation and Control," of Appendix A l

to 10CFR Part 50 requires instrumentation to monitor variables for accident conditions ac appropriate to assure adequate safety. In the past, General Design Criterion 13 was not interpreted to require instrumentation to directly monitor water level in the reactor vessel or the adequacy of core cooling. The conventional instrumentation available that could indicate inadequate core cooling includes core exit thermocouple, cold leg and hot leg resistance temperature detectors, in-core neutron detectors, and ex-core neutron detectors. Generally, such instrumentation is included in the reactor design to perform functions other than monitoring of core cooling or indication of vessel water level.

During the TMI-2 accident, a condition of low water level in the reactor vessel and inadequate core cooling existed and was not recognized for a long period of time. This problem was the result of a combination of factors including an insufficient range of existing instrumentation, inadequate emergency procedures, inadequate operator training, unfavorable instrument location (scattered information), and perhaps insufficient instrumentation.

O O WAPWR-RC 3.1-35 AMENDMENT 3

5 The purpose of this requirement is to provide the reactor operator with instrumentation that, together with improved operating procedures and i training, will enable him to readily recognize and implement actions to correct or avoid conditions of inadequate core cooling.

SP/90 Response The above required instrumentation (reactor vessel level instrumentation ,

J system and thermocouple / core cooling monitoring system) is included in Regulatory Guide 1.97, Revision 3, " Instrumentation for Light-Water-Cooled q Nuclear Power Plants to Assess Plant and Environs Conditions During and j Following an Accident," and, as such, it has been included in the overall 3

SP/90 post-accident monitoring design that addresses this regulatory guide. Refer to the discussion of item 23 below.

Control room instrumentation for the SP/90 design as a result of this regulation will be integrated with other control room instrumentation requirements as discussed in item 8 above.

23. Post-Accident Monitoring Instrumentation O

3 , 10CFR 50.34(f)(2)(xix) (II.F.3)

" Provide instrumentation adequate for monitoring plant conditions following an sccident that includes core damage."

Discussion General Design Criterion 13, " Instrumentation and Control," of Appendix A to 10CFR Part 50 requires instrumentation to monitor variables and systems over their anticipated ranges for accident conditions as appropriate to ensure adequate safety. General Design Criterion 19, " Control Room,"

requires that a control room be provided from which actions can be taken 3.1-36 AMENDMENT 3 O

WAPWR-RC UO60e:1d AUGUST 1989

to -maintain the nuclear power ' unit in a safe condition under accident conditions. In addition, General Design Criterion 64, "0nitoring Radioactivity Releaser," requires means for monitoring the reactor containment atmosphere, spaces containing components for recirculation of loss-of-coolant accident fluids, effluent discharge paths, and the plant I environs for radioactivity that may be released from postulated accidents.

The overall subject of adequate post-accident monitoring has been a

. concern of the NRC and the industry for many years. As a result of this initial concern which was amplified in' light of the TMI-2 accident, the NRC has issued guidance in the form of Regulatory Guide 1.97, Revision 3,

" Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident," which describes a method acceptable to the NRC staff for complying with the Commission's requirements to provide instrumentation to monitor plant variables and systems during and following an accident.

SP/90 Response

.O, Westinghouse has developed an interpretation of the requirements necessary I to meet the intent of Regulatory Guide 1.97 for a conventional operating

, plant applicant. The specific variables specified in this design basis interpretation of Regulatory Guide 1.97 are not entirely applicable to the SP/90 design as a result of differences from conventional designs. 3 Using the above mentioned Westinghouse design basis interpretation of  !

Regulatory Guide 1.97 as a starting point, a similar document for the

{ SP/90 has been developed. The results are documented in Subsection 7.5 of RESAR-SP/90 PDA Module 9," Instrumentation & Controls and Electric Power."

{

l Control room instrumentation for the SP/90 design as a result of this )

regulation will be integrated with other control room instrumentation ]

f}

' requirements as discussed in item B above. j WAPWR-RC 3.1-37 AMENDMENT 3 UO60e:1d AUGUST 1989 l

s

24. Power Supplies for Pressurizer Relief Valves, Block Valves, and Level Indicators ,

3 10CFR 50.34(f)(2)(xx) (II.G.1)

" Provide power supplies for pressurizer relief valves, block valves, and level indicators such that: (A)levelindicators are powered from vital buses, (B) motive and control power connections to the emergency power sources are through devices qualified in accordance with requirements applicable to systems important to safety, and (C) electric power is provided from emergency power sources."

Discussion Pursuant to NRC regulations in 10CFR Part 50, Appendix A. " General Design Criteria for Nuclear Power Plants," the loss of offsite power is considered to be an anticipated operational occurrence, since it is expected to occur one or more times during the life of a nuclear plant.

Following a loss of offsite power, stored and decay heat from the reactor would normally be removed by natural circulation using the steam generators as the heat sink. Alternatively, in the event that natural .

circulation in the reactor coolant system is interrupted, the feed and bleed mods of reactor coolant system operation can be used to remove decay heat from the reactor. This method of decay heat removal requires the use of the emergency core cooling system and the pressurizer power-operated relief valves.

Consistent with satisfying the basic requirements in General Design Criteria 10, 14, 15, 17 and 20, the pressurizer power-operated relief g

valves and associated block valves and level indicators must be supplied from emergency power buses.

O WAPWR-RC 3.1-38 AMENDMENT 3 O

UO60e:1d AUGUST 1989

s Nore specific: NRC guidance for implementation of this regulation is-contained in Item II.G.1 of NUREG-0737, " Clarification of TMI~ Action Plan ]

Requirements'." I l

SP/90 Response The SP/90 meets the above requirements, i.e.  !

(A) The four (4) redundant pressurizer level channels are powered by the four (4) redundant vital AC instrument buses.

(B) The. power supplies for the pressurizer power operated ' relief valves and their associates block valves are from Class 1E sources.

3 I (C) The Class' IE sources .in (B) above are backed up by the onsite _

emergency diesel generators.

1 l

O NOTE: [(f)(2)(xxi) through (f)(2)(xxiv) not applicable to SP/90) j i

)

25. Emergency Response Facilities J

10CFR 50.34(f)(2)(xxv) (III.A.1.2) 3

" Provide an onsite technical support center, an onsite operational support center, and, for construction permit applications only, a nearsite emergency operations facility."

Discussion O In addition to the above regulation, Article IV.E.8 of Appendix E, ,

" Emergency Planning and Preparedness for Protection and Utilization O WAPWR-RC 3.1-39 AMENDMENT 3 UO60e:1d AUGUST 1989

s Facilities," to 10CFR Part 50 requires that adequate provisions shall be made and described for emergency facilities and equipment, including a licensee onsite technical support center and a licensee near-site emergency operations facility from which effective direction can be given and effective control

• can be exercised during an emergency. (Notethat g

" effective control" must be interpreted to mean administrative control W versus actual control of the plant). ,

As one would expect, these regulations are quite general in that they l simply require emergency response facilities to be established. The NRC has, however, issued detailed guidance (e.g, functions, locations, size, structures, habitability, communications, instrumentation, etc.) for the of emergency response facilities in the form of NUREG-0737, design Supplement 1, " Requirements for Emergency Response Capability," and NUREG-0696, " Functional Criteria for Emergency Response Facilities."

SP/90 Response 3

The onsite technical support center is included in the SP/90 design. The on-site operational support center and the near-site emergency operations g

facility is the responsibility of each utility utilizing the SP/90 design.

26. Leakage Control Outside Containment 3 10CFR 50.34(f)(2)(xxvi) (III.D.1.1)

" Provide for leakage control and detection in the design of systems

)

outside containment that contain (or might contain) TID-14844 source term radioactive materials following an accident. Applicants shall submit a leakage control program, including an initial test program, a schedule for retesting these systems, and the actions to be taken for minimizing leakage from such systems. The goal is to minimize potential exposures to workers and public, and to provide reasonable assurance that excessive leakage will not prevent the use of systems needed in an emergency."

WAPWR-RC 3.1-40 AMENDMENT 3 O

D060e:1d AUGUST 1989

_ - _ - - - - _ _ _ - _ _ - - _ - _ _ - _ - _ _ ~ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _

O Discussion Qf 10CFR Part 20 and Part 100 specify radiation limits and guidelines that must be met by licensed facilities to assure protection of public health and safety. In a power reactor, many systems that will or may handle

' liquids or gases containing large radioactive' inventories after a serious transient or accident are located outside containment. Several of the engineered safety features and auxiliary systems located outside reactor O

V containment will or may have to function during a serious transient or accident with large radioactive inventories in the fluids they process.

The leakage from these systems, when operated, should be minimized or eliminated to prevent the release of significant amounts of radioactive materials to the environment. Historically, these systems are checked out during preoperational testing and startup testing but are not usually included in any periodic leak testing program. It is beneficial if the plant operating staff knows the leakage rates of these systems and maintains them at rates that are as low as practical.

The purpose of this regulation is to make every effort to eliminate or reduce the leakage from these systems, perform periodic tests to assure that the leakage from these systems is maintained as low as practical, and

. provide the plant staff with current knowledge of the system leakage rates.

SP/90 Response As described in the response to Item 12, the SP/90 Integrated Safeguards System, which is the only system potentially handling large quantities of radioactive fluids in a post-accident scenario, is divided into four 3 4 separate subsystems. Each of these four subsystems is located in a dedicated and separated safeguard component area, each having its own floor drain system. In addition, each recirculation path can be routinely tested because the suction source (i.e., the emergency water storage tank) is located inside containment. '

1 \

O

\

l l

WAPWR-RC 3.1-41 AMENDMENT 3 UO60e:1d AUGUST 1989 I

The above described features allow-for simple and effective monitoring of 3 potential leakage paths on a periodic basis. Detailed procedures to accomplish this will be developed at the FDA application stage.

27. Inplant Monitoring 3 10CFR 50.34(f)(2)(xxvii) (III.D.3.3)

" Provide for monitoring of inplant radiation and airborne radioactivity as appropriate for a broad range of routine and accident conditions." .

Discussion 10CFR Part 20, " Standards for Protection Against Radiation," provides criteria for control of exposures of individuals to radiation in restricted areas, including airborne iodine. Since iodine concentrates in the thyroid gland, airborne concentrations must be known in order to evaluate the potential dose to the thyroid. Historically, the concentration of iodine in atmosphere air has been determined by measuring the activity of iodine adsorbed in a carbon filter through which air has h

been pumped. The charcoal filter is removed from the air pump and allowed to ventilate to permit the noble gases to diffuse to the atmosphere. The filter is then counted for radioactivity content and the remaining activity is ascribed to iodine. This procedure is conservative; however, it is possible for sufficient noble gas to be adsorbed in the charcoal so that the resulting iodine determination may be unduly conservative (high). If the airborne iodine concentration is overestimated, plant personnel may be required to perform operational functions while using respiratory equipment, which sharply limits communication capability and may diminish personnel performance during an accident.  ;

I The purpose of this requirement is to improve the accuracy of measurement

)

of airborne iodine concentrations as well as to ensure adequate inplant l

monitoring of vital areas. i l l g

WAPWR-RC 3.1-42 AMENDMENT 3 UO60e:1d AUGUST 1989

1 SP/90 Response

[v")

The SP/90 design includes sufficient iodine monitors to sample all vital 3 areas. This information is provided in Subsection 11.5 of RESAR-SP/90 PDA (T

w)

Nodule 12. " Waste Management."

28. Control Room Habitability 3

10CFR 50.34(f)(2)(xxviii) (III.D.3.4)

[

" Evaluate potential pathways for radioactivity and radiation that may lead to control room habitability problems under accident conditions resulting in a TID-14844 source term release, and make necessary design provisions to preclude such problems."

Discussion Control room habitability deals with assuring that control room operators .

will be adequately protected against the effects of an accidental release of toxic and radioactive gases and that the plant can be safely operated ,

or shutdown under design basis accident conditions (in accordance with

, General Design Criterion 19 " Control Room," of Appendix A to 10CFR Part 50).

For plants recently designed, this TMI item (in general) has not presented (

a significant problem (beyond software documentation), since the current l NRC guidance and acceptance criteria for ensuring control room habitability was available during the design and licensing of these plants.

SP/90 Response The SP/90 Control Room Habitability System described in Subsection 6.4 of 3 RESAR-SP/90 PDA Module 13, " Auxiliary Systems", is designed to meet the above requirements.

O WAPWR-RC 3.1-43 AMENDMENT 3 UO60e:1d AUGUST 1989

s

29. Industry Experiences 3 10CFR 50.34(f)(3)(i) (I.C.5)

" Provide administrative procedures for evaluating ope-ating, design and construction experience and for ensuring that applicable important i industry experiences will be provided in a timely manner to those designing and constructing the plant."

Discussion OI This requirement deals with administrative procedures which by themselves de not impact any design.

SP/90 Response Westinghouse has always recognized the need to stay appraised of operating events to meet the need for feedback of operating experiences to design, construction, and operation. This has been accomplished through informal methods of screening various media sources (e.g., INPO, Westinghouse site managers daily reports, NRC Inspection and Enforcement Bulletins, Circulars, and Information Notices) for event information. Those events or issues identified as having potential significance are routed internally to appropriate cognizant personnel for their evaluation and follow-up action as necessary.

l The SP/90 design process included the use of operating and construction experiences in the early determination of the systems, components and structures required for increased safe operation of future nuclear power g

3 plants. This information was also utilized in the risk assessment reliability studies given in Volumes 1 and 2 of RESAR-SP/90 PDA Module 16, "Probabilistic Safety Study."

3.1-44 AMENDMENT 3 O

WAPWR-RC UO60e:1d AUGUST 1989 l

I  !

L

'A 30. Quality Assurance List l O l 10CFR 50.34(f)(3)(ii) (I.F.1) 3  !

^ " Ensure that the quality assurance (0A) list required by Criterion II, Appendix 'B, 10CFR Part 50 includes all structures, systems and components i'

important to safety."

p Discussion G l Appendix B " Quality Assurance Criterion for Nuclear Power Plants and Fuel 1 Reprocessing Plants," to 10CFR Part 50 establishes quality assurance requirements for all activities affecting the design, construction, and operation of those safety-related structures, systems, and components that prevent or mitigate the consequences tsf postulated accidents or could cause undue risk to the health and safety of the public. Criterion II of Appendix B further requires the identification of the structures, systems, i and components to be covered by the quality assurance program.

( .

Historically, the requirements of Appendix B have been mostly applied only to safety-related structures, systems, and components (e.g., for a conventional design this encompasses Safety Class 1, 2, and 3 structures,  :

systems,andcomponents). This approach has been (in general) accepted by i the NRC in the past even though Appendix A " General Design Criteria for Nuclear Power Plants," of 10CFR Part 50 requires the establishment of principle design criteria for structures, systems, and components important to safety; that is, structures, systems, and components that ,

provide reasonable assurance that the facility can be operated without undue risk to the health and safety of the public. Structures, systems, and components important to safety can and do include equipment that has been historically classified as Non-Nuclear Safety. A Non-Nuclear Safety p

classification has been translated into a quality assurance program less stringent than that required by Appendix B of 10CFR Part 50.

O WAPWR-RC 3.1-45 AMENDMENT 3 UO60e:1d AUGUST 1989 l

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ i

The purpose of this regulation. is to ensure that appropriate quality i assurance is applied to all structures, systems, and components important to safety versus only those that are safety-related.

SP/90 Response O

Table 3.2-1, " Classification of Structures, Components and Systems," of RESAR-SP/90 PDA Module 7 " Structural / Equipment Design," includes a 0-list 3

which identifies whether a structure, component or system is required to be in compliance with 10CFR50, Appendix B or not. This listing was added .

as part of the 6P/90 quality assurance program. The table, including the quality assurance listing will be reviewed during the FDA stage and will include identifying all equipment important to safety.

31. Quality Assurance Program 3 10CFR 50.34(f)(3)(iii) (I.F.2)

" Establish a quality assurance (QA) program based on consideration of:

(A) ensuring independence of the organization performing checking g >

functions from the organization responsible for performing the functions; (B) performing quality assurance / quality control functions at construction sites to the maximum feasible extent; (C) including QA personnel in the documented review of and concurrence in quality related procedures associated with design, construction and installation; (D) establishing criteria for determining QA programmatic requirements; (E) establishing qualification requirements for QA and QC personnel; (F) sizing the QA staff commensurate with its duties and responsibilities; (G) establishing procedures for maintenance of "as-built" documentation; and (H) providing a QA role in design and analysis activities."

O WAPWR-RC 3.1-46 AMENDMENT 3 O j U060e:1d AUGUST 1989 1

/

l l

l

5

[

Q Discussion i

Various TMI-2 accident investigations and inquiries identified problems relating to the quality assurance organization, authority, reporting, and O

Q inspection. This reg 01ation is a result of NRC actions taken to improve the quality assurance program for design, construction, and operations to provide greater assurance that these activities are conducted in a manner commensurate with their importance to safety.

" SP/90 Response Westinghouse has established and is implementing a quality assurance program (WCAP-8370/7800 (Rev.11/7),"EnergySystemsBusinessUnit/ Nuclear Fuel Business Unit Quality Assurance Plan"), approved by the NRC, that 3

complies with 10CFR Part 50, Appendix B, " Quality Assurance Criterion for Nuclear Power Plants and Fuel Reprocessing Plants," and the considerations listed .in the above regulation. This program currently addresses p Westinghouse design and construction activities and may be revised in the V future to include onsite construction activities and operations.

32. Dedicated Containment Penetrations 10CFR 50.34(f)(3)(iv) (II.B.8) 3

" Provide one or more dedicated containment penetrations, equivalent in size to a single 3-foot diameter opening, in order not to preclude future installation of systems to prevent containment failure, such as a filtered vented containment system."

Discussion As discussed in more detail in Section 3.2, there are rulemaking efforts currently underway to establish policy, goals, and requirements related to accidents involving core damage greater than the present design basis.

O V

WAPWR-RC 3.1-47 AMENDMENT 3 UO60e:1d AUGUST 1989

One of the design requirements being considered in these efforts is the need for a new structure for controlled filtered venting of the reactor containment structure.

The purpose of this requirement is to ensure the capability of installing such a system should it be determined necessary.

SP/90 Response y The SP/90 plant does not include tr.e dedicated penetration called for in >

this requirement. The ratione.le for this omission is as follows:

o The SP/90 severe accident analyses as documented in RESAR-SP/90 PDA Module 16 "Probabilistic Safety Study," do not indicate the need for containment venting.

o If containment venting is to be a feature for future plants, its use is only anticipated for preventing slow, long term (i.e., >24 hours) 3 overpressurization of the containment. g o Long-term overpressurization can be prevented with a vent size much smaller than tha 3 foot equivalent diameter called for in this requirement.

o It is Westinghouse's position that in case containment venting becomes a future requirement, the operating purge exhaust penetration should be used for that purpose. l o The containment penetration of the SP/90 operating purge exhaust system will be designed so as not to preclude its future use as a containment venting path.

In the above manner, the need of adding a potentially unnecessary 1

penetration with associated containment integrity concerns has been eliminated, while maintaining the flexibility to incorporate containment venting at a future date if deemed necessary. i WAPWR-RC 3.1-48 AMENDMENT 3 UO60e:1d AUGUST 1989 I

s 133. Containment Design 10CFR 50.34(f)(3)(v) (II.B.8)

" Provide preliminary design information at a level of detail consistent with that normally required at the construction permit stage of review sufficient to demonstrate that:

(A)(1) Containment integrity will be maintained (i.e., for steel containments by meeting the requirements of the ASME Boiler and Pressure Vessel Code,Section III, Division 1, Subsubarticle NE-3220, Service Level C Limits, except that evaluation- of instability is not required, considering pressure and dead load alone. For concrete containments _by meeting the requirements of the ASME Boiler and Pressure Vessel Code,Section III, Division 2 Subsubarticle CC-3720, Factored Load Category, considering pressure and dead load alone) during an accident that releases hydrogen generated from 100 percent fuel clad metal-water reaction accompanied by either hydrogen burning or the added pressure from post-accident'inerting assuming carbon dioxide is the inerting agent. As a minimum, the specific code requirements set forth above, appropriate for each type of containment, will be met for a combination of dead load and an internal pressure of- 45 psig. Modest deviations from these criteria will be considered by the staff, if good cause is shown by an applicant.

Systems necessary to ensure containment integrity shall also be j demonstrated to perform their function under these conditions.

(2) Subarticle NE-3220, Division 1, and subarticle CC-3720, Division 2, of Section III of the July 1, 1980 ASME Boiler and Pressure Vessel Code, which are referenced in paragraph (f)(3)(v)(A)(1) and (f)(3) (v)(B)(1) of this section, were approved for incorporation by reference by the Director of the Office of the Federal Register. A notice of any changes made to  ;

y the material incorporated by reference will be published in the Federal

\ Register. . . .

lO WAPWR-RC 3.1-49 AMENDMENT 3 UO60e:1d AUGUST 1989 l

i s I (B)(1) Containment structure loadings produced by an inadvertent full actuation of a post-accident inerting hydrogen control system (assuming carbon dioxide), but not including seismic or design basis accident loadings will not produce stresses in steel containments in excess of the limits set forth in the ASME Boiler and Pressure Yessel Code,Section III, I Division 1, Subsubarticle NE-3220, Service Level A Limits, except that evaluation of instability is not required (for concrete containments the loadings specified above will not produce strains in the containment liner in excess of the limits set forth in the ASME Boiler and Pressure Vessel g Code,Section III, Division 2, Subsubarticle CC-3720, Service Load W Category), (2) The containment has the capability to safely withstand pressure tests at 1.10 and 1.15 times (for steel and concrete containments, respectively) the pressure calculated to result from carbon dioxide inerting."

I Discussion The accident at THI-2 resulted in a severely damaged or degraded reactor core with a concomitant releaso of radioactive material to the primary coolant system and a large amount of fuel cladding metal-water reaction in the core with hydrogen generation well in excess of the amounts required to be considered for design purposes by historical Commission regulations. The accident revealed design and operational limitations that existed relative to mitigating the consequences of the accident and determining the status of the facility during and following the accident.

This regulation, in conjunction with the hydrogen control regulations of items 5 and 14, is intended to assure that containment structural integrity is maintained during severe accident conditions.

SP/90 Response The SP/90 containment will be designed to meet the requirements of the relevant sections of the ASME Boiler end Pressure Vessel Codes.

WAPWR-RC 3.1-50 AMENDMENT 3 O

UO60s:1d AUGUST 1989 l

1

_ _ _ _ _ _ . _ )

I 1

34. Hydrogen Recombiners x

i 10CFR 50.34(f)(3)(vi) (II.E.4.1)

L "For plant designs with external hydrogen recombiners, provide redundant dedicated containment penetrations so that, assuming a single failure, the recombiner systems can be connected to the containment atmosphere." .

i Discussion In accordance with 10CFR 50.44, " Standards for Combustible Gas Control System in Light Water Cooled Power Reactors," plant designs since about lete 1970 must include a combustible gas control system (such as recombiners) as the primary means for controlling combustible gases following a loss-of-coolant accident.

Certain plant designs satisfied this requirement with provisions for post-accident installation and operation of an external hydrogen recombiner for combustible gas control. For example, TMI-2 had this external recombiner capability. The design of the external recombiner hookup at TMI-2 used the 36-inch containment penetrations for the normal

. containment purge system by tapping 4-inch lines off the purge lines outside the containment building between the building and the outer containment isolation valves. To place the hydrogen recombiner into service required the opening of the inboard 36-inch containment isolation valve in both a containment purge system inlet and outlet line. With this design, once the hydrogen recombiner is put into operation, containment

( integrity is vulnerable to a single active failure. That is, a spurious or inadvertent opening of one of the 36-inch outboard containment isolation valves would have resulted in the venting of the containment to the environment. In addition, the design of the system to include use of i the large (36-inch) containment purge penetrations resulted in the f} operation of the recombiner beyond the design capacity of the unit.

I

'1 WAPWR-RC 3.1-51 AMENDMENT 3 UO60e:1d AUGUST 1989 i

_ - _ - _ _ _ _ _ _- - a

Since the THI-2 event, the NRC has revised 10CFR 50.44 to also require dedicated containment penetrations for external recombiners.

SP/90 Response This requirement does not apply to Westinghouse designed plants that 3

ine rporate internal hydrogen recombiners.

l l The SP/90 design will include a manually actuated recombiner system which is redundant, qualified, and installed inside containment.

i

35. Management Plan 3 10CFR 50.34(f)(3)(vii) (II.J.3.1)

" Provide a description of the management plan for design and construction activities, to include: (A) the organizational and management structure singularly responsible for direction of design and construction of the proposed plant; (B) technical resources director by the applicant; (C) details of the interaction of design and construction within the h

applicant's organization and the manner by which the applicant will ensure close integration of the architect engineer and the nuclear steam supply vendor; (D) proposed procedures for handling the transition to operation; (E) the degree of top level management oversight and technical control to be exercised by the applicant during design and construction, including the preparation and implementation of procedures necessary to guide the effort."

Discussion One of the major findings as a result of the THI-2 accident was the need to improve staffing to oversee design and construction activities. This regulation is intended to address this finding.

WAPWR-RC 3.1-52 AMENDMENT 3 O

The NRC has issued draft NUREG-0731, " Guidelines for Utility Management O Structure and Technical Resources," which is expected to be used by utilities as guidance in meeting this regulation.

SP/90 Response This regulation is directed at utility management organizational and administrative capabilities and is not applicable to Westinghouse in relation to the SP/90 design.

O O

O O WAPWR-RC 3.1-53 AMENDMENT 3 UO60e:1d AUGUST 1989

i s

3.2 SEVERE ACCIDENT REVIEW AND RELATED CONSIDERATIONS 1

Discussion 1

l( The TM1-2 accident and the results of subsequent reviews and investigations prompted the Commission to reconsider certain aspects of its licensing l policy. One of the conclusions from the post-TMI investigations was that I attention should be given to the probability and consequences of severe

[] accidents (accidents in which substantial damage is done to the reactor core)  !

V and that a policy statement on the acceptance level of risk to the public health and safety was needed.

l In October 1980 the Commission published an advanced notice of proposed rulemaking concerning severe accidents (Severe Accident Design Criteria,

, l 45 FR 65474). Subsequently, it was decided to handle this issue in a policy statement, and a statement entitled " Proposed Commission Policy Statement on l Severe Accidents and Related Views of Nuclear Reactor Regulation" was 3 published for comment in April 1983(48FR16014). After consideration of the comments, the final policy statement entitled " Policy Statement on Severe Reactor Accidents Regarding Future Designs and Existing Plants" was issued on August 8,1985(50FR32138). 1 The policy statement on severe reactor accidents identifies the NRC intentions for rulemaking and regulatory actions for resolving safety issues related to reactor accidents more severe than design basis accidents.

On the basis of currently available information, the Commission has concluded

( that existing plants pose no undue risk to public health and safety; the main focus of the Severe Accident Policy Statement is on approval or certification of new standard designs.

O V

EAPWR-RC 3.2-1 AMENDMENT 3 B423e:1d AUGUST 1989

s The Comission policy affirms that a new design for a nuclear power plant can be shown to be acceptable for severe accident concerns if it meets the following criteria and procedural requirements:

o Demonstration of compliance with current Comission regulations including TMI requirements for new plants (10CFR50.34(f)).

o Demonstration of technical resolution of applicable Unresolved Safety Issues and high and medium priority Generic Safety Issues, with special focus on reliability of DHR system and AC and DC electrical supply systems.

o Completion of a Probabilistic Risk Analysis and consideration of severe accident vulnerabilities the PRA exposes.

3 o NRC staff review, using an approach that stresses deterministic engineering analysis and judgment complemented by PRA.

The Comission initiated efforts to explicitly define an acceptable level of '

risk to the public from nuclear power plant operation in its 1981 Federal Register Notice entitled

• Development of a Safety Goal - Preliminary Policy Consideration". In February 1982 the NRC published a " Proposed Policy Statement on Safety Goals for Nuclear Power Plants" (40FR7023) which included several qualitative safety goals and quantitative probabilistic guidelines for severe accidents.

After several drafts and revisions as a result of comments and recommendations g received by the NRC, on March 14, 1983, the Comission published a " Policy W Statement on Safety Goals for the Operation of Nuclear Power Plants" l (48FR10772) for a two year trial use and evaluation period. The final policy statement, with corrections, was issued August 21,1986(51FR30028).

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WAPWR-RC 3.2-2 AMENDMENT 3 B423e:1d AUGUST 1989

'The safety goal policy statement includas the following qualitative safety goals and quantitative objectives.

e Qualitative Safety Goals

" Individual members of the public should be provided a level of protection from the consequences of nuclear power plant operation such that individuals bear no significant additional risk to life and health."

o " Societal risks to life and health from nuclear power plant operation should. be comparable to or less than the risks of generating electricity by viable competing technologies and should not -be a significant addition to other societal risks."

o Quantitative Objectives "The risk to an average individual in the vicinity of a nuclear power 3 plant of prompt fatalities that might result from reactor accidents should not exceed one-tenth of one percent (0.1%) of the sum of prompt fatality risks resulting from other accidents to which members of the

. U.S. population are generally exposed."

"The risk to the population in the area near a nuclear power plant of cancer fatalities that might result from nuclear power plant operation should not exceed one-tenth of one percent (0.1%) of the sum of cancer fatality risks resulting from all other causes."

The quantitative objectives for determining achievement of the qualitative i safety goals are based on health effects - prompt and latent cancer mortality l risk to an individual in the vicinity (one mile) or to the population in the area (ten miles) of a nuclear power plant.

The possible effects of sabotage or the diversion of nuclear material are not presently included in the safety goals as there presently is no basis on which to provide a measure of risk of these matters.

WAPWR-RC 3.2-3 AMENDMENT 3 5423e:1d AUGUST 1989

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4 Although not a specific performance objective of the safety goal policy, the 9 Commission has proposed a general performance guideline for implementing the safety goal policy, i.e., the overall mean frequency of a large release of radi m ive materials to the environment from a reactor accident should be g less than 1 in 1,000,000 par year of reactor operation. W '

SP/90 Response The SP/90 development program has provided for implementation of a severe accident review in accordance with NRC severe accident and safety goal policies. A demonstration of compliance with current Commission regulations and regulatory guidance is addressed in Section 6.0 of this module. The TMI requirements reflected in the CP rule (10CFR50.34(f)) are discussed in Section 3.1. Technical resolution of Unresolved Safety Issues and medium and high priority Generic Safety Issues, as applicable to SP/90 design, is addressed in Sections 4.0 and 5.0, respectively. The results of a probabilistic risk assessment and a review of severe accident vulnerabilities 3

exposed by the PRA is presented in Module 16. The Probabilistic Safety Study provided in Module 16 provides for implementation of the Safety Goal and g

Severe Accident criteria as discussed below.

While the safety goal policy statement includes the risks of normal operation l

as well as accidents, the risks from routine emissions are considered small compared to the safety goals. Compliance with Federal guidance and regulations, as documented in a plant specific environmental report, will i ensure that these risks are kept at a very low level. The evaluation provided in Module 16 therefore discusses health effects, both prompt and latent cancer mortality risks, from the standpoint of a severe accident and subsequent containment failure resulting in a large release of radioactive material to the environment.

The Safety Goals Policy Statement does not include an explicit core-melt design objective. Nonetheless, the objective of providing reasonable assurance that a severe core-damage accident will not occur et a U.S. nuclear power plant is recognized. An assessment of the SP/90 design, provided in r

WAPWR-RC 3.2-4 AMENDMENT 3 5423e:1d AUGUST 1989

l Nodule 16, shows the core melt frequency to be less than 1 in 100,000 per year of reactor operation for internal events. For purposes of evaluation, the SP/90 PRA postulates that core melt occurs when conditions have been identified in an event tree sequence that might yield core damage, i.e., loss of reactor decay heat removal. Neither recovery of systems nor use of O vmergency non-essential safeguards methods that hypothetically could be l attempted by the operators are addressed. J Nodule 16 includes an evaluation of containment performance and an assessment O- of the frequency of a serious release, i.e., grossly exceeding design basis

' leakage. The containment event tree provides a comprehensive treatment of core degradation and melt progression considerations which impact the j probability of containment failure and the type and amount of radioactivity released. Containment failurcs considered include failure of containment isolation; early, intermediate, and late containment failure; containment

'ypass; and basemat failure.

A quantitative analysis of SP/90 severe accident consequences was performed 3 using radioactive source terms derived from postulated SP/90 accident scenarios. The consequence analysis is limited to airborne pathway of fission product release since previous studies have found the comparative consequences of rainout to fish flesh and of liquid pathways to be small.

l The SP/90 consequence analysis provided in Module 16 was performed assuming the plant is at each of two sites (Salem, New Jersey and Byron, Illinois) which are expected to be representative of possible future sites which might be chosen.

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Estimates of the number of early fatalities, the number whose thyroid whole body and bone marrow limits will be exceeded, and the number of latent cancer fatalities are provided. 200 REM whole body represents the threshold dose for early fatalities as specified ia NUREG-1150, Reactor Risk Reference Document.

The evaluation provided in Module 16 includes a pro'babilistic assessment of the risk to the public from severe reactor accidents and an evaluation of O WAPWR-RC 3.2-5 AMENDMENT 3 B423e:1d AUGUST 1989

...---.__---~_________m___________ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ . _ . _

severe accident vulnerabilities in accordance with the NRC severe accident and safety goal policies. A discussion of the IDCOR issues of severe accident phenomena and modeling considered in the SP/90 risk assessment is included in 3 Nodule. 16. While it is recognized that further information relating to severe accidents may be developed from ongoing NRC programs including the Severe l Accident Source Term Program and the Sever ~ Accident Research Program, these programs are not expected to significantly impact the SP/90 design.

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WAPWR-RC 3.2-6 AMENDMENT 3 5423e:Id AUGUST 1989 l

!i 3.3 OTHER POST TMI ISSUES Discussion The. TMI-2 action plan, NUREG-0660, provides a comprehensive. and integrated listing of issues under coulderation to improve the safety of power reactors based on experience gained from the TMI-2 accident. This includes items that

.were judged to be tamediately necessary for operating reactors or required for p near term operating license applications (later clarified in NUREG-0737) items considered to be necessary for pending construction permit applicants (incorporated into~ a new rule, 10CFR50.34(f), and further clarified in NUREG-0718. " Licensing Requirements for Pending Appilcations for Construction Permits and Manufacturing License") as well as items requiring further development.

The CP rule 10CFR50.34(f), provides for implementation of the TMI Action Plan items required for CP applications pending at the effective date of the rule, February 1982. However, this rule does not address all of the open issues identified in NUREG-0660, some of which requi s research or study before 3 judgments can be made as to whether existing requirements should be modified.

The TMI Action Plan items, including those not yet resolved on a generic basis, have been incorporated into the Generic Safety Issue program.

l NUREG-0933, "A Prioritization of Generic Safety Issues", issued periodically, establishes a prio'rity ranking for open issues based on the potential safety implication and provides the current status of the program for resolution of each issue.

SP/90 Resnonse The NRC " Policy Statement on Severe Reactor Accidents Regarding Future Designs and Existing Plants" (50FR32138 August 8,1985) requires a demonstration of technical resolution of all applicable medium and high priority Generic Safety Issues, which includes the TMI Action Plan items under development from NUREG-0660. A design response for each applicable TMI Action Plan item under development is provided in the following sections.

O l BAPMR-RC 3.3-1 AMENDMENT 3 ,

9131e:Id AUGUST, 1989

_______ _ - - - - _ - - - - - - - - - - - - - - _ - . - - - -.. o

s 3.3.1 NUREG-0737 NUREG-0737, " Clarification of TMI Action Plan Requirements," contains those post-THI requirements that have been approved for implementation by the Commission for operating plant licensees and applicants. In many cases, the specific requirements of NUREG-0737 are not identical to those of However, the NRC has NUREG-0718/10CFR 50.34 discussed in Section 3.1.

determined that certain of the items contained in NUREG-0737 are not applicable at the construction permit stage and are, therefore, not included as requirements in NUREG-0718/10CFR 50.34. Westinghouse believes that the NRC does not intend to imply that certain requirements imposed on operating or near-term operating plants are not applicable to a later vintage plant, but simply that certain requirements can be more appropriately addressed at the operating license stage. Therefore, the SP/90 design will include appropriate '

consideration of the additional requirements of NUREG-0737.

1. Pressurizer Water Level (NUREG-0737. Item II.K1.17)

Discussion This item is really applicable to certain older generation Westinghouse l

, operating plants that utilized a low pressurizer level coincident eith low pressurizer pressure logic to provide a safety injection signal. This design feature is not utilized in current-day Westinghouse designs. ,

l SP/90 Resoonse f This design feature has not been included in the SP/90 design.

2. Thermal Mechanical Report - Effect of High Pressure Injection on Vessel Integrity for Small-Break LOCA with no Auxiliary Feedwater (NUREG-0737, Item II.K.2.13)

)

AMENDMENT 3 el.

BAPWR-RC 3.3-2 AUGUST, 1989 9131e:1d

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Discussion Refer 'to the discussion of Unresolved Safety Issue A-49, " Pressurized Thermal Shock," in Section 4.0 (item 27).

O O' 3. Installation and Testing of Automatic Power-Operated Relief Valve Isolation System (NUREG-0737. Item II.K.3.1) g Discussion U

Refer to the discussion of item 4 in Section 3.1.

.4. Automatic Trip of Reactor Coolant Pumps During LOCA (NUREG-0737 Item II.K.3.5)

Discussion For this item the NRC considered a requirement for plant designs to incorporate automatic tripping of the reactor coolant pumps in the case of a small-break LOCA. This item and its impact on the SP/90 design is fully discussed in Section 6.4, Items 92 and 93.

'5 . Emergency Preparedness (NUREG-0737 Item III.A.2)

Discussion Refer to the discussion of 10CFR Part 50, Appendix E. " Emergency Planning

,q and Preparedness for Production and Utilization Facilities," in Section k/ 6.1.1 (item 1).

O HAPHR-RC 3.3-3 AMENDMENT 3 9131e:1d AUGUST. 1989

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s 3.3.2 NUREG-0660 ]

1 l l Shortly after the initial recovery phases following the March 28, 1979 l incident at Three Mile Island Unit 2 (THI-2), various task forces and I investigating groups were s'et up, both inside and outside of the NRC, to make recommendations for plant design and operating changes to ensure that a THI-2 type event or similar event does not happen again. The requirements and recommendations from these task forces and investigating groups were consolidated and documented in NUREG-0660, "NRC Action Plan Developed as a Result of the THI-2 Accident," first published in May 1980. NUREG-0737,

" Clarification of THI Action Plan Requirements," issued in November 1980, provided further guidance regarding the THI-2 action plan items that were approved for implementation.

Subsequently, the NRC issued NUREG-0718, Revision 2, " Licensing Requirements for Pending Applications for Construction Permits and Manufacturing License,"

which specifies those NRC Action Plan (NUREG-0660) items that are required by a pending applicant prior to receiving a construction permit. In addition, the NRC has issued a revision to 10CFR 50.34, " Contents of Applications; 3

Technical Information," that essentially incorporates the post-THI requirements of NUREG-0718 into their regulations. Further development of THI

. action Plan (NUREG-0660) items which have not been resolved on a generic basis is addressed in the NRC Generic Safety Issue program, NUREG-0933.

The following discussions pertain to high and medium priority THI Action Plan items under development, in relation to the SP/90 design.

Issue I.A.4.2(4): Long Term Simulator Upgrade-Review Simulators for Conformance to Criteria  ;

Discussion Nuclear power plant simulators are recognized as an important part of reactor operator training. The TMI Action Plan, NUREG-0660, called for a number of t actions to impreve simulators and their use. There is significant interaction  !

{

among the simulator-related action items and clear separation is difficult. J BAPHR-RC 3.3-4 AMENDHENT 3 9131e:1d AUGUST, 1989

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Item I.A.4.2; has a number of components dealing with long-term upgrades.- The NUREG-0660 description calls for research to improve the use of simulators in D. - training operators, develop guidance on the need .for and nature of operator action during accidents, and gather data on operator performance. Specific research items mentioned. include simulator capabilities, safety-related O operator action, and simulator experiments. The item also calls for the upgrading of training simulator standards. The final portion of Item -I.A.4.2 I cal's for a review of simulators to assure their conformance to the criteria. 1 A: significant portion of the. activities to be conducted under this action plan l item has been completed. For example, ANSI /ANS 3.5 was revised and issued in 1981. The regulatory guide endorsing this standard, Regulatory Guide 1.149, Nuclear Power Plant Simulators for Use in Operator. Training," as well as  ;

numerous research reports have been pubitshed.

Item I.A.4.2(4) concerns the long-term training simulator improvement criteria which were also established in Regulatory Guide 1.149. However, the review of submittals from simulator owners for conformance with the criteria is an on-going task which is still not complete. Therefore, the outstanding 3 portions of this issue that have yet to be completed are the continuation of

. simulator research and the review for conformance to acceptability criteria.

'SP/90 Resnonse

. This issue is related to criteria for establishing the acceptability of training' simulators and is not applicable to the SP/90 design scope.

Simulator capability is the responsibility of each utility utilizing the SP/90 design.

Issue I.D.3: Safety System Status Monitoring Discussion This TMI Action Plan item recommends that a study be undertaken to determine the need for all licensees and applicants not committed to Regulatory 1.47 to install a bypass and inoperable status indication system or similar system.

HAPHR-RC 3.3-5 AMENDMENT 3 AUGUST, 1989 9131e:1d m-_____-__2_ _ _ _ _ - _ _ _ _ _ _ _ _ - _ _ -

s 10CFR 50.55a(h) requires that protection systems meet the requirements set forth in IEEE Standard 279, " Criteria for Protection Systems for Nuclear Power g

Generating Stations." Section 4.13 of IEEE Standard 279 requires that, if the protective action of some part of the protection system has been bypassed or deliberately rendered inoperative for any purpose, this fact shall be continuously indicated in the control room. Regulatory Guide 1.47 describes an acceptable method of complying with the requirements of IEEE Standard 279.

Implementation of a well-engineered bypass and inoperable status indication system could provide the operator with timely information on the status of the plant safety systems. This operator aid could help eliminate operator errors such as those resulting from valve misalignment due to maintenance or testing errors.

If the system is integrated with the overall control room, then it could be expected that it would reduce operator error, which in turn will lower the risk associated with operation of the monitored safety systems.

3 Plants not yet licensed or undergoing licensing are committed to Regulatory Guide 1.47. Supplemental guidance for the implementation of the recommendations of Regulatory Guide 1.47 is provided in Branch Technical

, Position ICSB 21.

SP/90 Resoonse The Bypassed and Inoperable Status Indication (BISI) System hardware has been incorporated in the integrated I&C architecture described in Chapter 7. The bypassed and inoperable information necessary to meet the intent of Regulatory Guide 1.47, " Bypassed and Inoperable Status Indication for Nuclear Power Plant Safety Systems," has been integrated into the plant display system. A physically independent system is not being supplied. Detailed design drawings of the plant display system will not be available until the final design phase. The drawings will not be specific to BISI.

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HAPHR-RC 3.3-6 AMENDMENT 3 AUGUST, 1989 9131e:1d

$g All protection system and systems . actuated or controlled by the protection O system will be ' automatically indicated if inoperability was intentionally induced or . the system . bypassed. Also included will be those systems which directly support automatically initiated systems but which themselves may not be automatically initiated because they are normally in the operating mode.

V Related support systems may have their own subsystem bypass and inoperability

' indication as.. well as . input to the primary system indication, i.e., status information for support systems will be pyramided up to all related primary safety systems.

The information available. to the operating staff with respect to inoperability and bypass is at least as comprehensive as that of an independent system.

Both overall system. state (i.e., the ability to perform) and actuation status (i.e., failure to perform) will be available. These capabilities for each of these purposes will take the form of binary lights for the ISS. The alarm system will drive these lights based on its results in recognizing .the completion or lack of completion of the patterns for the " readiness" and for the " active" system status. The status of the systems not having these lights will be available by accessing system mimics of the display system.

The design philosophy of that portion of the display system that fulfills the requirements for a BISI System is the same as for the remainder of the display system. This approach to display bypassed and inoperable status is part of the overall functional decomposition performed to determine what information is provided in the control room and the manner in which it is displayed. High level goals are identified and then decomposed into system level indications, component level summaries, and support system status. With the BISI requirements incorporated at the conceptual level of the display system in the O integrated system, the same operator mental model and display rules are used throughout the system, minimizing the potential for operator confusion and errors.

That equipment rendered inoperable for maintenance less frequently than once per year will not necessarily be automatically indicated. The display system will have the capability to indicate manual initiation of bypass of safety features on a system level. Under administrative control, manual bypass O HAPWR-RC 3.3-7 AMENDMENT 3 AUGUST, 1989 9131e:1d

s indication can be input or removed. The automatic indication feature cannot be overridden by operator action.

3 The BISI System, while not a Class 1E system, will not degrade the Class IE systems with which it interfaces if subjected to a credible event. The isolation provisions it uses to satisfy this requirement will meet Class 1E '

criteria.

Issue I.D.4: Control Room Design Standard Discussion This issue emphasizes a need for guidance on the design of control rooms to incorporate human factor considerations.

Control rooms and control panels which incorporate human factor considerations

.can greatly enhance operator performance. This could contribute to a reduction in operator error and, therefore, a potential reduction in the frequency of core-melt accidents. A possible solution for this issue could be the . development of an NRC Regulatory Guide endorsing industry standard (s) with the intention of providing: (1) guidance for the design of control rooms and

. remote shutdown panels, and (2) the evaluation criteria for use in use in the licensing process.

3 Subsequent NRC review indicated that all operating and NTOL plants are conducting detailed control room design reviews in response to the THI Item I.D.1 NUREG-0700, and acceptable substitutes are being used as control room

- design standards for this effort. Applications for future light water reactors shall include per 10 CFR 50.34(G), an evaluation of the facility against the SRP Section 18.1 which addresses control room design and references NUREG-0700 as appropriate to guidance for control room design.

This issue has been resolved by industry compliance with TMI Item I.D.1 and 10 CFR 50.34(G) and the revision of SRP 18.1 to include reference to NUREG-0700.

Thus, staff actions have negated the need for evaluation of industry control room design standards and for the development of a Regulatory Guide endorsing HAPWR-RC 3.3-8 AMENDMENT 3 AUGUST, 1989 9131e:1d

4 those E standards. NUREG-0700 ' and acceptable . substitutes are the de facto I- control ' room d6 sign standards for evaluating commercial nuclear power plants

.in the . United States. Design standards for advanced ' control ras will be addressed as a research _ issue' under the Human Factors Research Program.

Therefore, this issue was RESOLVED and no new requirements were established.

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52/90 Response 3

Chapter 18 of RESAR-SP/90 PDA Module 15 " Control Room / Human Factor, Engineering," discusses in detail the Westinghouse control room design process to be utilized for the SP/90.

l Issue I.D.5(1): Operator-Process Communication l

Discussion Issue I.D.5(1) was documented in the TMI Action Plan and focused on the need to evaluate the operator machine interface in reactor control rooms. The ,

emphasis of this portion'of the overall issue was the use of lights, alarms, 3 and annunciators.

The method of presentation of information can significantly enhance the performance of the control room operators and thereby potentially affect operator error.

It 'was proposed that current practice and use of lights, alarms, and annunciators be reviewed to assess how well they facilitate operator-machine interaction and minimize errors. NRC has studied the area of control room O alarms and annunciators and the results were reported in NUREG/CR-2147.

on this report, NRC issued a Research Information Letter (RIL-124) which Based provided a recommendation for further action.

This item was RESOLVED and no new requirements were established.

i O HAPHR-RC 3.3-9 AMENDMENT 3 l AUGUST, 1989 l 9131e:1d 4

Q SP/90 Resoonse ,

The design process for the SP/90 control room placed heavy emphasis on human j factors engineering methodology. Westinghouse employed a man-machine l I

interface design philosophy that analyzes the complete range of control tasks i

and then allocates tasks between operator and automatic systems to better ensure optimization of plant safety and power production. RESAR-SP/90 PDA l l Module 15 " Control Room / Human Factors Engineering", describes these methods in detail.

l

' Issue I.D.5(2): Plant Status and Postaccident Monitoring Discussion This issue was documented in the THI Action Plan and focused on the need to improve the ability of reactor operators to prevent, diagnose, and properly respond to accidents. The emphasis was on the information needs (i.e.,

indication of plant status) of the operator.

In order for the operators to perform their functions it is necessary that they receive all the necessary information on the plant status. This can Anhance operator performance (and therefore reduce operator error).

Accident sequences should be analyzed to determine the information required to provide unambiguous indication of plant status. Specific instrumentation and ESF status monitoring needs would then be determined. PHR instrumentation requirements were analyzed in NUREG/CR-1440 and BHR instrumentation require-ments were analyzed in NUREG/CR-2100. ESF Status Monitoring requirements were also studied in NUREG/CR-2278. Research Information Letter (RIL) No. 98 was g

issued in August 1980. This RIL transmitted "the results of completed research describing an improved method for analyzing accident sequences."

Revision 2 to Regulatory Guide 1.97 was issued in December 1980. (See also Item II.F.3, " Instrumentation for Monitoring Accident Conditions.") Present plans include implementation of this guide at all plants.

O MAPHR-RC 3.3-10 AMENDMENT 3 AUGUST, 1989 9131e:1d

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This item was RESOLVED and new requirements were established.

SP/90 Resnonse Section 7.8 of RESAR-SP/90 PDA Module 15 " Control Room / Human Factors l Engineering," describes the main control board, the supervisory console, and all other panels located in the primary operating area.

The SP/90 control room design has emphasized improvements in the area of plant status indication and accident monitoring. Use of improved plant process data systems, parameter display systems and alarm systems should enhance operator performance and reduce operator error.

Issue I.D.5(3): Improved Control Room Instrumentation Research, On-Line Reactor Surveillance System l

Discussion 3

The objective of this issue is to perform research to determine the O- feasibility of detecting and diagnosing nuclear power plant operating anomalies using a continuous on-line noise surveillance system and to perform demonstrations of such a system in an NRC-licensed commercial plant.

Item I.D.5(3) was documented in the TMI Action Plan based on the work being performed by Oak Ridge National Laboratory. A continuous on-line automated surveillance system was installed at Sequoyah-1 (PWR) and information has been obtained throughout the first fuel cycle.

The demonstration at Sequoyah has been completed. A similar demonstration at an operating BHR is underway. The system has the potential to provide diagnostic information to produce anomalous behavior of operating reactors which could be used to maintain safe conditions.

O HAPWR-RC 3.3-11 AMENDMENT 3 AUGUST, 1989 9131e:Id

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Noise surveillance and diagnostic techniques associated with the on-line reactor surveillance system have shown their safety significance and the results of the research have been and are being used by NRC in various regulatory activities.

9, '

Based on the ongoing programs, NRC staff concludes that the technical resolution of this issue has been identified.

SP/90 Response The SP/90 design includes a digital metal impact monitoring system (DMIMS) which is detailed in Subsection 4.4.6.4 of RESAR-SP/90 PDA Module 4 " Reactor Coolant System." The DMIMS is not a Class IE system, but serves as a diag-nostic aid to reliably detect loose parts in the reactor coolant system before

< damage can occur. This system is not provided for its safety significance, "but to improve plant availability. There are other non-safety systems avall-

-able for use, such as the acoustic leak detection system and neutron-noise monitoring system, which are not currently included in the SP/90 design.

Westinghouse will consider the implementation of additional monitoring systems for inclusion in the final design of the SP/90 and will respond to any edditional regulatory requirements which may be promulgated in the resolution of this issue.

Issue I.F.1: Expand Quality Assurance List Discussion The objective of task I.F is to improve the quality assurance program for design, construction, and operations to provide greater assurance that plant design, construction, and operational activities are conducted in a manner commensurate with their importance to safety.

The THI Action Plan has identified that, "Several systems important to the safety of THI were not designed, fabricated, and maintained at a level equivalent to their safety importance. They were not on the Quality Assurance 3.3-12 AMENDHENT 3 HAPHR-RC AUGUST, 1989 9131e:1d

i s-

-(@) List ' for the plant.

This condition exists at other plants and results primarily from the lack of clarity in NRC guidance on graded protection. One 4 of the difficulties in establishing a- QA list based on safety importance is the absence of relative risk assessments to equipment."

"NRC will develop guidance for licensees to expand their QA lists to cover equipment important to safety and rank the equipment in order of its importance to safety. Experience in use of the revised NRR review procedure

'for developing QA lists for individual operating license applicants will also be factored into the genaric guidance to be developed and when determining backfit requirements. (There is a task presently underway to define the applicability of 10 CFR 50, Appendix B to 10 CFR 50, Appendix A required equipment.)"

The principal benefits to be derived from the expanded QA list is the knowledge that adequate guidance is provided the plant owner to establish quality assurance programs and requirements which are commensurate with the safety importance of the structure, system, and components as determined from 3

completed risk assessments. Currently, the quality assurance requirements are applied principally to structures, systems, and components that prevent or mitigate the consequences of postulated accidents that could cause undue risk to the health and safety of the public (10 CFR 50, Appendix B). This guidance will not only result in the inclusion or addition of other systems important to safety to the plant owners QA list which previously were excluded but will also aid in clarifying the quality assurance level of effort which is deemed necessary.

SP/90 Resnonse O Table 3.2-1, " Classification of Structures. Components and Systems," of  ;

RESAR-SP/90 PDA Module 7 " Structural / Equipment Design," includes a Q-list I

which identifies whether a structure, component or system is required to be in

compliance with 10CFRSO, Appendix B or not. This listing was added as part of the SP/90 quality assurance program. The table, including the quality assurance listing will be reviewed during the FDA stage and will include identifying all equipment important to safety.

HAPHR-RC 3.3-13 AMENDMENT 3 AUGUST, 1989 9131e:1d

Issue II.C.1: Interim Reliability Evaluation Program The objective of Task II.C is to develop and implement improved systems-oriented approaches to safety review. The NRC will employ risk-assessment methods to identify particularly high-risk accident sequences at individual plants and determine regulatory initiatives to reduce these high-risk sequences. .

The Interim Reliability Evaluation Program (IREP) is a planned multiple reliability evaluation program to develop and standardize the reliability methodology involved in performing reliability and safety type studies of this ,

depth. This program was conceived in NUREG-0660 as a pilot study and then a f scaled-up study of an additional 6 plants.

I

.This issue is concerned with the completion of a shortened 5-plant version of  !

the IREP program. To date, the pilot study has been completed on the Crystal River plant and the results reported in NUREG/CR-2515. Scaled-up analyses have been completed on 4 other plant's and the results reported on 2 plants so far: Arkansas Nuclear One-Unit One (NUREG/CR-2787) and Browns Ferry Unit One 3

(NUREG/CR-2802). Remaining is one additional plant, probably a Mark II BWR-type plant. To be included in this analysis would be other comman cause

< initiators, e.g., fires, seismic events, and floods, which were considered in f

(

the other IREP analyses.

Based on the value/ impact score, this issue would have received a medium I priority ranking. However, given the potential public risk reduction, it was deemed to have a high priority. Work completed by the staff resulted in the publication of the following reports for the two remaining plants:

NUREG/CR-3085 and NUREG/CR-3511 for Millstone Unit I and Calvert Cliffs l Unit 1, respectively. A primary output of the IREP was NUREG/CR-2728 which is a guide that documents methods, codes, and data used in the IREP. This guide is intended to provide guidance for PRAs performed subsequent to IREP. Thus, this item was RESOLVED and no new requirements were established. l O'

3.3-14 AHENDHENT 3 HAPHR-RC AUGUST, 1989 9131e:1d

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l SP/90 Resnonse This issue applies- more to operating plants and the completion of plant specific Probabilistic Risk Assessment (PRA) as part of a reliability c program. However, an internal events PRA was performed for. the SP/90 plant design and is documented in RESAR-SP/90 PDA Module 16. "Probabilistic Safety Study." The PRA was employed as an initial design tool to determine the most reliable system design and resulted in several modifications to the overall standard plant systems which are utilized in current operating plants.

O For the Final Design Approval Application, a complete Level 3 PRA will be l

performed and will employ state-of-the-art reliability methodology.

Issue II.C.2: Continuation of Interim Reliability Evaluation Program IREP is a planned multiplant reliability evaluation to develop and standardize the reliability methodology involved in performing reliability and safety-type studies of this depth. It was conceived in NUREG-0660 that a National 3 Reliability Evaluation Program (NREP) study, performed by the plant owners, O should follow the IREP effort.

This issue is concerned with the continuation of the IREP program to cover all the remaining operating reactors which were not covered in the initial IREP studies either performed by NRC or performed by plant owners. Also, consideration is being given to include plants under design or construction.

Possible solutions to this issue may range from the NRC sponsoring an analysis of all plants, having the individual utilities perform an analysis on all or O some plants, or reducing the effort to a limited type study. The plan selected for this analysis consists of three parts: (1) performance of an NREP by the plant owners on 4 plants currently without a risk / reliability type analysis, (2) a careful review by the NRC of 7 other plants that currently have an existing PRA, and (3) an appraisal of the interim results of these reviews a year after implementation to consider the advisability of future extension of the NREP program to other plants. These 11 plants would be the same ones chosen for the first group of SEP Phase III plants.

HAPHR-RC 3.3-15 AMENDHENT 3 9131e:1d AUGUST, 1989

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Although the value/ impact score of this issue would only warrant a medium priority ranking the large potential risk reduction (brought about by the reduction in core-melt frequencies for those plants that are above 10-4/RY) l indicated that a high priority ranking should be assigned.

Work completed by the staff on this item was closely related to the l accomplishments under Item IV.E.5. Whereas Item II.C.2 called for the initiation of IREP studies (i.e., plant-specific PRAs) on all remaining l operating reactors, Item IV.E.5 called for the development of a plan for the systematic assessment of the safety of all operating reactors. The Integrated Safety Assessment Program (ISAP), presented in SECY-84-133 and SECY-85-160, j l

provided for a comprehensive review of selected operating reactors to address all pertinent se.fety issues and to provide an integrated cost-effective implementation plan for making needed changes. Under ISAP, each plant would be subject to an integrated assessment of safety topics, a probabilistic safety assessment, and an evaluation of operating experience.

NRC guidance, as described in the Severe Accident Policy Statement (see Item 3 II.B.8), states that OLs will be expected to perform plant-specific PRAs in order to discover instances of particular vulnerability to a core-melt or poor containment performance, given a core-melt. Thus, this item was RESOLVED and

,no new requirements were established.

SP/90 Resconse RESAR-SP/90 PDA Module 16, "Probabilistic Safety Study," describes the methodology, analysis and results of an internal events only PRA for the SP/90 design. A complete Level 3 PRA will be performed at the time of an FDA application.

Issue II.C.4: Reliability Engineering

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HAPWR-RC 3.3-16 AMENDMENT 3 AUGUST, 1989 9131e:1d j

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s Discussion There is currently no requirement for the plant owners to develop and implement a reliability assurance program. Such programs, typically, determine system ava11 abilities, identify high component failure rates, d determine basic causes for component failures, identify possible corrective actions, and perform other similar activities in what is generally called reliability engineering. In the absence of a requirement, it is difficult to determine the nature and extent that is now being exercised by the plant owners to implement a reliability assurance program. This issue was identified in the THI Action Plan as the final item of Task II.C.

The technical issue at the time the THI Action Plan was published was that the essential elements and process of a reliability program applicable to operational safety had not yet been identified. Although NRC requirements, such as Appendices A and B of 10 CFR 50, strongly reflect reliability principles (i.e., safety margins, redundancy, diversity, and corrective action), these principles had been applied to nuclear power plants primarily 3 in the design phase and not in the operations phase. Reliability engineering practices at nuclear power plants had not yet resulted in strategies to help achieve and maintain the ' designed-in' capability for reliable operation during the operating lifetime of the plants.

The concept of an operational reliability program is based on a simple closed-loop strategy: monitoring and evaluating plant performance, identifying and prioritizing potential problems, diagnosing the causes, taking appropriate corrective actions, and verifying the effectiveness of these l

actions. The elements of a reliability program were summarized by the staff in RIL 158, " Operational Safety Reliability Program," which was issued in l September 1988. The elements of a reliability program are included among recent NRC initiatives to improve maintenance and better manage the effects of aging, to improve technical specifications, and to develop and use plant 5 performance indicators. Also, an operational reliability program that is 'an acceptable means of meeting the station blackout rule (10 CFR 50.63) will be described in Revision 3 to Regulatory Guide 1.9 as part of the proposed resolution of Issue B-56, " Diesel Generator Reliability."

3.3-17 AMENDHENT 3 HAPHR-RC AUGUST, 1989 9131e:1d

s Based on the above, the staff concluded that the sr.fety concern of this issue was addressed in other NRC programs and the issue was considered RESOLVED in October 1988 with no new requirements.

SP/90 Resconse This issue is associated with the implementation of a reliability assurance program for use during plant operation and is not applicable to Westinghouse in regard to SP/90 design.

Issue II.E.4.3: Containment Integrity Check Discussion The objective of task II.E.4 is to improve the reliability and capability of nuclear power plant containment structures to reduce the radiological consequences and risks tc the public from design basis events and degraded-core and core-melt accidents.

THI Action Plan Item II.E.4.3 proposes a requirement for the performance of a feasibility study to evaluate the need and possible methods for performing a periodic or continuous test to detect unknown gross openings in the containment structure. A prime example of the type of operational error this issue is directed at is the incident at the Palisades plant. At Palisades, the reactor was operated for about 1.5 years while the containment isolation valves in a purge system bypass line were unknowingly locked in the open position.

Should a LOCA resulting in major fuel damage occur in a plant which has an undetected breach in the containment building, severe offsite exposure would be expected.

Issue II.E.4.3 deals with containment leakage during postulated (i.e., design basis) accidents and does not address the issue of containment integrity and associated radiation consequences during severe accidents. This last issue is O

HAPHR-RC 3.3-18 AMENDMENT 3 9131e:1d AUGUST, 1989

s being addressed as part of implementation of the Commission's policy on severe accidents and, more specifically, in the Individual Plant Examination. (IPE) and ' Containment Performance Improvement programs. Thus, this issue was RESOLVED and no new requirements were established.

\ SP/90 Response i

This issue is concerned with periodic inspection and testing associated with containment integrity and is not applicable to SP/90 design. The utility will (q/ comply with applicable NRC requirements.

Issue II.E.6.1: In Situ Testing of Valves, Test Adequacy Study ,

1 Discussion-The objective of task II.E.6 is to evaluate whether current requirements for valve testing provide adequate assurance of performance under design conditions. 3 0 The purpose of TMI Action Plan item II.E.6.1 is to establish the adequacy of 1

current requirements for safety-related valve testing. It recommends a study which would result in recommendations for alternate means of verifying performance requirements.

1 Valve performance is critical to the successful functioning of a large number of the plants' safety systems.

It ce91d be assumed that a study would be conducted for both PHRs and BWRs and O that it could result in recommendations for additional testing and/or maintenance on all safety-related valves. A program to implement the recommendations would then be required at all plants.

SP/90 Resnonse This issue is associated with in situ testing of valves and is not directly l

applicable to Westinghouse in relation to SP/90 design.

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Issue II.F.3: Instruments for Monitoring Accident Conditions (NUREG-0660) l Discussion l

The objective of this task is to provide instrumentation to monitor plant variables and systems during and following an accident. Indications of plant i variables and status of systems important to safety are required by the plant operator (licensee) during accident situations to: (1) provide information needed to permit the operator to take preplanned manual actions to accomplish safe plant shutdown; (2) determine whether the reactor trip, engineered safety features systems, and manually-initiated systems are performing their intended functions (i.e., reactivity control, core cooling, maintaining reactor coolant j system integrity, and maintaining containment integrity); (3) provide information to the operator that will enable him to determine the potential

,for a breach of the barriers to radioactivity release (i.e., fuel cladding, reactor coolant pressure boundary, and containment) and if a barrier has been breached; (4) furnish data for deciding on the need to take unplanned action if an automatic or manually-initiated safety system is not functioning 3 properly or the plant is not responding properly to the safety systems in operation; (5) allow for early indication of the need to initiate action necessary to protect the public and for an estimate of the magnitudt of the jmpending threat; and (6) improve requirements and guidance for cla'ssifying nuclear power plant instrumentation control and electrical equipment important to safety.

After the THI-2 event, Task II.F of the THI Action Plan addressed several concerns regarding the availability and adequacy of instrumentation to monitor plant variables and systems during and following an accident.

Prior to the THI-2 event, Regulatory Guide 1.97 " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident," (August 1977) had been used as guidance during licensing reviews. Item II.F.3 called for this regulatory guide to be up' dated to include the THI-2 concerns.

O MAPHR-RC 3.3-20 AMENDHENT 3 9131e:ld AUGUST, 1989

4 Revision 2 of Regulatory Guide 1.9{ was published in December of 1980 and implementation is being carried out as discussed in SECY-82-111 and a letter issued to all licensees of operating reactors. This item was RESOLVED and new requirements were established.

SP/90 Response Subsection 7.5 of RESAR-SP/90 PDA Module 9. " Instrumentation & Controls' and Electric Power," describes in detail instrumentation, important to safety, employed by the operator for monitoring conditions in the reactor coolant system, the secondary heat removal system and ' the containment including engineered safety functions and the system employed for a safe shutdown condition.

The SP/90 control room design described in Subsection 7.8 and Section 18.0 of RESAR-SP/90 PDA Module 15. " Control' Room / Human Factors Engineering," provides a discussion of the SP/90 alarm system which has the attributes of a safety parameter display system (SPDS).

3' O It is stated throughout the aforementioned subsections that the SP/90 design for the control room, instrumentation monitoring system and alarm systems meet the requirements of Regulatory Guide 1.97, Revision 3. Any deviations from l

l 'the R.G. 1.97 positions will be documented during the FDA phase of the licensing process.

Issue II.H.2: Obtain Technical Data on the Conditions Inside the TMI-2 Containment Structure Discussion The objectives of task II.H.2 is to obtain and factor into regulatory programs g safety-related and environmental information from the THI-2 cleanup.

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Pertinent techr.ical information is to be obtained on the conditions of the TH1 facility as cleanup operations proceed. The information to be gathered and disseminated is divided into two distinct categories; i.e., ' data to be l obtained prior to gaining access to the primary system and that to be obtained )

after ' access to the primary system. In the first category, information will be obtained and developed in instrumentation and electrical equipment survivability under the accident conditions, environmental conditions in the containment and auxiliary buildings, fission-product release, transport l anddeposition, decontamination, dose reduction, and waste handling and debris in the containment building, in particular the containment sump.

After access to the primary system is obtained, the primary system pressure )

boundary will be characterized including the steam generators, pumps, and other mechanical and structural components. Techniques will be developed for a non-destructive assay of fuel distribution in the primary system, for j assessing criticality control during examination and cleanup operations, and for fuel removal, packaging, shipment and disposal. There will also be detailed pre-access reactor and core damage assessments, followed by careful 3

in situ and away-from-site fuel and reactor internals examinations.

NRC has cooperated with DOE and industry on research and data-gathering

,rograms p within the THI-2 containment. These programs concentrate on retrieving data which characterize the progression of the accident and the resulting radiological source term. The programs address the examination of the reactor internals (especially the reactor core and fuel), the primary system piping and vessel, the dose reduction effort during decontamination, and various mechanical, electrical, and instrumentation equipment.

Recent core examinations indicate a large flow of molten material (more than 15 tons) into the lower plenum after the accident had been in progress for about 225 minutes. Additional research to examine the effects of this. molten material on the vessel bottom and vessel components is being pursued. .

O HAPHR-RC 3.3-22 AMENDMENT 3 AUGUST, 1989 9131e:1d

s r The information which may be obtained through this TMI Action Plan . item will k]/ 'be used in the pursuance of other generic safety issues such as:

USI A-45 Shutdown Decay Heat Removal Requirements w USI A-48 Hydrogen Control Measures and Effects of Hydrogen Burns on Safety Equipment II.B.5 Research on Phenomena Associated with Core Degradation on Fuel Melting q II.B.7 Analysis of Hydrogen Control )

Q II.B.8 Rulemaking Proceedings in Degraded Core Accidents  !

II.E.3.4 Alternate Decay Heat Removal Concepts SP/90 Resnonse This issue is associated with obtaining technical information on the condition of the TMI facility, and is not directly applicable to Nestinghoure in relation to the SP/90 design. Westinghouse will continue to follow the THI cleanup efforts and utilize lessons learned in the continuing SP/90 development program as applicable.

Issue II.J.4.1: Revise Deficiency Reporting Requirements Discussion The objective of task II.J.4 is to clarify deficiency report requirements to obtain uniform reporting and earlier identification and correction of problems. q Item II.J.4.1 called for the NRC to revise, as necessary, the event-reporting i O requirements of CFR Part 21 to assure that all reportable items are reported promptly and that the information submitted is complete. Improvements will be j

implemented by rule changes as appropriate and coordinated with those made I

under THI Action Plan Item I.E.6. The reports received as a result of these l f rule changes will provide increased information on component failures that 4

affect safety so that prompt and effective corrective action can be taken.

O 3.3-23 AMENDMENT 3 HAPHR-RC AUGUST, 1989 9131e:Id

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w The information will also be used as input to an augmented role of the NRC's

. vendor and construction inspection program. I NRC draft has proposed changes to 10 CFR Part 50.55(e) and 10 CFR Part 21. A Federal Register Notice of proposed changes to 10 CFR Part 50.55(e) and 10 CFR 3 Part 21 and publishing of the final rule is anticipated in 1988.

- SP/90 Resnonse This issue is associated with event-reporting requirements and is not applicable to Westinghouse in relation to SP/90 design.

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O HAPWR-RC 3.3-24 AMENDHENT 3 AUGUST, 1989 9131e:1d

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s 3.3.3 HUMAN FACTORS PROGRAM PLAN ISSUES, NUREG-0985 v

The many investigations which followed the accident at THI 2 identified the need to bring human factors considerations into the requirements and )

regulations for the design, construction, and operation of nuclear power l plants. NUREG-0660 contains many items which address human facters concerns.

The "U.S. Nuclear Regulatory Commission Human Factors Program Plan."

NUREG-0985, was subsequently developed to provide a more current and I

- comprehensive consideration of the outstanding human factors issues related to the design, operation, and maintenance of nuclear power plants.

The issues presented in this section include those outlined in the Human Factors Program Plan (HFPP) and documented in NUREG-0985. This plan describes the human factor-related work required to complete the NUREG-0660 human factors tasks as well as the additional human factors-rein ed efforts.

. identified during implementation of NUREG-0660 tasks, that require NRC attention.

The following discussions pertain to high and medium priority Human Factors 3 1 Program Plan issues in relation to the SP/90 design.

Issue: HF1.1: Shift Staffing Discussion Task HF1 was developed to assure that the number and capabilities of the staff at nuclear power plants are adequate to provide safe operation.

1ssue HF1.1 was to determine the minimum appropriate shift crew staffing composition. A determination was to be made from developed personnel I projection and alloe.ation models and from evaluations of job and task analysesand PRA data. Staffing practice of foreign and domestic utilities l

were surveyed to evaluate current practices, regulations, and staffing levels l

considering such variables as plant size, control room arrangement and O MAPHR-RC 3.3-25 AMENDMENT 3 9131e:1d AUGUST, 1989

r configuration, and plant layout. The issue consists of two parts: (1) the Staffing Rule and (2) conforming amendments to Regulatory Guide 1.114 and SRP Section 13.1.2.

The Staffing Rule which is officially known as " Licensed Operator Staffing at Nuclear Power Units" was published in the Federal Register on July 11, 1983 (48 FR 31611) with an effective date of January 1,1984; this rule is now I included in 10 CFR 50.54. The proposed conforming requirements to Regulatory Guide 1.114 and SRP Section 13.1.2 contain no requirements beyond those included in the Staffing Rule. Implementation of these requirements will be verified by resident inspectors. Nc, further verification will be necessary upon issuance of the Regulatory Guide and SRP changes.

SP/90 Resoonse The Control Room design process for the SP-90 control room requires as input from the utility a definition of the crew staffing. Using this input along with the techniques of plant processes functional decomposition and the 3 Rasmuessen model of real-time process operator decision-making (as described in Mcdule 15 of the Westinghouse RESAR-SP/90), the elements of the control room workstation (the alarm system, process data displays, controls) are sferived and designed. Once the elements are understood, the elements are organized and coordinated into a workstation design, again, based on the staffing assumptions, the result of the functional decomposition and the decision-making model. Finally, the workstations are coordinated into a total control room layout / design recognizing the interaction of the operating staff.

As a result, the results of HF1.1 initiative can be clearly incorporated in the design process of the SP/90 control room.

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HAPHR-RC 3.3-26 AMENDHENT 3 AUGUST, 1989 9131e:1d

l Issue HF4.1: Inspection Procedure for. Upgraded Emergency Operating Procedures  !

O l Discussion Task HF.4 is to provide assurance that plant procedures are adequate and can be used effectively. The objective is to provide procedures which will guide the operators in maintaining the plant is a safe state under all operating conditions, including the ability to control upset conditions without first having to diagnose the specific initiating event.

Item HF4.1 addresses criteria to evaluate and inspect EOPs by the regions.

These criteria have been prepared by NRR and OIE and were published as an OIE Temporary Instruction. Similar criteria and inspection modules will be developed when the guidelines for the upgrading of other procedures are completed.

SP/90 ResDonse It is intended that procedures in the SP/90 control room will be computerized. The computerization, however, is more than simply making the computer be a "page turner." It is intended that the procedures will be:

a) Derived from the functional decomposition, the decision-making model, and the resulting operator task analysis. The procedures, then, would be addressing the various activities or tasks in the decision-making model, i.e., Monitoring Planning, taking control action.

b) As a result of their derivation, the procedures become an element in O the operator's working environment. Therefore, we intend that the various procedures will appear, along with other decision-support process data, in the displays of the SP/90 control room.

This process ensures that the procedures are fully, i.e., conceptually as well as physically, integrated into the control room presentation. As a result, O lpPHR-RC 3.3-27 AMENDMENT 3 AUGUST, 1989 9131e:1d

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the NRC inspection and evaluation' procedures for procedures (EPO and otherwise) need to be able to effectively evaluate such a treatment of the gI operating procedures, i

Issue HF4.4: Guidelines for Upgrading Other Procedures l Discussion On the basis of current efforts to evaluate the quality of and the problems associated with existing plant procedures, NRR is evaluating the need to develop technical guidance for the industry to use to upgrade normal operating procedures (ops) and abnormal operating procedures (AOPs) as the staff hasdone for EOPs. Future work in this aret includes performing a regulatory analysis to determine whether regulatory oction for other plant procedures is warranted, and, if so, to develop formal regulatory requirements.

SP/90 Response 3

A discussion of how Westinghouse plans to develop and incorporate procedures (OP, AOP, EDP, etc.) in the SP/90 design was covered in the answer to HF.4.1.

Issue HFS.1: Local Control Stations Discussion The objective of task HF5 is to ensure that the man-machine interface (HMI) is adequate for the safe operation and maintenance of nuclear power plants. This '

objective will be attained by developing: (1) human factors engineering guidelines for correcting W I problems; and (2) regulatory guidance for integrating human factors engineering into new designs and into advanced technological improvements incorporated into existing designs. This activity will also provide for the preparation of evaluation tools for: (1) the next generation of nuclear power plants; and (2) expected changes or upgrading to designed plants in the area of data and information management and improved O

3.3-28 AMENDMENT 3 HAPHR-RC AUGUST, 1989 9131e:Id

n annunciator systems. In addition, . these efforts will improve the staff's V capability to evaluate reac' tor incidents involving MMI errors, r

The regulatory efforts to date dealing with the MMI have been limited t'o the control room and the remote shutdown panel.. Further guidance is necessary regarding local control stations and auxiliary operator interfaces.

Additional guidance may also be required regarding improvements to existir,g annunciator systems.

Information will be developed to determine if guidance on local control station design and auxiliary operator interfaces with these stations is required. To accomplish this task, job / task analyses of control room crew activities will be conducted to identify and describe communication and control links between the control room and the auxiliary control stations.

Inaddition, the functions of the auxiliary personnel will be analyzed from the task analyses to estimate the potential impact of auxiliary personnel job errors on plant safety.

SP/90 Resnonse 3 Westinghouse intends to fully apply the MMI design process discussed in RESAR-SP/90 PDA Module 15, " Control Room / Human Factors Engineering," to the

' local control stations.

Issue HF5.2: Review Criteria for Human Factors Aspects of Advanced Controls and Instrumentation Discussion Improved annunciator systems utilizing advanced technologies are expected to become available. Guidelines for the utilization and evaluation of these longer-term annunciator improvements will be developed. These guidelines will be based upon evaluations of results from advanced concept activities being performed by governmental and commercially sponsored research activities.

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The existing human engineering guidelines for nuclear power plant control rooms primarily address the control, display, and information concepts and  ;

I technologies which are now being used in process control systems. While these guidelines are adequate for the current generation of nuclear power plants, 1 they may not be sufficient for advanced and developing technologies which may (

be introduced into existing and future designs. This concern is addressed bythe combination of Item HF5.2 dealing with annunciators, Item HF4.5 dealing with automation and artificial intelligence, Item HF5.3 dealing with operational aids, and Item HF5.4 dealing with computers and computer displays.

Solutions to this combined issue will be changes to the SRP, guidance for the industry such as a Regulatory Guide, and the necessary staff expertise to evaluate proposed designs for the WI based on advanced technology.

SP/90 Resoonig The SP/90 control room utilizes many advanced HMI techniques that have become available and practical with the recent progress in digital computer 3 technology. Among these are an advanced alarm system, an advanced display system which assembles displays from a library of windows, soft or multi-function controls, etc. In the course of applying these new

. technologies, Westinghouse starts with first principles to develop a good, sound theoretical design basis and will do some operator testing to verify the application. However, extensive human factors testing is extremely difficult and costly. Westinghouse would welcome any additional insight into substantiating HMI design requirements through the results of NRC sponsored testing.

Issue HF8: Maintenance and Surveillance Program Discussion The purpose of the Maintenance and Surveillance Program (HSP) effort is to provide direction for the NRC's efforts to assure effective nuclear power O

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plant maintenance. The program will. be based on the current NRC regulatory O  : approach to maintenance. and an evaluation of. the effectiveness of current

- industry efforts in the maintenance area.

r The NRC's . current regulatory approach to nuclear power plant maintenance is concentrated on: (1) QA during design - construction, and operation for structures, systems and components important to safety (10_ CFR 50, Appendix 8), and (2) surveillance requirements to assure that the necessaryavailability and quality of such systems and components is maintained (10 CFR 50.36).

O. Despite extensive surveillance testing requirements, the NRC's rules. and regulations provide no clear programmatic treatment of maintenance. NRC additionally require.s reporting of significant events (10 CFR 50.72),

including personnel errors and procedural inadequacies which could prevent fulfillment of safety functions and exceeding of TS limits.

The NRC does not stipulate maintenance requirements for systems which are not safety-related. Many challenges to safety systems originate from systems / components which are classified as Act safety-related. The relationship between non-safety grade control systems and safety systems is 3 being addressed in USI A-47.

l The MSP is intended to integrate the NRC's efforts to assure effective nuclear power plant maintenance and to do so in a manner that is consistent with and responsive to the Commission's 1984 Policy and Planning Guidance. The program addresses the problems and issues which exist and proposes development of

- alternative NRC approaches to regulating nuclear utility maintenance activities consistent with the Policy and Planning Guidance. The scope of the program includes all aspects of maintenance required to carry out a systematic I O skintenance and surveillance program. It includes conventional maintenance and repair plus such things as surveillance and test activities, equipment isolation, post-maintenance testing, independent verification, maintenance management, administrative control, personnel selection and training, O procedures, and technical documentation.

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'O Operating experience has indicated that at least one third of the abnormal nuclear power plant occurrences may be directly attributed to maintenance I errors. In addition, reviews of operating experiences show a high frequency of degraded system performance caused by the lack of maintenance and improperly perfbrmed maintenance including human error during repair and )

surveillance testing. A recent NUMARC study of 600 events showed that 51% of all maintenance problems can be traced to human factors problems and another l 431 are attributable to either design or manufacturing.

The proposed solution to this issue is to implement a systematic maintenance program as addressed in the NRC's preliminary MSP with the following five objectives:

1 (1) To assure that needed maintenance is being accomplished, especially in counteracting system and equipment aging effects, by taking appropri-ate preventive and corrective action to minimize equipment failures. 1 (2) To reduce failures from improper maintenance to an acceptable level and to assure safety through effective maintenance management, personnel selection and training, procedures, administrative control, and design for maintainability.

(3) To assure proper integration of maintenance operations an'd other organizational interfaces for maintenance activities which can affect plant safety.

(4) To improve the effectiveness of nuclear power plant maintenance programs in reducing the number of challenges to safety systems (e.g.,

reactor scrams).

(5) To optimize surveillance requirements to assure equipment availability when required without excessive equipment out-of-service intervals for l testing and to eliminate the unnecessary exposure for transient trips i due to excessive test frequencies of logic and initiation systems.

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SP/90 Resnonse 1

A major objective for the design of the SP/90 is to reduce the probability of i operating abnormalities caused by problems in I&C maintenance and surveillance testing activities. Four major design features were employed to achieve this objective.

1) Continuous On-line Testing The incorporation of continuous on-line testing features and specific automatic actions to be taken upon detection of a fault coordinated with system architectural features to prevent detected faults from causing abnormal plant operating conditions. The use of distributed digital technology in the implementation of protection and control system I&C made this feature practical.

In essence the effectiveness of this design feature is superior to a manually performed preventative maintenance program.

3 Recent studies performed by Westinghouse have developed a quantification method to support the contention that these continuous testing features can reduce the frequency of periodic surveillance testing - for a given overall reliability requirement.

2) Built-in Automatic Test Equipment The incorporation of built-in automatic testing equipment for the performance of the periodic surveillance tests. As stated in the requirements section the NRC does impose extensive surveillance testing requirements on the safety related systems and components. The protection system design includes built-in automatic testing equipment for the performance of periodic surveillance tests. This feature assures the  !

t performance of a complete and comprehensive test within a minimum time l

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period to reduce .the time period that part of the protection system is out of service and also reduce the probability of human error during the gjI performance of the tests.

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3) Automatic System Reconfiguration Incorporation of automatic system reconfiguration features for testing and l maintenance activities. Addresses one of the principal human performance I

problem areas.

A major source of human error in testing and maintenance occurs during the reconfiguration of the system to permit taking equipment out of service for these activities while at power.

i The protection system design includes two features that provide this automatic reconfiguration. One is the global and local bypass feature that automatically reconfigure the protection system coincidence logic from 2-out-of-4 to 2-out-of-3 or, in special uses, from 2-out-of-3 to 1-out-of-2 when a protection function or protection channel is taken outof service. This automatic reconfiguration feature reduces the chances of challenges to the safety system caused by spurious or random equipment malfunctions while in a 1-out-of-3 coincidence logic configuration (this is common in present designs since the channel or function out of service for testing or maintenance is put in a tripped condition rather than bypassed).

1 The second feature for automatic reconfiguration is the signal selector. I

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The signal selector provides functional isolation between the protection system and the control system for the case where control system process variable signals are taken from redundant measurements made in the protection system. The signal selector compares the redundant protection g system signals provided at the output of the isolation amplifiers and W through a specific selection algorithm rejects signals that are not a l O.

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member of the redundant set because of differences caused by testing or maintenance activities. The signal selector also rejects signals that fail the selector test because of malfunctions in the protection system.

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p The signal' selector eliminates a major cause of abnormal plant operating conditions for testing activities in present plant designs. In these designs the control system signal is taken from a single protection system selected by a switch on the main control board. If the I&C technician

'q fails to inform the control room operator before he takes a protection

's/ channel or function out of service for maintenance or test, an erroneous signal can be sent to the control system with a resulting transient upset.

4) Modular Equipment Design Features The modular equipment design features employed in both the protection and  ;

control system I&C packaging design significantly improve the man-machine interface for maintenance diagnostic and repair activities.

( The hardware aspects of the modular design permit access to equipment on a preplanned basis for removal of cards and the use of a card-keying feature that prevents insertion of wrong cards. In addition, software keys are employed to assure consistent configuration sets for new cards (for example all memory components on a card are consistent with a particular revision).

In summary Westinghouse has incorporated design features that address the possible solutions to maintenance and surveillance problems that are listed in item HF8 of NUREG-0933 page 4 HF8. The relationship between these possible O solutions and the design features is given in the table.

O HAPHR-RC 3.3-35 AMENDMENT 3 AUGUST, WS9 9131e:1d

3 Table 1 O Summary of M Design Features for Improved MSP' Possible Solutions HFB-2 W Desian Features

1) To assure that needed maintenance is being o Continuous on-line accomplished, especially in counteracting system o Built-in automatic test and equipment aging effects, by taking equipment appropriate preventive and corrective action to minimize equipment failures.

2). To reduce failures from improper maintenance to o Modular equipment design

.an acceptable level and to assure safety through o Automatic system effective maintenance management, personnel reconfiguration selection and training, procedures, administrative o Built-in automatic test control, and design for maintainability equipment l

3) To assure proper integration of maintenance o Automatic system operations and other organizational interfaces reconfiguration for maintenance activities which can affect plant safety.

4) To improve the effectiveness of nuclear power o Automatic system plant maintenance programs in reducing the reconfiguration ,

[ number of challenges to safety systems (e.g.

l reactor scrams)

5) To optimize surveillance requirements to assure o Built-in automatic test equipment availability when required without equipment

, excessive equipment out-of-service intervals o Continuous on-line for testing and to eliminate the unnecessary testing exposure for transient trips due to excessive test frequencies of logic and initiation systems

• Maintenance and Surveillance Program O

O MAPHR-RC 3.3-36 AMENDMENT 3 9131e:1d AUGUST, 1989

q 4.0 UNRESOLVED SAFETY ISSUES D

The NRC continuously evaluates the safety requirements used in their reviews against new information as it becomes available. Information related to the

( safety of nuclear power plants can come from a variety of sources including

( experience'from operating reactors; research results; NRC staff and Advisory Committee on Reactor Safeguards (ACRS) safety reviews; and vendor, architect engineer, and utility design reviews.

O i Each timo a new concern or safety issue is identified, the need for immediate

(,/

actica to ensure safe operation is assessed by the NRC. In some cases, immediate NRC action is taken to ensure the safety of operating plants. In 3

other cases, immediate licensing actions or changes in licensing criteria are not considered necessary, but further study by the NRC may be deemed appropriate before judgments are made as to whether existing requirements should be modified to address the issue for new plants or if backfitting is appropriate for the long-term operation of plants already under construction or in operation.

These issues are called " generic safety issues" or " unresolved safety issues" and they do have a potential impact on all plant designs including the SP/90 design. NRC " generic safety issues" are discussed in Section 5.0. This section is devoted to the discussion of " unresolved safety issues".

The NRC defines an Unresolved Safety Issue as "a matter affecting a number of nuclear power plants that poses important questions concerning the adequacy of existing safety requirements for which a final resolution has not yet been developed and that involves conditions not likely to be acceptable over the lifetime of the plant it affects."

Each year the NRC reviews their task action plans and generic issues to define

/ a current set of Unresolved Safety Issues which is reported to Congress.

These annual reports usually identify those Unresolved Safety Issues that were technically resolved from the previous annual report.

O 4.0-1 AMENDMENT 3 WAPWR-RC B944e:1d AUGUST 1989

_ _ - - _ _ _ _ _ - - _ - - _ - - - _ - - - - - - )

s l

The purpose of this section is to assess each Unresolved Safety Issue relative i

3 lto its impact, or potential impact, on the SP/90 design. f The following current list of Unresolved Safety Issues has been obtained from l

3 lNUREG-0933, "A Prioritization of Generic Safety Issues."

c O

3l Water Hammer (A-1)*

, o Asymmetric Blowdown Loads on the Reactor Primary Coolant Systems (A-2)*

1 3 o WestinghouseSteamGeneratorTubeIntegrity(A-3)*

O l

3l o Combustion Engineering Steam Generator Tube Integrity (A-4)**

3l o Babcock and Wilcox Steam Generator Tube Integrity (A-5)**

3l o Mark I Short Term Program (A-6)**

3 o MarkILongTermProgram(A-7)** g 3 o Mark II Containment Pool Dynamic Loads (A-8)**

o AnticipatedTransientsWithoutScram(A-9)*

3l o BWR Feedwater Nozzle Cracking (A-10)**

o Reactor Vessel Materials Toughness (A-11)*

o Fracture Toughness of Steam Generator and Reactor Coolant Pump 3l Supports (A-12)*

• NRC technical resolution for each of these Unresolved Safety Issues has been issued.

3 ** Issues which are not applicable to RESAR-SP/90.

WAPWR-RC 4.0-2 AMENDMENT 3 O

E944e:1d AUGUST 1989

o Systems Interactions in Nuclear Power Plants (A-17)

\,

o Qualification of Class 1E Safety-Related Equipment (A-24)*

l

.,e o- Reactor Vessel Pressure Transient Protection (A-26)*

o Residual Heat Removal Requirements (A-31)*

o Control of Heavy Loads Near Spent Fuel (A-36)*

o Determination of Safety Relief Valve Pool Dynamic Loads and Temperature Limits for BWR Containments (A-39)** l3 o SeismicDesignCriteriaShortTermProgram(A-40) o Pipe Cracks in Boiling Water Reactors (A-42)** l3 o Containment Emergency Sump Performance (A-43)* l3 o StationBlackout(A-44)*

l3

,o Shutdown Decay Heat Removal Requirements (A-45) o Seismic Qualification of Equipment in Operating Plants (A-46)* l3 o Safety Implications of Control Systems (A-47) o Hydrogen Control Measures and Effects of Hydrogen Burns on Safety Equipment (A-48) o Pressurized Thermal Shock (A-49)* l3 0

O WAPWR-RC 4.0-3 AMENDMENT 3 8944e:1d AUGUST 1989 l

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _]

i w

1. Issue A-1: Water Hammer Discussion The issue was raised after the occurrence of various incidents of water hammer that involved steam generator feedrings and piping, emergency core cooling systems, RHR systems, containment spray, service water, feedwater, and steam lines. The incidents have been attributed to such causes as rapM condensation of steam pockets, steam-driven slugs of water, pump startup with partially empty lines, and rapid valve motion. Most of the damage has been relatively minor and involved pipe hangers and restraints. However, there have been several incidents which have resulted in piping and valve damage. This item was originally identified in NUREG-0371 and was later determined to be a USI.

No water hammer incident has resulted in the release of radioactivity outside of plants. However, because of the continuing incidence of water 3- hamer events, the number of phenomena, and the potential safety significance of the systems involved, the staff believed that systematic review procedures should be developed to ensure that water hammer is given appropriate consideration in CP and OL reviews and in the review of operating reactors.  ;

This USI was RESOLVED on March 15, 1984 with the publication of NUREG-0927, Rev. I and the following SRP Sections: 3.9.3, Rev. 1; 3.9.4, Rev. 2; 5.4.6, Rev. 3; 5.4.7, Rev. 3; 6.3, Rev. 2; 9.2.1, Rev. 3; 9.2.2, Rev. 2; 10.3, Rev. 3; and 10.4.7, Rev. 3. The revised SRP Sections will 1

be used only for the review of " custom plant" CP applications and for standard plant applications docketed after the issuance of these revised SRP Sections (which are intended for referencing in CP applications).

h Thus, this USI affects all future plants only.

l WAPWR-RC 4.0-4 AMENDMENT 3 5944e:1d AUGUST 1989 L._____ _ _ _ _ _ _ _ _ _ _ _ _ _

SP/90 Response The SP/90 design includes an emergency feedwater system (EFWS) and a startup feedwater system (SFWS). The EFWS is a safety system utilizing four pumps; two electric motor-driven and two steam turbine-driven. The EFWS functions similarly to a conventional auxiliary feedwater system except that following reactor trip and during' normal plant startup/

shutdown and hot standby the SFWS is utilized. The EFHS is designed for such events as main steam line breaks, main feedwater line breaks, steam generator tube ruptures, loss- of-coolant accidents, loss of all AC power, and any other event in which the main and startup feedwater systems are not available. The SFWS is a control grade system utilizing one motor-driven pump and provides heated feedwater during normal plant startup/ shutdown and hot standby, thus reducing the probability of water hammer in the feedwater or steam generator feedrings.

1 3

Special valve designs are being evaluated for the SP/90 design to reduce the probability of water hammer events in feedwater systems due to rapid O motion of check valves. Also, feedwater piping configurations will be designed to minimize the potential for water hammer.

. The current NRC criteria related to water hammer do not require any i additional measures beyond those already implemented in current l

Westinghouse ~ designs. However, additional design features as described I f above have been incorporated in the SP/90 design.

2. Issue A-2: Asymmetric Blowdown Loads on the Reactor Primary Coolant Systems Discussion This issue concerns asymmetric loadings which could act on the reactor's primary system as the result of a postulated double-ended rupture of the piping in the primary coolant system. The magnitude of these loads is O

WAPWR-RC 4.0-5 AMENDMENT 3 E944e:1d /.UGUST 1989 ,

s 1

1 potentially large enough to damage the supports of the reactor vessel, the reactor internals, and other primary components of the system. Therefore, g

the NRC initiated a generic study to gain a better understanding of these loads and to develop criteria for an evaluation of the response of the i primary systems in pressurized water reactors to these loads.

3 l This USI was resolved with the publication of NUREG-0609, "Asymetric Blowdown Loads on PWR Primary Systems." This report provides an accept-  !

able basis for performing plant analyses for asymmetric loss-of-coolant accident loads. Guidelines and criteria are given for the evaluation of the loading transients, structural components, and fuel assemblies.

Westinghouse has addressed the subject of asymmetric loss-of-coolant ,

accident loads in the design and analysis of plants currently under NRC review. These analyses have demonstrated compliance with NUREG-0609.

As discussed in Section 5.1 (item 18), Westinghouse has performed extensive material testing and fracture mechanics evaluations to demonstrate that pipe breaks need not be postulated in the reactor coolant system. Both the NRC and ACRS have endorsed this concept and the h

methodology developed to support it.

SP/90 Response As part of the SP/90 PDA application, Westinghouse has applied revised pipe break criteria to all SP/90 high energy fluid systems in order to reduce / eliminate the need to postulated pipe breaks. These criteria do not require pipe breaks to be postulated in high energy fluid system piping unless sotae mechanism (e.g., corrosion, water hammer) exists which g

could result in a pipe break. The bases for this position are the material properties of the piping and the methodology previously developed for reactor coolant system piping.

O WAPWR-RC 4.0-6 AMENDMENT 3 E944e:1d AUGUST 1989

)

. I With the elimination of pipe breaks Westinghouse intends.to eliminate the structural effects considered in the piping structural evaluation. Some of these structural' effects include- blowdown loads and jets from previ- l ously postulated' pipe breaks; pipe whip restraints on piping; and .)

pressurization effects from previously postulated pipe breaks. As part of O- the FDA application analyses will be performed to demonstrate that the 3

structural integrity of the reactor vessel supports, reactor internals, and other primary system components are maintained within acceptable

3. . Issue A-3: Westinghouse Steam Generator Tube Integrity-Discussion Pressurized water reactor steam generator tube integrity can be degraded by various mechanisms, including corrosion induced wastage, cracking,  ;

reduction in tube diameter (denting, which leads to primary side stress corrosion cracking) vibration induced fatigue cracks and wear or fretting due to loose parts in the secondary system. The primary concern is the capability of degraded tubes to maintain their integrity during normal operation and under accident conditions (LOCA or a main steam line break) w ith adequate safety margins.

Steam generator tube integrity concerns for the three steam generator 3 suppliers, Westinghouse, Combustion Engineering, and Babcock and Wilcox are being addressed by an integrated NRC program for USIs, A3, A4 and AS.

This program addresses the areas of steam generator integrity, plant systems response, human factors, radiological consequences and the response of various organizations to a steam generator tube rupture. A generic risk assessment is to be provided on the risk from steam generator rupture events. The report is to identify actions that would be effective in significantly reducing the incidence of steam generator tube degradation. The frequency of tube ruptures and the corresponding potential for significant non-core melt radiological releases, and O WAPWR-RC 4.0-7 AMENDMENT 3 5944e:1d AUGUST 1989

s l

occupational radiological exposures and which would be effective in mitigating the consequences of SGTR events.

Major findings emerging from this program have included the following:

SGTRs do not contribute a significant fraction of the early and latent cancer fatality risks associated with other reactor events at a given site. The increment of risk associated with SGTR events in a small -

fraction of the accidental and latent cancer fatality risks to which the general public is routinely exposed.

)

The staff has recommended actions which have been found to be )

effective measures for significantly reducing (a) the incidence of tube degradation, (b) the frequency of tube ruptures and the corresponding potential for significant non-core melt releases, and 3

(c) occupational exposures and which would be effective in mitigating the consequences of SGTRs.

The staff has issued GL 85-02 to inform licensees and applicants of the staff recommended actions.

g It appears that both industry initiative and the staff recommendations have been responsible for significant improvements to plant programs over the past few years.

The staff will continue to monitor steam generator operating ,

experiences as an indicator of the effectiveness of utility programs.

Final publication of NUREG-0844 constitutes resolution of USI's A-3, i

A-4, and A-5.

O l 1 WAPWR-RC 4.0-8 AMENDMENT 3 O

5944e:1d AUGUST 1989 i

s SP/90 Response The steam generator design for the SP/90 includes features to minimize tube integrity problems, such as:

O _.

a,c 9

l 3

The Staff Recommended Actions of Generic Letter 85-02 are addressed in the steam generator design of the SP/90 as described in the following: The numbers correspond to the numbers of the items in the Generic Letter.

la. Prevention and detection of loose parts (inspections)

This recomended action included the recommendations to inspect the tubesheet in the region of the tube lane and the annulus between the tube bundle and the shell for the presence of loose parts or foreign O

WAPWR-RC 4.0-9 AMENDMENT 3 B944e:1d AUGUST 1989

s objects or damage to the tubes. Any objects found should be removed and any tubes with visible evidence of damage should be eddy current inspected.

Response: The steam generator design for the SP/90 includes access provisions which will permit the recommended inspections.

1.b Prevention and detection of loose parts (quality assurance)

This recommended action deals with procedures to control loose parts in steam generators during maintenance activities.

Response: This recommendation does not involve steam generator design.

2.a Inservice inspection program (full length tube inspection) 3 This recommendation is for full length addy current inspection of the steam generator tubes from tube end on the hot leg side to tube end of

~ '

the cold leg side.

Response: The steam generator design will permit full length tube inspections. The longer radius of the row 1 tubes will facilitate

~

such inspections.

2.b Inservice inspection program (inspection interval)

This recommendation is to limit the time permitted between eddy current ins'pections to not greater than 72 months.

O Response: This recommendation does not involve steam generator design.

WAPWR-RC 4.0-10 AMENDMENT 3 E944e:1d AUGUST 1989

l 3.a Secondary water chemistry program O)

This recommendation is to incorporate the EPRI secondary water  !

ci,emistry guidelines.

Response: The steam generator design is compatible with EPRI secondary water chemistry guidelines.

3.b Condenser inservice inspection program This recommendation is to implement an inservice inspection program.

Response: The condenser inspection recommendations and the design provisions for such inspections are the responsibility of the utility.

4. Primary to secondary leakage limit 3

This recommendation is to apply the leakage limits in the Standard Technical Specifications to plans with less restrictive limits.

Response: The primary to secondary leakage limit in the Technical

. Specifications for the SP/90 will be consistent with the Standard Technical Specifications.

5. Coolant iodine activity limit This recommendation is to apply the primary coolant iodine limits in the Standard Technical Specifications to plants with less restrictive limits.

Response: The primary coolant iodine activity limit in the Technical Specifications for the SP/90 will be consistent with the Standard Technical Specifications.

I O l WAPWR-RC 4.0-11 AMENDMENT 3 B944e:1d AUGUST 1989 L--------_--_----------------------------- -A

I 1

6. Safety injection signal reset.

This recommendation is related to the control logic associated with the safety injection pump suction flow.

Response: This recommendation is not applicable to the design of the )

SP/90 safety injection system.

3 I The Staff Actions resulting from this issue relating to steam generator design and operation yet to be completed are in the area of guidelines for eddy current techniques and additional tube inspections. The SP/90 steam generator design will not prevent the use of improved eddy current techniques or additional inspections. Steam generator design features may reduce the need for proposed additional inspections such as denting '

inspection.

4. Issue A-4: Combustion Engineering Steam Generator Tube Integrity j

]

Discussion This issue was combined with issue A-3 by the NRC, but is not applicable to Westinghouse steam generator designs (see Item 3 above).

5. Issue A-5: Babcock and Wilcox Steam Generator Tube Integrity Discussion j l

This issue was combined with issue A-3 by the NRC, but is not applicable j to Westinghouse steam generator designs (see Item 3 above). l l

6. Issue A-6: Mark I Short Term Program This issue is not applicable to Westinghouse pressurized water reactor designs.

WAPWR-RC 4.0-12 AMENDMENT 3 O

B944e:Id AUGUST 1989

W p 7._ Issue A-7: Nark I Long-Term Program ) '

A) <

This issue is not applicable' to Westinghouse pressurized water reactor.  !

' designs.

8. Issue A-8: Mark I'I Containment Pool Dynamic Loads j I

~This issue is not' applicable to Westinghouse pressurized water reactor designs.

9. Issue A-9: Anticipated Transients Without Scram j Discussion Nuclear plants. have safety and control systems-to limit the consequences of temporary abnormal operating conditions or " anticipated transients".

Some deviations from normal operating conditions may be minor; others, occurring less frequently, may impose significant demands on plant equipment. In some anticipated transients, rapidly shutting down the

j. nuclear reaction (initiating a " scram"), and thus rapidly -reducing the generation of heat in the reactor core, is an important safety measure.

A potentially severe " anticipated transient" where the reactor shutdown system does not " scram" as desired, is an " anticipated transient without scram", or ATWS. The technical raport on ATWS for water-cooled reactors (WASH-1270) discussed tha probability of an ATWS event as well as an appropriate safety objective for the event. After several years of discussions with vendors and evaluations of vendor models and analyses, the staff published in 1975 a status report on each vendor analysis. This 3 i

report included detailed guidelines on analysis models and ATWS safety objectives. This item was originally identified in NUREG-0371 and was later determined to be a USI.

4 1 The staff's technical findings on the issue were published in Volume 4 of WUREG-0460. The USI was RESOLVED on June 26, 1984 with the publication of J

l-WAPWR-RC 4.0-13 AMENDMENT 3  !

5944a:1d AUGUST 1989

s a final rule. Federal Register, Vol. 49, No. 124, pp. 26036-26045, "10 CFR Part 50, Reduction of Risk from Anticipated Transients Without Scram (ATWS) Events for Light-Water-Cooled Nuclear Power Plants."

The requirements as shown below apply to all commercial light-water cooled g nuclear power plants. W

1. Each pressurized water reactor must have equipment from sensor output to final actuation device, that is diverse from the reactor trip system, to automatically initiate the auxiliary (or emergency) feedwater system and initiate a turbine trip under conditions indicative of an ATWS. This equipment must be designed to perform its function in a reliable manner to be independent (from sensor output to the final actuation device) from the existing reactor trip system.

2. Each pressurized water reactor manufactured by Combustion Engineering or by Babcock and Wilcox must have a diverse scram system from the sensor output to interruption of power to the control rods. This 3

scram system must be designed to perform its function in a reliable manner and be independent from the existing reactor trip system (from sensor output to interruption of power to the control rods).

3. Each boiling water reactor must have an alternate rod injection (ARI) system that is diverse (from the reactor trip system) from sensor output to the final actuation device. The ARI system must have redundant scram air header exhaust valves. The ARI must be designed to perform its function in a reliable manner and be independent (from the existing reactor trip system) from sensor output to the final actuation device.

4. Each boiling water reactor must have a standby liquid control system (SLCS) with a minimum flow capacity and boron content equivalent in control capacity to 86 gpm of a 13 percent sodium pentaborate solution. The SLCS and its injection location must be designed to perform its function in a reliable manner. The SLCS initiation must WAPWR-RC 4.0-14 AMENDMENT 3 O

B944e:1d AUGUST 1989

- - _ - _ - _ _ _ _ - - - - - -- a

3

(,

be automatic and must be:' designed to perform its function in a

. reliable manner for. plants granted a construction permit after '

July 28, 1984, and for plants granted a construction ' permit prior to July 28, - 1964, L that have already been designed and built to include this feature;

5. Each boiling water reactor must have equipment to trip the- reactor coolant recirculating pumps automatically under conditions indicative O of an ATWS. This equipment must be designed to perform' its function in a reliable manner.

SP/90 Response The SP/90 plant' includes the following design features related to ATWS.

o The . design of the integrated protection system (IPS) is highly reliable. The.IPS is based on two-out-of-four logic throughout and

-features continuous on-line testing. The system contains " fail-safe" features to the extent practical, i.e., it is designed to generate a 3 reactor trip signal when failures occur.

.- o The reactor trip switchgear consists of eight circuit breakers arranged in a two-out-of-four matrix and located in two separate cabinets. The trip is implemented by undervoltage trip attachments and shunt trip devices on the circuit breakers. To generate a trip, power is interrupted to the undervoltage trip attachment, and the shunt trip attachment is energized. Either device will trip the breaker. The eight breaker configuration permits testing of the reactor trip breakers without the use of auxiliary bypass breakers,

o. The reactor trip switchgear can be actuated manually from the main control board via reactor trip switches hard wired to the shunt trip and undervoltage coils on each circuit breaker. In addition, it is possible to trip from the main control board the motor generator sets that provide power for control rod operation.

O WAPWR-RC 4.0-15 AMENDMENT 3 E944e:1d AUGUST 1989

l o The moderator temperature coefficient (MTC) is significantly more j negative than in the case of current plants, typically by a factor of )

three to four.

o ATWS considerations will be factored into the design of the pressurizer safety and relief valves during the detailed design phase, i

o An ATWS mitigating system is included in the SP/90 design to generate l turbine trip and emergency feedwater start signals independent (including sensors) from the IPS.

o Detailed analyses of limiting ATWS transients will be performed at the FDA stage to demonstrate that ATWS acceptance criteria are met.

The acceptance criteria, as well as the assumptions to be used in the ATWS analysis will be agreed upon between the NRC and Westinghouse as part of the process to develop the Licensing Review Basis (LRB) for the SP/90 3 plant. If the ATWS analyses to be performedud' ring the FDA stage do not demonstrate compliance with acceptance criteria, Westinghouse will consider additional design features te mitigate ATWS transients, including incremental redundance in the reactor trip system.

RESAR-SP/90 PDA Module 16, "Probabilistic Safety Study," includes a probabilistic analysis of ATWS sequences. The core melt frequency  ;

contribution from this event has been calculated to be 5.5 E-08 per year. j Recently Westinghouse has completed a number of ATWS analyses for current '

plants. A review of this effort has indicated that the SP/90 analysis g .

include a non-conservative assumption in that common mode failure of the W l rods to enter the core because of mechanical problems was not assumed.

The probability of this common mode failure has been evaluated as 1.0 E-6 per demand. Applying this value to the SP/90 ATWS analysis would have the effect of increasing the core melt frequency for this event by a factor of 3.3.

WAPWR-RC 4.0-16 AMENDMENT 3 91 l E944e:1d AUGUST 1989 l

On the other hand, the SP/90 analysis contains several conservatism:

o The assumed number of transients (10 per year) is much higher than the SP/90 design goal of 1 per year, and is well above current operating n experience. Moreover, ,o credit was taken for the fact that some of V the transients are initiated as a result of reactor trip.

o No credit was taken for the startup feedwater system; this system is A automatically started by the integrated control system on loss of main U feedwater, which is typically the limiting ATWS event.

3 When all of the above factors are considered, it is expected that the core melt frequency due to ATWS will be less than 1.0 E-7 per year for the SP/90 plant. In addition, it should be noted that the probability of a severe release following an ATWS induced core melt is very low, such that this event has negligible impact on public risk.

Based on the above discussion, Westinghouse is of the opinion that the I SP/90 design adequately addresses ATWS issues and that no additional hardware design features are required at the PDA stage. 4

10. Issue A-10: BWR Feedwater Nozzle Cracking

)

This issue is not applicable to Westinghouse pressurized water reactor designs.

11. Issue A-11: Reactor Vessel Materials Toughness Discussion Stael commonly used in the construction of reactor pressure vessels (RPV)

D exhibits fracture toughness that varies greatly with temperature. Steel has relatively high toughness at high temperatures but low toughness at )

low temperatures. The temperature or temperature range where the transi- I tion from high-toughness (ductile) to low-toughness (brittle) behavior occurs is commonly referred to as the ductile-brittle transition tempera-i WAPWR-RC 4.0-17 AMENDMENT 3 E944e:1d AUGUST 1989 L

i 1

ture. Thus the temperature-dependent fracture toughness has three more-or-less distinct zones: a lower shelf with low toughness, an intermediate transition region, and an upper shelf with high toughness. l l

{

Charpy-impact (C) test.dataintheform of specimen-fracture energy, as I a function of temperature, reflect the ductile-brittle transition. The transition temperature can be identified in several ways, the simplest of which is to report the temperature at an arbitrary C yenergy level (for ]

example, 35 ft-lb). The upper shelf energy is the energy level of the q upper asymptote of the ECv = f(T) curve.

The embrittling effect of neutron radiation may so change the mechanical properties that the steel in a RPV would fail to meet the toughness requirements of 10CFR50. This could result from either too large a temperature increase in the reference-transition temperature (RTNDT), or j too large an energy decrease in the C yupper-energy level, or both. The I magnitude of the irradiation-induced changes depends, among other things, on the chemistry and metallurgical condition of the steel. The effect of copper content can be singled out because it plays a major role in the behavior. Copper was introduced by the practice (later abandoned) of g

coppercoating the consumable electrode weld wire to protect it from rusting and to increase its electrical conductivity. Experiments have shown that the radiation-induced changes in both the transition temperature and the C, y increase with copper content and the most sensitive steels include wald metals with relatively high-copper content.

Because some high copper welds exhibited relatively low initial upper ,

shelf energy levels, it was found to be more significant with respect to l violation of regulatory requirements than the corresponding transition j temperature increase. l Regulatory Guide 1.99 (Rev.1) shows conservative measures of the changes in transition temperature and upper shelf with fluence, copper and phos-phorus contents are shown parametrically. The guide is updated as sig-nificant additional data from surveillance or test reactor programs become available. Conservatism was included by constructing the curves as upper bounds of property changes rather than averages.

WAPWR-RC 4.0-18 AMENDMENT 3 5944e:1d AUGUST 1989

_ _ ~ - - - - - _ _ _ . _ _ _ - - . _ - - - -

l.

4 Guidance for licensees to provide justification for continued operation is given in NUREG 744. Rev. 1 (Generic Letter 82-26). In accordance with the requirements of 10CFR50,- Appendix G, all licensees should take' the following course of_ action. The upper. shelf energy at the plant-specific and of life (EOL) should be established in accordance with 10CFR50 and the ASME code. If the EOL upper shelf energy y_ 50 ft-1b, the reactor pres-l sure vessel is acceptable (other factors, detailed in 10CFR50 and in the Code, remain in_ force).- If the EOL upper shelf energy 5 50 ft-lb, either a safety analysis should be performed to' demonstrate that' the vessel can operate with adequate margin or a thermal anneal could be performed to restore the material toughness. To be acceptable, the analysis .must show adequate margin under normal, upset, emergency, faulted, and test conditions. The analysis may follow either the method recommended by the NRC or a method of equal or better reliability.

Appendix G to 10CFR50 essentially adopts the method of ASME Code Appendix G, with additional restrictions related to the presence of fuel or criticality. However, 10CFR50, Appendix G, extends the applicability of the design rules to operations, and fluence effects that must be considered.- Because the resulting pressure and temperature limitations must be included in the plant Technical Specification, which controls plant operation, the 10CFR50 Appendix G rules apply to all operating plants.

The need to include rules for emergency and faulted condition control in the ASME Code,Section III, Appendix G, is not clear. The Section III rules are of value only to the extent that they influence the construction and it is not apparent that such rules would have that effect. Although material selection might be influenced, indications are that the current acceptance criteria are satisfactory in that they provide adequate lifetime fracture resistance.

O' Results from reactor vessel surveillance programs indicates that as many as 20 operating PWRs will have beltline materials with marginal toughness, relative to the requirements of Appendices G and H of 10CFR50, after comparatively short (approximately 10 effective power years) periods of WAPWR-RC 4.0-19 AMENDMENT 3 5944e:1d AUGUST 1989

s operation. The specific requirement that may be voided is that of paragraph V.B. Appendix G, 10CFR50. For vessels failing to satisfy that requirement, paragraph V.C.3, Appendix G, 10CFR50, must be satisfied (along with the rest of V.C); that is, the owner must perform an analysis demonstrating the existence of adequate operational safety margins against .

fracture. For plants currently under licensing review, reactor vessels generally have acceptable fracture toughness. However, a few plants under licensing review have reactor vessels that have been identified as having the potential for marginal fracture toughness within their design life; these vessels will have to be reevaluated in the light of the new criteria for long-term acceptability.

Techniques for periodic surveillance of reactor vessel welds are discussed in Regulatory Guide 1.150, Rev. 1, " Ultrasonic Testing of Reactor Vessel Welds During Preservice and Inservice Examinations."

SP/90 Response For the SP/90 design the core baffle / reflector region has been optimized to provide increased shielding of the reactor vessel to reduce neutron g

irradiation in the reactor vessel beltline region to less than that of the best currently operating plants. The reduced neutron irradiation will 1ead to increased fracture toughness of the material in the reactor vessel beltline. Residual copper content of the SP/90 reactor vessel beltline material will be at or below that specified for the most recent l Westinghouse reactor vessels. Residual copper content is a key contributor to the loss of reactor vessel material toughness in the presence of neutron irradiation.

The design features summarized above will ensure that the SP/90 reactor pressure vessel will maintain high fracture toughness properties through-out plant life, and thus, will not require additional analysis under 10CFR g Part 50, Appendix G. Therefore, fN1 resolution of this issue has no W '

additional impact on the SP/90 design.

i WAPWR-RC 4.0-20 AMENDMENT 3 e

5944e:1d AUGUST 1989

12. Issue A-12: Fracture -Toughness -of Steam Generator and Reactor Coolant O' Pump Supports Discussion-This issue deals with the potential for lamellar tearing and low fracture toughness' of the steam' generator and reactor coolant pump support materials..

NUREG-0577, " Potential for Low Fracture Toughness and Lamellar Tearing on PWR Steam Generator and Reactor Coolant Pump Supports," was formally 3

issued by the NRC in 1983. This NUREG- categorizes operating plants relative to the adequacy of the plant's steam generator and reactor coolant pump supports with respect to fracture toughness. In general, the conclusions of NUREG-0577 are that supports for the reactor coolant pumps and steam generators in recently licensed pressurizer water reactors have adequate fracture toughness. Westinghouse believes that designing and fabricating these supports in accordance with Subsection NF of Section III of the ASME Code provides adequate ausurance of acceptable fracture toughness of materials, and ensures compliance with NUREG-0577. l3 SP/90 Response A new SRP that endorses Subsection NF of Section III of the ASME Code is expected to be issued in the near future. With respect to lamellarl 3 tearing, the current Westinghouse design for supports does not contain the thick, heavy weldments of the type possibly susceptible to lamellar tearing.

The SP/90 steam generator and reactor coolant pump supports will be designed and fabricated in accordance with Subsection NF of Section III of the ASME Code. Once finalized by the NRC, any new requirements of NUREG-0577 and the proposed SRP beyond ASME Code requirements will be reviewed for impr.ct, and the level of compliance will be documented as part of the FDA ap% ication for the SP/90 design.

O WAPWR-RC 4.0-21 AMENDMENT 3 E944e:1d AUGUST 1989 i

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13. Issue A-17: Systems Interactions in Nuclear Power Plants O

Discussion Systems interactions are those events that can occur in a plant due to one or multiple systems or components acting upon one or more other systems in a manner not intended by design.

The design and analyses by the plant designers, and the subsequent review and evaluation by the NRC staff take into consideration the inter-disciplinary areas of concern and account for systems interaction to 3 a large extent. National standards and regulatory criteria provide requirements that, if met, reduce the probability of adverse systems interactions.

3 Nevertheless, there is some question regarding the interaction of various plant systems, both as to the supporting roles such systems play and as to the effect one system can have on other systems, particularly with regard to whether actions or consequences could adversely affect the presumed redundancy and independence of safety systems.

g The problem to be resolved by this task is to identify where the present design, analysis, and review procedures may not acceptably account for potentially adverse systems interaction and to recommend the regulatory action that should be taken.

The NRC scope of work to resolve this issue includes:

3 ]

i o Search for common cause events. Oi I

o Review trends / patterns of cocinon cause events. I o Comparison with Indian Point 3 systems interactions experience.

o Screen common cause events for safety significance.

1 WAPWR-RC 4.0-22 AMENDMENT 3 0'

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i o Review and define sec.rch methods. )

1 J o Compare and evaluate search methods.

SP/90 Response The systems interaction concern is a major consideration being addressed in the SP/90 design. The SP/90 design incorporates several features that will reduce the probability of any adverse interactions occurring. These features include safeguards fluid system designs with reduced or elimina-ted interconnections, reduced or eliminated normal operation functions, and improved redundancy and diversity. With regard to instrumentation and control, the protection and control functions are performed by separate systems; this separation extends to the DC and vital AC power supplies.

3 Furthermore, the SP/90 plant layout provides improved physical separation between redundant divisions of safety related equipment as well as between safety related equipment and the control systems.

O o ee i or the se'eo >> t 8 1 9 i te 8er 18 > tem, inte ctie -

issue early in the design phase. All systems interactions that have been identified in th(. past are being addressed by either hardware changes or

, analyses to show the applicable safety criteria are met. Also, a key consideration in the plant layout, safety system design, and equipment selection is to avoid any unacceptable systems interactions.

In addition to considering systems interactions in the design phase of the plant, a comprehensive systems interactions analysis will be performed as part of the SP/90 final design. A description of the systems interaction study to be performed will be documented in the FDA application.

1 WAPWR-RC 4.0-23 AMENDMENT 3 5944e:1d AUGUST 1989

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14. Issue A-24: Qualification of Class 1E Safety-Related Equipment '

O Discussion l

l This NRC task is concerned with developing adequate design criteria for  !

electrical equipmer.t in safety systems such that it will perforn its function in adverse environmental conditions as a consequence of certain postulated accidents. The NRC requires that such equipment (principally equipment associated with the emergency core cooling, containment isola-tion, and cleanup systems) be environmentally qualified.

Specific electrical equipment of concern during postulated accident conditions includes: I o Instrumentation needed to initiate the safety systems and provide diagnostic information to the plant operators (e.g., electrical penetrations into containment, any electrical connectors to cabling which transmit signals, and the instruments themselves). J o Control power to motor operators for certain valves (e.g., emergency core cooling and containment isolation valves located inside containment).

o Fan cooler motors for those plants that utilize fan coolers for containment heat removal.

NUREG-0588, Revision 1 " Interim Staff Position on Environmental Qualification 'of Safety-Related Equipment," establishes the methods and procedures to be used to environmentally qualify safety-related electrical equipment and supplements the requirements given in the 1971 and 1974 versions of IEEE Standard 323, " Standard for Qualifying Class 1E Equipment ,

for Nuclear Power Generating Stations." This NUREG does not address in .

detail all areas of the qualification issue, since some areas (e.g.,

effects of aging, sequential versus simultaneous testing, including I

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s h synergistic effects, and the potential for combustible gas and chloride formation in equipment containing organic materials) are not yet fully defined.

l In addition, the NRC has codified a new regulation, 10CFR Part 50.49,

" Environmental Qualification of Electric Equipment Important to Safety for Nuclear Power Plants," January, 1983. It prescribes aging and testing for i

synergistic effects. Each holder of an operating license was required by Way 20, 1983, to identify the electric equipment important-to-safety already qualified and submit a schedule for the qualification or replace-ment of the remaining electric equipment important-to-safety. The final environmental qualification of the electric equipment was required by the end of the second refueling outage occurring after March 31, 1982 or by March 31, 1985, whichever is earlier. Applicants for operating licenses were required to perform an analysis to ensure that the plant can be safely operated, pending completion of equipment qualification required by this section. The rule requires that:

o A program shall be established for qualifying electric equipment and systems that are essential to emergency reactor shutdown, containment isolation, reactor core cooling, and containment and reactor heat

- removal or that are otherwise essential in preventing significant release of radioactive material to the environment.

o All electric equipment covered by this rule shall be listed and this list shall be maintained in auditable form.

o The electrical equipment qualification program must include temperature and pressure, humidity, chemical effects, radiation, aging, submergence, synergistic effects, and margins.

h o Electric equipment must be qualified by testing an identical item of equipment, testing a similar item of equipment (with a supporting analysis to show acceptability), experience with identical or similar O

' WAPWR-RC 4.0-25 AMENDMENT 3 5944e:1d AUGUST 1989

equipment under similar conditions (with a supporting analysis to show acceptability), or analysis in lieu of testing (if type testing is precluded by the physical size of the equipment or by the state-of-the-art).

o A record of the qualification must be maintained in an auditable form to permit verification that each item of electric equipment is quali-fied for its application and meets the specified performance requirements.

Also, the NRC has issued Revision 1 to Regulatory Guide 1.89, " Environ-mental Qualification of Electric Equipment for Nuclear Power Plants," for comment that describes a method acceptable to the NRC staff to demonstrate compliance with the requirements of 10CFR 50.49.

SP/90 Response Westinghouse has an ongoing environmental qualification program which has resulted in successful qualification of electrical equipment for recently licensed plants. Currently qualified equipment which is intendec to be used for the SP/90 design will be reassessed relative to its position within containment, and any anticipated changes in the potential environ-ment it will experience. Analyses will be performed to demonstrate that the current Westinghouse generic envelope is valid for the SP/90 design.

In addition, a detailed SP/90 environmental qualification report will be '

prepared which will address all of the documentation requirements of the current rulemaking. Finally, Westinghouse will completely document and l justify any deviations from the NRC Regulatory Guide 1.89 positions in the FDA application for the SP/90 design.

g 1

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15. Issue A-26: Reactor Vessel Pressure Transient Protection.

Discussion h Since 1972, there have been numerous reported incidents of " pressure transients" at various pressurized water reactor facilities. A pressure transient occurs when the pressure-temperature limits included in the -

technical specifications for the facility have been. exceeded. There has p been greater than 33 such events. Half of these events occurred before b the plant achieved initial criticality (i.e., before initial operation of ]

the reactor); the majority occurred during startup or shutdown operations.

In all of these incidents fracture mechanics and fatigue calculations indicated that the reactor vessels were not damaged and continued operation of the vessels was acceptable. Nevertheless, the NRC concluded that appropriate regulatory actions were necessary to reduce the frequency >

of pressure transient events and restrict future transients to acceptable pressures. The NRC deemed that action was necessary to conserve reactor vessel safety margins over the lifetime of the vessel.

O The NRC staff's review of this safety issue was completed in September 1978 with the issuence of NUREG-0224, " Reactor Vessel Pressure Transient Protection for Pressurized Water Reactors."

Upgraded procedural controls were implemented at operating pressurized water reactor facilities which significantly reduced the occurrence of pressure transient events. In addition, most operating plants incorpor-ated equipment modifications involving the addition of a second lower set point on existing power-operated relief valves, the addition of new spring-loaded relief valves, or modifications to allow use of existing spring-loaded relief valves.

Branch Technical Position RSB 5-2, "Overpressurization Protection of Pressurized Water Reactors While Operating at Low Temperatures," ,

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establishes the current NRC acceptance criteria for a low temperature  :

l_

overpressurization protection system.

l In summary, Branch Technical Position RSB 5-2 states that: )

I o A system should be designed and installed which will prevent exceeding the applicable technical specifications and 10CFR Part 50, Appendix G l

l limits for the reactor coolant system while operating at low l temperatures.

o The system should be able to perform its function assuming any single O '

l active component failure. Analyses using appropriate calculational techniques must be provided which demonstrate that the system will provide the required pressure relief capacity assuming the most j limiting single active failure. The cause for initiation of the event (e.g., operator error, component malfunction) should not be considered as the single active failure. The analyses should assume the most limiting allowable operating conditions and systems configuration at the time of the postulated cause of the overpressure event.

o The system should be designed using IEEE Standard 279-1971, " Criteria for Protection Systems for Nuclear Power Generating Stations," as guidance.

o To assure operational readiness, the overpressure protection system should be testable.

o The system must meet the requirements of Regulatory Guide 1.26, g

" Quality Group Classifications and Standards for Water , steam , and W Radioactive-Waste-Containing Components of Nuclear Power Plants," and Section III of the ASME Code.

o The overpressure protection system should be designed to function during an operating basis earthquake.

i WAPWR-RC 4.0-28 AMENDMENT 3 ,

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I q o The overpressure protection system should not depend on the'

.V availability of offsite power to perform its function.

o Overpressure protection systems which take credit for an active

.g component (s) to mitigate the consequences of an overpressurization

() event should include additional analyses considering inadvertent system initiation / actuation or provide justification to show that existing analyses bound such'an event.

I o If pressure relief is from a low pressure system, not normally connectad to the primary system, the overpressure protection function should not be defeated by interlocks which would isolate the low pressure system from the primary coolant system.

Low temperature overpressure protection systems were implemented on all PWR operating plants.

3 SP/90 Response O The SP/90. design does include a low temperature overpressurization protection capability. Westinghouse has not identified any deviations 3 from the NRC Branch Technical Position RSB 5-2 acceptance criteria.

16. Issue A-31: Residual Heat Removal Requirements Discussion The safe shutdown of a nuclear power plant following an accident not O' related to a loss-of-coolant accident has been typically interpreted as achieving a " hot standby" condition (i.e., the reactor is shutdown, but system temperature and pressure are still at or near normal operating values). Considerable emphasis has been placed on the hot standby s condition of a power plant in the event of an accident or abnormal occurrence. A similar emphasis has been placed on long-term cooling.

O WAPWR-RC 4.0-29 AMENDMENT 3 5944e:1d AUGUST 1989 L________-----_--__-..

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Even though it may generally be considered safe to maintain a reactor in a hot standby condition for a long time, experience shows that there have been events that required eventual cooldown and long-term cooling until h

the reactor coolant system was cold enough to perform inspection and repairs. For this reason the ability to transfer heat from the reactor to the environment after" a shutdown is an important safety function.

Therefore, the NRC beliaves it is essential that a power plant be able to h

go from hot standby to cold shutdown conditions (when this is determined to be the safest course of action) under any accident conditions.

This NRC task is concerned with establishing specific design requirements for the systems that are employed to achieve and maintain a safe shutdown including cooldown from hot standby to cold shutdown (e.g., reactor coolant system, main steam system, auxiliary feedwater system, chemical and volume control system, borated refueling water system, residual heat removal system, component cooling water system, essential service water system, supportive heating, ventilation and air conditioning systems, emergency diesel generators, spent fuel cooling system, and supportive portions of the instrument air system).

Regulatory Guide 1.139, " Guidance for Residual Heat Removal," Branch Technical Position RSB 5-1, " Design Requirements of the Residual Heat Removal System," and Standard Review Plan 5.4.7, " Residual Heat Removal (RHR) Systems,"containregulatorypositions and acceptance criteria for the system (s) used to take the reactor from normal operating conditions to cold shutdown. Specifically, the system (s) must:

o Be safety grade o Be single failure proof o Function with or without offsite power o Be capable of being operated from the control room o Be capable of achieving cold shutdown within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> O

WAPWR-RC 4.0-30 AMENDMENT 3 5944e:1d AUGUST 1989

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,A 'In addition, the residual heat removal pump system must meet specific isolation, pressure relief, pump protection, and testability requirements. .)

l SP/90 Response j

-p The SP/90 design includes safety grade cold shutdown capability.

l

)

Westinghouse 'has not identified any deviations from the NRC Regulatory l Guide 1.139 Branch Technical Position RSB 5-1, and Standard Review Plan 5.4.7 positions and acceptance criteria.

17. Issue A-36: Control of. Heavy Loads Near Spent Fuel .

Discussion i Overhead handling systems (cranes) are used to lift heavy objects in the vicinity of spent fuel in light-water-cooled nuclear power plants. If a heavy object (e.g., a spent fuel shipping cask or shielding block) were to fall or tip onto spent fuel in the storage pool or the reactor. core and damage the fuel, there could be a release of radioactivity to the environ-ment and a potential for radiation overexposure to inplant personnel. If many fuel assemblies are damaged, and the damaged fuel contained a large amount of undecayed fission products, radiation releases to the environ-ment could exceed the guidelines of 10CFR Part 100, " Reactor Site Criteria."

Additionally, a heavy object could fall on safety-related equipment and prevent it from performing its intended function. If equipment from redundant shutdown paths were damaged, safe shutdown capability may be O defeated.

The purpose of this task was to provide an evaluation of current NRC requirements and existing licensee design esasures, operating procedures, and technical specifications associated with the movement of heavy loads near spent fuel pools inside or outside containment, and over the reactor O WAPWR-RC 4.0-31 AMENDMENT 3 5944e:1d AUGUST 1989

s core during refueling. The current NRC requirements and review procedures in effect at the time this issue was identified, were given in Standard Review Plans 9.1.2, " Spent Fuel Storage," 9.1.4, " Light Load Handling System (Related to Refueling)," 15.7.4, " Radiological Consequences of Fuel Handling Accidents," and 15.7.5, " Spent Fuel Cask Drop Accidents." These Standard Review Plans provide procedures for review of the spent fuel storage pool, the fuel handling system, radiological consequences of fuel handling accidents, and spent fuel cask drop accidents. Regulatory Guide 1.13, " Spent Fuel Storage Facility Design Basis," provides additional guidance in this area. Further, the Standard Technical Specifications, included in all new operating licenses, include a prohibition on the movement of loads over spent fuel in the storage pool that weigh more than the equivalent weight of a fuel assembly. These load restrictions have been successfully demonstrated, for recently licensed plants, as providing assurance that miscellaneous loads (which have not been reviewed from the standpoint of rigging) will not be carried over stored fuel, and in the i event such loads are dropped, radioactivity release is limited and critical array does not result from rack distortion.

Although it is the NRC's view that continued operation with currently licensed facilities' designs, operating procedures, and technical speci-fication limitations that meet the criteria listed above presents no undue risk to the health and safety of the public, the advent of increased (higher density storage configurations) and longer term storage of spent fuel assemblies in spent fuel storage pools caused the NRC to reevaluate the above requirements.

As a result o

'f this reevaluation the NRC expanded this issue to also include the control of heavy loads over safe shutdown equipment (i.e.,

safety-related equipment and associated subsystems that would be required to bring the plant to cold shutdown conditions or provide continued decay heat removal following the dropping of a heavy load). The NRC has docu-  !

mented their technical resolution of this issue in NUREG-0612 " Control of I

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/~3 Heavy Loads at Nuclear Power Plants," and issued Standard Review Plan.

9.1.5, " Overhead Heavy Load Handling Systems," which includes NUREG-0612 as one of the NRC acceptance criteria.

3 SP/90 Response

.qp) .

No deviations from SRP Section 9.1.5 have been identified for the SP/90 design. The OHLHS design satisfies the requirements of General Design 3 p Criteria 4 and 61, and follows the guidelines of Regulatory Guides 1.13

( and 1.29.

18. Issue A-39: Determination of Safety Relief Valve Pool Dynamic Loads and Temperature Limits for BWR Containments This issue is not applicable to Westinghouse pressurized water reactor designs.

19. Issue A-40: Seismic Design Criteria - Short Term Program Discussion

, NRC regulations require that nuclear power structures, systems, and components important to safety be designed to withstand the effects of natural phenomena such as earthquakes. Detailed requirements and guidance regarding the seismic design of nuclear plants are provided in NRC regulations and regulatory guides. The seismic design process required by current NRC criteria includes the following sequence of events:

O (a) Define the magnitude or intensity of the earthquake which will produce the maximum vibratory ground motion at the site (the safe shutdown earthquake or SSE).

(b) Determine the free-field ground motion at the site that would result if the SSE occurred.

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(c) Determine the motion of site . structures by modifying the free-field motion to account for the interaction of the site structures with the underlying foundation soil.

I (d) Determine the motion of the plant equipment supported by the site structures.

(e) Compare the seismic loads, in appropriate combination with other loads, on structures, systems and components important to safety, with '

the allowable loads.

While this seismic design sequence includes many conservative factors, certain aspects of the sequence may not be conservative for all plant sites. At present, it is believed that the overall sequence is adequately conservative. The objective of this program is to investigate selected areas of the seismic design sequence to determine their conservatism for all types of sites, to investigate alternate approaches to parts of the 3 design sequence, to quantify the overall conservatism of the design sequence, and to modify the NRC criteria in the Standard Review Plan if changes are found to be justified. In this manner, this program will g

provide additional assurance that the health and safety of the public is protected, and if possible, reduce costly design conservatism by improving current seismic design requirements, and by improving NRC's capability to quantitatively assess the overall adequacy of seismic design i for nuclear plants in general. I The program for resolution of USI A-40 consists of two phases: (1) tasks concerning se'ismic input definitions, and (2) tasks concerning the response of structures, systems, and components. All technical work has g

been completed on both phases. The technical work accomplished on each of the tasks is summarized and evaluated in NUREG/CR-1161, " Recommended Revisions to Nuclear Regulatory Commission Seismic Design Criteria."

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I A review of'the recommendations by the NRC staff 'has been made and a

. proposed . staff position developed in a form of recommended changes.in SRP for revision. In some cases, the recommendations have already been incorporated into the SRP in a general revision in 1981.

O Further SRP changes, including alternatives to the staff position on lj soil-structure interaction (SSI) are under consideration. l SP/90 Response The problem, as defined under this issue, is. that while the presently  ;

established seismic design sequence includes many conservative factors, certain aspects of the sequence may not be conservative for all plant sites.

For the Westinghouse SP/90, it is anticipated that the resolution of.this 3 issue will show that its design will have adequate conservatism.

Westinghouse employs generic, enveloping seismic design criteria and applies established seismic evaluation methodology which has been successfully applied to current plant designs and which meets current NRC regulations and regulatory guidance. Should the NRC criteria in the

, Standard -Review Plan be changed as a result of the resolution of this issue, and become applicable to the SP/90 design, the Westinghouse design will conform to modified criteria or to an acceptable alternate approach.

The manner in which this will be implemented will be decided, if and when, the criteria are modified.

20. Issue A-42: Pipe Cracks in Boiling Water Reactors This issue is not applicable to Westinghouse pressurized water reactors.

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21. Issue A-43: Containment Emergency Sump Performance i Discussion I This issue deals with a concern for the availability of adequate recirculation cooling water following a LOCA when long-term recirculation of cooling water from the PWR containment sump, or the BWR RHR system suction intake, must be initiated and maintained to prevent core-melt.

This water must be sufficiently free of LOCA generated debris and potential air ingestion so that pump performance is not impaired thereby )

seriously dograding long-term recirculation flow capability. The concern I applies to both PWRs and BWRs. )

The RHR suction strainers in a BWR are analogous to the PWR sump debris screen and adequate recirculation cooling capacity is necessary to prevent core-melt following a postulated LOCA. ]

The issue was declared a USI in January 1979 and published in NUREG-0510.

3 The technical concerns evaluated are as follows:

(1) PWR sump (or BWR RHR suction intake) hydraulic performance under post-LOCA adverse conditions resulting from potential vortex formation g

and air ingestion and subsequent pump failure.

(2) The possible transport of large , quantities of LOCA generated insulation debris resulting from a pipe break to the sump debris screen (s), and the potential for sump screen (or suction strainer) blockage to reduce net positive suction head (NPSH) margin below that required for the recirculation pumps to maintain long-term cooling.

(3) The capability of RHR and containment spray system (CSS) pumps to continue pumping when subjected to possible air, debris, or other effects such as particulate ingestion on pump seal and bearing systems.

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l The resolution of .SI A-43 was presented to the Commission in SECY-85-349.

The staff is implementing the resolution of USI A-43 through the following actions: )

I I (1)The staff's technical findings (NUREG-0897, Revision 1) were published I for use as an information source by applicants, licensees, and the staff.

(2) SRP Section 6.2.2 and Regulatory Guide 1.82 were revised to reflect V the staff's technical findings reported in NUREG-0897, Revision 1.

This revised licensing guidance applies only to reviews of: l (a) future construction permit applications and preliminary design j approvals (PDAs); (b) final design approvals (FDAs) for standardized designs which are intended for. referencing in future construction l permit applications that have not received approval; and (c) applica- l 3

tions for licenses to manufacture. -This revised guidance became effective 6 months after issuance of Regulatory Guide 1.82, Revision 1. i 1

(3) Generic Letter 85-22 (for information only) was sent to all holders of j an operating license or construction permit outlining the safety concerns regarding potential debris blockage and recirculation failure due to inadequate NPSH. It was recommended (but not required) that licensees utilize Regulatory Guide 1.82, Revision 1, as guidance for conduct of the 10 CFR 50.59 analysis for future plant modifications involving replacement of insulation on primary system piping and/or equipment. If, as a result of NRC staff review of licensee actions i associated with replacement or modification to insulation, the staff decides that SRP 6.2.2, Rev. 4 and/or Regulatory Guide 1.82, Rev. 1,

\ criteria should be applied by the licensees, and the staff seeks to impose these criteria, then NRC will treat such actions as plant-specific backfits pursuant to 10 CFR 50.109.

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SP/90 Response An important design feature of the SP/90 is the in-containment Emergency O

Water Storage Tank (EWST) which has replaced the conventional outside-containment refueling water storage tank (RWST). The four SP/90 ECCS subsystemy and the four SP/90 containment spray subsystems are initially aligned to draw water from the in-containment EWST therefore no realign-h ment of these systems is required for long term recirculation. The incontainment EWST is an annular tank which is located below the contain-ment floor and sized to contain sufficient borated water to fill the  ;

refueling canal during refueling operations.

Fo11 ming a postulated loss of coolant accident, water discharged from the break would; (1) collect on the containment floor; (2) flood all compart-ments belew the containment floor elevation such as the reactor vessel cavity, and (3) spill back into the EWST via several physically separated spillways located in the containment floor and outside the loop compart-ments. Each spillway is protected by rough screens and trash racks to prevent debris from entering the EWST. The elevation of each spillway is several inches above the containment floor, therefore, the containment floor serves as a large settling pond for the recirculation water.

3' Inside the EWST, there are four physically separate EWST sump pits located below the EWST floor elevation. Each sump pit is dedicated to one of the four ECCS and containment spray subsystems. Rough screens and fine screens are provided at each of the four EWST sump pits in addition to the rough screens and trash racks provided at each of the EWST spillways. The EWST therefore' serves as a second settling pond for the recirculation water.

Evaluations are performed to establish the minimum post accident EWST water levels and to verify that the ECCS and containment spray pump net g positive suction head (NPSH) requirements are satisfied for all normal or W i WAPWR-RC 4.0-38 AMENDMENT 3 B944e:1d AUGUST 1989 l

l accident system operation. The. SP/90 EWST configuration therefore meets 6 all NPSH and sump design requirements currently specified in SRP 6.2.2g NRC Regulatory Guide 1.1, and 1.82.

The SP/90 EWST configuration in conjunction with the ECCS piping 9 configuration also provides a unique means for performing full flow system verification not only during preoperational testing but performance anytime during the plant life. Therefore all operational verification requirements specified in NRC Regulatory Guide 1.79 are satisfied by the 9 SP/90 design.

22. Issue A-44: Station Blackout Discussion The complete loss of AC electrical power to the essential and nonessential switchgear buses in a nuclear power plant is referred to as a " Station Blackout." Because many safety systems required for reactor core decay heat removal are dependent on AC power, the consequences of a station blackout could be a severe core damage accident. The technical issue involves the likelihood and duration of the loss of all AC power and the potential for severe core damage after a loss of all AC power.

The issue of station blackout arose because of the historical experience 3 regarding the reliability of AC power supplies. There had been numerous reports of emergency diesel generators failing to start and run in operating plants. In addition, a number of operating plants experienced a total loss of offsite electrical power. In almost every one of these loss of offsite power events, the onsite emergency AC power supplies were available to supply the power needed by vital safety equipment. However, in some instances, one of the redundant emergency power supplies had been available. In a few cases, there was a complete loss of AC power, but 9 during these events AC power was restored in a short time without any serious consequences.

O WAPWR-RC 4.0-39 AMENDMENT 3 5944e:1d AUGUST 1989

The results of WASH-1400 showed that, for one of the two plants evaluated, a station blackout accident could be an important contributor to the total risk from nuclear power plant accidents. Although this total risk was found to be small, the relative importance of station blackout accidents was established. This finding and the concern for diesel generator reliability based on operating experience raised station b'lackout to a USI in the 1979 NRC Annual Report. A detailed action plan for resolving this issue was published in NUREG-0649, Revision 1.

The final evaluation of station blackout accidents at nuclear power plants was performed by the staff and published in NUREG-1032. In resolving this h

issue, the staff performed a regulatory analysis which was documented in NUREG-1109. In June 1988, this USI was resolved with the publication of a new rule (53 FR 23203) and Regulatory Guide 1.155. Thus, this issue was RESOLVED and new requirements were established.

3 SP/90 Response The SP/90 design includes the following design features specifically aimed at mitigating the consequences of a station blackout.

o The emergency feedwater system (EFWS) includes two turbine-driven emergency feedwater pumps. These pumps are independent of AC and DC, and the rooms they are located in are cooled in a passive manner.

Only one of two pumps is required for decay heat removal.  ;

o The chemical and volume control system (CVCS) contains a backup seal injection pump. This pump takes suction from the spent fuel pit and -l is powered from a small (~100 kW) diesel generator which is l independent of off-site and on-site AC power supplies. The diesel l generator and pump are started automatically on loss of normal seal )

injection. .

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'o 'ne Class 1E batteries are sized ~for four . hours of- operation under O blackout conditions. -This assumes normal operation for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and

]

selective load shedding by the operators thereafter.-

Permanently installed connections are provided between the power l o l

• source of the backup seal injection pump and the Class 1E batteries.

This will allow the operators to recharge the batteries in order to L maintain vital functions such as monitoring of RCS and SG parameters and emergency lighting.

o Emergency response guidelines will be developed as part of the FDA application to- ensure correct operator action during station blackout. These will cover the operation of the above equipment, as 3 well as any other equipment that may be useful in a station blackout condition.

These features allow the plant to be maintained at hot standby conditions for approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. At that time, additional water supplies need to be provided in order to allow continued operation of. the emergency feedwater and backup seal injection pumps.

Westinghouse believes that the SP/90 design exceeds the requirements of the final rule 10 CFR50.63 and Regulatory Guide 1.155, and that therefore the Station Blackout Issue should be resolved for this plant.

23. Issue A-45: Shutdown Decay Heat' Removal Requirements Discussion In March 1981, this issue was identified as a USI in NUREG-0705. A program was initiated to evaluate the safety adequacy of the decay heat

/ removal (DHR) function in operating LWRs and to assess the value and 3 impact (i.e., the benefit and cost) of alternative measures to improve the overall reliability of the DHR function.

O WAPWR-RC 4.0-41 AMENDMENT 3 8944e:1d AUGUST 1989

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The program employed PRAs and . deterministic evaluations of those DHR l

systems and support systems required to achieve hot shutdown and cold '

shutdown conditions in both PWRs and BWRs. Systems analysis techniques were used to assess the vulnerability of DHR systems to various internal and external events. The analyses were limited to transients, small-break LOCAs, and special emergency challenges such as fires, floods, earth- I quakes, and sabotage. Cost-benefit analysis techniques were used to assess the net safety benefit and cost of alternative measures to improve the overall reliability of the DHR function.

Six plants were analyzed after an initial selection process which O

considered vendor, product line, other issues in which each particular plant might be involved, operational status, and utility willingness to participate. In terms of expected core damage frequency caused by DHR failure /RY, the range for the six plants was 7 x 10-5 to 4 x 10 ~4 with an average value of 2 x 10~4 if credit is allowed for feed-and-bleed 3 operation on the PWRs and containment venting on the BWRs. Neither the above ranges nor the averages were found to be significantly changed when several other existing, reliable PRA results were also included.

g Based on its analysis, the staff found that the six plants met the health effcets quantitative objectives in the Commission's Safety Goal (i.e.,

O.1% of the expected accident or cancer fatality risks from causes .iot related to nuclear plants). Guidance for an acceptable core damage frequency was not explicitly provided. However, in order to provide assurance that (1) core damage due to a DHR failure-related event will not occur in the lifetime of the present population of plants, (2) consistency

-5 is maintained with the 10 /RY contribution to core damage frequency from station blackout expected after resolution of USI A-44, and (3) the frequency of a severe release will be less than the Commission's safety

~0 goal guidance of 10 /RY, the staff selected a goal that core damage due to failure of DHR function should be less than 10 -5 /RY. This x staff-selected goal was intended only for application to the resolution of USI A-45. The results indicated that the DHR related frequency of core damage at certain plants may be considerably above this goal. To address i WAPWR-RC 4.0-42 AMENDMENT 3 1

5944e:Id AUGUST 1989 l

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l the question of whether corrective actions could be cost-effective, six possible alternatives addressing potential DHR vulnerabilities were I

identified and then evaluated.

In presenting its proposed for resolution of USI A-45 in SECY-88-260, the

(

' staff recognized the ongoing actions in implementing the Commission's Severe Accident Policy, one of which was a generic letter to require all plants in operation or under construction to undergo a systematic Q examint. tion termed the Individual Plant Examination (IPE) to identify any O plant-specific vulnerabilities to severe accidents. The IPE analysis is intended to examine and understand the plant emergency procedures, design, operations, maintenance, and surveillance to identify vulnerabilities.

The analysis will examine both the DHR systems and those systems used for other functions. It is anticipated that a future extension of the IPE program will require examination of externally-initiated events, some of which significantly contribute to DHR failure-related core damage frequency.

3 To resolve USI A-45, one of the alternatives proposed by the staff was to have each licensee perform a risk assessment for its plant. This assessment would be done in conjunction with the IPE program. Available

. options for acceptable risk assessments include performing a Level-1 PRA (enhanced) or performing an analysis using the IDCOR IPEM. Thus, USI A-45 was RESOLVED with the requirement for plant-specific analyses to be conducted under the IPE program.

SP/90 Response The SP/90 has several systems which have the capability to remove decay heat from the reactor core. The SP/90 secondary side safeguards system will employ an emergency feedwater system combined with a startup

( feedwater system. This system provides two independent systems operated from diverse energy sources which serve to remove decay heat from the primary system via the steam generators to the secondary system. If either the EFWS or the SFWS cannot remove decay heat via the steam O

WAPWR-RC 4.0-43 AMENDMENT 3 E944e:1d AUGUST 1989

s generators, then the Integrated Safeguards System (ISS) can be used. The ISS incorporates designed-in " feed and bleed" capability. Bleed is achieved by opening safety grade pressurizer power operated relief valves, which establishes a path from the pressurizer to the pressurizer relief tank to the in-containment emergency water storage tank (EWST). The high head safety injection (HHSI) pumps take suction from the EWST and feed borated water back to the primary system, thereby maintaining inventory.

The ISS also includes the residual heat removal function which can be used in the long term after primary system pressure and temperature have been reduced sufficiently.

The primary side and secondary side safeguards systems for the SP/90 design (as discussed above) provide the capability of removing decay heat 3 from the reactor core while maintaining sufficient water inventory to ensure adequate core cooling. The combination of these systems is highly reliable because it incorporates both redundancy and diversity.

RESAR-SP/90 PDA Module 16, "Probabilistic Safety Study" indicates that the core melt frequency due to internal events is 1.5E-6 per year. Less than half of SP/90 core, melts are due to RHR (DHR) system failures and the goal g of 1.0E-5 per year is therefore exceeded by better than one order of W magnitude. The design of the SP/90 RHR systems is therefore not impacted by the resolution of this USI.

24. Issue A-46: Seismic Qualification of Equipment in Operating Plants Discussion The design criteria and methods for the seismic qualification of mechanical and electrical equipment in nuclear power plants have undergone significant change during the evolution of the commercial nuclear power h

industry. Consequently, the margins of safety provided in existing equipment to resist seismically induced loads and perform the intended safety functions may vary considerably. The NRC believes that the seismic qualification of equipment in operating plants must, therefore, be h

WAPWR-RC 4.0-44 AMENDMENT 3 B944e:Id AUGUST 1989 L_____._____ _

reassessed to ensure the ability to bring the plant to a safe shutdown (V] condition when it is subject to a seismic event. The objective of the NRC task program to address this issue was to establish an explicit set of guidelines that could be used to judge the adequacy of the seismic qualification of mechanical and electrical equipment .at all operating l N plants'in lieu of attempting to backfitting of the current design criteria that apply to new plants. This guidance will concern equipment required to safely shst down the plant, as well as equipment whose function is not c, required for safe shutdown, but whose failure could result in adverse conditions which might impair shutdown functions.

Work completed on this USI resulted in the publication of NUREG/CR-3017, NUREG-CR-3875, NUREG/CR-3357, NUREG/CR-3266, NUREG-1030, and NUREG-1211.

The resolution of USI A-46 was mainly based en work completed by the Seismic Qualification Utility Group (SQUG) and EPRI using the seismic and test experience data approach and reviewed and endorsed by the Senior Seismic Review and Advisory Panel (SSRAP) and the NRC staff. The scope of the review was narrowed down to equipment required to bring each affected plant to hot shutdown and maintain it there for a minimum of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. A walk-through of each plant is required to inspect equipment in the scope.

Evaluation of equipment will include: (a) adequacy of equipment 3 anchorage; (b) functional capability of essential relays; (c) outliers and deficiencies (i.e., equipment with non-standard configurations); and (d) seismic systems interaction. This issue was RESOLVED and requirements were issued in Generic Letter 87-02 in February 1987.

SP/90 Response

' In response to this issue and other industry needs Westinghouse has developed a seismic requalification program for operating plants. This program consists of a data search to ident.ify what type of seismic qualification (if any) is available, use of existing data to qualify equipment by similarity, qualification by analysis, testing, or a combination of analysis and testing. Current NRC acceptance criteria and l l

l O 4.0-45 AMENDMENT 3 WAPWR-RC E944e:1d AUGUST 1989

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regulatory guidance for new plant designs are given in Standard Review Plan 3.10 " Seismic and Dynamic' Qualification of Mechanical end Electrical Equipment," and Regulatory Guide 1.100, " Seismic Qualification of Electric h

Equipment for Nuclear Power Plants," which endorse IEEE Standard 344-1975, "Recomended Practices for Seismic Qualification of Cless 1E Equipment for Nuclear Power Generating Stations."

For Westinghouse plants designed and built to the requirements of IEEE Standard 344-1975, seismic design requirements implemented by Westinghouse for equipment important to safety are consistent with the latest NRC regulatory requirements. Additionally, the NRC has found the methods used by Westinghouse acceptable on a number of recent plant applications.

3 As indicated above, this issue is primarily concerned with plants licensed prior to the issuance of current NRC regulatory requirements regarding seismic qualification, and therefore, the ultimate resolution of this issue is not expected to impact the SP/90 design.

Specifically for the SP/90, Westinghouse will: (A) completely document g and justify any deviations from the NRC Standard Review Plan 3.10 and W Regulatory Guide 1.100 seceptance criteria and positions in the FDA application for the SP/90 design, and (B) apply established Westinghouse seismic qualification methods which have been successfully applied to current plant designs and which meet current NRC regulatory requirements.

25. Issue A-47: Safety Implications of Control Systems Discussion Instrumentation and control systems utilized by nuclear plants are O

composed of safety grade protection systems and non-safety grade control systems. Safety grade systems are used to (1) trip the reactor when specified parameters exceed allowable limits; and (2) protect the core from overheating by initiating ECCS systems. Non-safety grade control i

WAPWR-RC E944e:1d 4.0-46 AMENDMENT 3 O1I AUGUST 1989

i 3

systems are used to maintain the plant within prescribed parameters during shutdown, startup and normal load varying power operation. Non-safety grade systems are not relied on to perform any safety functions during or

!. following postulated accidents, but are used to control plant processes.

/ Although non-safety grade control system failures are not likely to result k- in accidents or transients that could lead to serious events or result in conditions that safety systems are not able to cope with, in-depth studies have not been performed. Concerns have been identified in which a failure or malfunction of the non-safety grade control system can (1) potentially cause steam generator or reactor vessel overfill; or (2) can lead to a transient that could cause severe vessel overcooling. In addition, there I is the potential for control system failures to result in plant conditions which may result in unacceptable risk.

The scope of work to resolve this issue includes:

1. Evaluate control system failures on four reference plant designs (one for each of the NSSS vendors) that could: (1) cause transients or accidents to be potentially more severe than those identified in the FSAR analysis; (2) adversely affect any assumed or anticipated opera-tor action during the course of atransientoraccident;(3)cause

. technical specification safety limits to be exceeded; or (4)cause transients or accidents to occur at a frequency in excess of those established for abnormal operational transients and design basis accidents. I I

1 risks posed by potentially significant control system

2. Evaluate failures identified during the review and assess the effectiveness of improvements in reducing those risks.

3. Evaluate risk reduction and costs of various fixes to perform j value/ impact analysis of the failures.

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O l WAPWR-RC 4.0-47 AMENDMENT 3 5944e:1d AUGUST 1989 i

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4. Evaluate the generic applicability of the failures identified during ,

the review of the reference plants.

The objective of this USI is to evaluate the need for requiring control systems changes in operating reactors and to verify the adequacy of 3 licensing requirements and to propose, if needed, additional criteria and guidelines.

h A draft resolution package has been developed by NRC and will be issued for public comment.

h SP/90 Respense Current Westinghouse Condition II analyses of transient events of moderate frequency, that could be initiated by the single failure of a control system, show that the consequences meet acceptance criteria for Condition II events.

Since the functional requirements and design specifications for the SP/90 control systems will be no less stringent than these for current plants, it is expected that an analysis similar to that performed on recently licensed plants would likewise show that the consequences of failures in control systems of the SP/90 would be bounded by FSAR type analyses.

Consequently, no hardware impacts on the SP/90 control systems are anticipated. However, a control system failure study, as part of an overall systems interactions study, will be performed and documented j 3 during the FDA application for the SP/90 design. The objectives of this study are to.

O\

1 o Minimize the potential for reactor shutdown or safeguards system actuation by failures in reactor control or protection systems. l i

o Reduce the number of possible interactions between control and protection systems which could lead to a degraded accident condition. l WAPWR-RC 4.0-48 AMENDMENT 3 O

5944e:1d AUGUST 1989 l

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o Reduce the probability and consequences of failures in control systems l

\,s on plant safety and operability. l

26. Istue A-48: Hydrogen Control Neasures and Effects of Hydrogen Burns on

]

, Safety Equipment l

( l v 1 Discussion I 1

l Postulated reactor accidents which result in a degraded or melted core can result in generation and release to the containment of large quantities of hydrogen. The hydrogen is formed from the reaction of the zirconium fuel )

cladding with steam at high temperatures and/or by radiolysis of water.

Experience gained from the TMI-2 accident indicates that the NRC may {

require more specific design provisions for handling larger hydrogen releases than required by regulations (particularly for smaller, low 3 {

pressurecontainmentdesigns). I The purpose of the NRC task program to address this issue is to investigate means to predict the quantity and release rate of hydrogen following degraded core accidents and various means to cope with large releases to the containment such as inerting of the containment or controlled burning. The potential effects of proposed hydrogen control measures on safety including the affects of hydrogen burns on 3

safety-related equipment will be investigated.

The NRC has issued a revision to 10CFR Part 50.34, " Contents of Applications; Technical Information," which incorporates post-TMI

(^'

requirements into their regulations. This revision, known as the "CP/ML Rule," encompasses the issue of hydrogen control for new plant designs.

This issue was originally related to all types of LWR and containment  ;

combinations. With the publication of rule changes related to degraded 3 w cores, the issue was redefined to be only related to BWR/ Mark III and Ice Condenser Plants.

l 0 WAPWR-RC 4.0-49 AMENDMENT 3

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l 5944e:1d AUGUST 1989 j 1

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l SP/90 Response 3 This issue as currently defined is not applicable to SP/90. Refer to GSI 121 for hydrogen issues for large dry containments that arose after the rule changes were issued.

1

27. Issue A-49: Pressurized Thermal Shock I i

1 Discussion I Transients and accidents can be postulated to occur in pressurized water O1 l

l reactors (PWRs) that result in severe overcooling (thermal shock) of the reactor vessel, concurrent with high pressure. In these pressurized ther-

]

mal shock (PTS) events, rapid cooling of the reactor vessel internal

]

surface causes a temperature distribution across the reactor vessel wall that produces a thermal stress with maximum tensile stress at the inside surface of the vessel. The magnitude of the thermal stress varies with the rate cf change of temperature and is compounded by coincident pressure stresses.

~

PTS events are postulated to result from a variety of causes. These include system transients, some of which are initiated by instrumentation l and control system malfunctions (including stuck open valves in either the primary or secondary system), and postulated accidents such as small break loss-of-coolant accidents, main steam line breaks, and feedwater line breaks.

As long as the fracture resistance of the reactor vessel material is relatively high, these events are not expected to cause vessel failure.

However, the fracture resistance of the reactor vessel material decreases with the integrated exposure to fast neutrons. The rate of decrease is dependent on the chemical composition of the vessel wall and weld  !

materials. If the fracture resistance of the vessel has been reduced sufficiently by neutron irradiation, severe PTS events could cause small l

WAPWR-RC 4.0-50 AMENDMENT 3 O1 l E944e:1d AUGUST 1989

s flaw that might exist near the inner surface to propagate into the vessel wall. The assumed initial flaw might be enlarged into a crack through the vessel wall of sufficient extent to threaten vessel integrity and, therefore, core cooling capability.

The toughness state of reactor vessel materials can be characterized by a

" reference temperature for nil ductility transition" (RTNDT). As the temperature decreases, the metal gradually loses toughness over a tempera-

-/ ture range of about 100'F. RTNDT is a measure of where this toughness transition occurs. Its value depends on the material and the integrated neutron irradiation. Correlations, based on tests of irradiated speci-mens, have been developed to calculate the shift in RTNDT as a function of neutron fluence for various material compositions. The value of RT NDT at a given time in a vessel's life is used in fracture mechanics calculations to determine whether assumed pre-existing flaws could propa- '

gate as cracks when the vessel is subjected to overecoling events.

The NRC amended 10CFR 50.61 in July of 1985 to (1) establish a screening criterion related to the fracture resistance of PWR vessels during (PTS) events, (2) require analyses and schedule for implementation of flux reduction programs that are reasonably practicable to avoid exceeding the

. screening criterion, and (3) require detailed safety evaluations before plant operation beyond the screening criterion value.

The value of RTNDT can be selected so that the risk from PTS events for reactor vessels with smaller RTNDT values is acceptable. Higher values of RTNDT might also be shown to be acceptable, but the demonstration would require ' detailed plant-specific evaluations and possibly modifica-tions. A value for RTNDT as a screening criterion determines the need for, and timing of, further plant-specific evaluations.

3 A wide spectrum of postulated overcooling events could occur. Postulated events were grouped into categories, estimates were made of their expected frequency, and stylized characterizations of the temperature and pressure O

WAPWR-RC 4.0-51 AMENDNENT 3 5944e:1d AUGUST 1989 l

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< time-histories were developed for each event category. Estimates are based on a generic study of Westinghouse-designed pressurized water reac-tor systems, and are considered also to be generally representative of PWR f

systems designed by Combustion Engineering. Because there are some signi- I ficant differences between those designs and PWRs designed by Babcock &

Wilcox~that affect the characteristics and estimated frequencies of PTS events, information was also developed for the Babcock & Wilcox designs.

h By combining the estimated frequencies of postulated events with the probabilistic fracture mechanics results, some estimates of the proba-bility of vessel failure resulting from PTS events were developed. These h

estimates were used by the NRC to better understand the residual risks q inherent in the use of the screening criterion approach for further evaluations and resolution of the PTS issue <.

On the basis of these studies, the NRC staff concluded that PWR reactor pressure vessels with conservatively calculated values of RT NDT less than 270'F for plate material and axial welds, and less than 300*F for circumferential welds, present an acceptably low risk of vessel failure from PTS events. These values were chosen as the screening criterion.

g The RT f reactor vessels for some plants will remain below the

~

NDT screening criterion (acceptable) throughout the service life. For many q other reactor vessels, fuel management programs could be instituted that I would result in core configurations reducing neutron flux at critical l locations, thereby slowing the increase of RT NDT so that the screening criterion would not be exceeded. Further refinements in materials infor-mation, analyses of PTS event frequencies and scenarios, and plant-specific analyses of alternative measures to reduce PTS risk may provide a basis for continued operation with RT NDT values in excess of the screening criterion. The preparation and review of such analyses and j determination of their acceptability will require substantial time. g )

However, the effectiveness of flux reduction programs depend on early W implementation. Practicable flux reduction programs should be implemented l WAPWR-RC 4.0-52 AMENDMENT 3 O1I B944e:Id AUGUST 1989

to maintain reactor vessel RT NDT below the screening criterion, without waiting for possible plant-specific determinations for. higher values.

Licensees may submit additional plant-specific analyses to justify (new information, improved analyses or evaluations of alternative measures) operation with less restrictive flux reduction programs in the future.

When it is determined that even with flux reduction measures that the

~

vessel RTNDT is still projected to exceed the screening criterion, an .

analysis of the vessel fracture mechanics properties and including the effects on PTS risk will be required at least three years before the scresning criterion would be exceeded.

SP/90 Response Design improvements to the safeguards systems in the SP/90 will limit thermal shock to the reactor vessel during postulated accidents. The l 3 primary side safeguards system (i.e., the integrated safeguards system) will inject water to the reactor coolant system at temperatures significantly higher (e.g., about 100'F) than that at certain conventional 3 O operating plant designs during postulated loss-of-coolant accident conditions. This is due to the location of the suction water source being within the containment building where the temperature is expected to 3

always be greater than 100'F.

In addition, the improvements in the reactor vessel material specifications 3

and the reduced neutron exposure (as discussed in item 11 above) will further mitigate the impact of a thermal shock event on the reactor vessel.

The design improvements discussed above will make the SP/90 less susceptible to severe pressurized overcooling events than current operating plants. Therefore, no additional impact on the SP/90 design is anticipated as a result of the final resolution of this issue.

O WAPWR-RC 4.0-53 AMENDMENT 3 E944e:1d AUGUST 1989

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s 5.0 GENERIC SAFETY ISSUES - TASK ACTION PLAN CATEGORY A, B, C, AND D ISSUES, 3

NUREG-0371, NUREG-0471 As mentioned in Section 4.0, the NRC continuously evaluates the safety g requirements used in its, reviews against new information as it becomes

() available. In 1978 the NRC published NUREG-0410 "NRC Program for the Resolution of Generic Issues Related to Nuclear Power Plants." This NUREG identified over 130 specific generic safety issues and assigned each issue to one of four categories.

o Category A "Those generic technical activities judged by the staff to warrant priority attention in terms of manpower and/or funds to attain early resolution. These matters include those the resolution of which could: (A) provide a significant increase in assurance ot the health and safety of the public, or (B) have a significant impact upon the reactor licensing process."

O l

o Category B "Those generic technical activities judged by the staff to De important in assuring the continued health and safety of the public but for which early resolution is not required or for which the staff perceives a lesser safety, safeguards, or environmental significance than Category A matters."

o Category C "Those generic technical activities judged by the staff to have little direct or immediate safety, safeguards, or environmental significance, but which could lead to improved staff understanding of particular p\_/ technical issues or refinements in the licensing process."

O WAPWR AMENDMENT 3 B788e:1d 5.0-1 AUGUST 1989

l o Category D l

"Those proposed generic technical activities judged by the staff not O

to warrant the expenditure of manpower or funds because little or no importance to the safety, environmental, or safeguards aspects of nuclear reactors or to improving the licensing process can be attributed to the activity."

Since the issuance of NUREG-0410, certain generic safety issues have been resolvad with the issuance of regulatory criteria or guidance, and new generic safety issues have been identified.

New Generic Safety Issues have not been categorized in the same manner as previous issues. NUREG-0933, "A Prioritization of Generic Safety Issues,"

issued semiannually, establishes a priority ranking of HIGH, MEDIUM, LOW, and DROP for previously categorized generic safety issues as well as for newly identified issues.

o HIGH priority items are those for which an-important safety deficiency is involved, a substantial safety improvement is likely to be attainable at a reasonable cost, or the uncertainty of the safety assessment is unusually large.

3 o MEDIUM priority items are less demanding of early resolution but have a potential for substantial and worthwhile safety improvements or reductions in uncertainty of' analysis.

o LOW priority items are those for which little or no prospects exist for substantial and worthwhile safety improvements, h o DROP category covers issues which have been evaluated and are considered without merit or with negligible safety significance.

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WAPWR AMENDMENT 3 B78Be:1d 5.0-2 AUGUST 1989 l

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The above priority ranking has been applied to the previously categorized task plan items (NUREG-0410), as well as to new generic issues, TMI Action Plan items under development (NUREG-0660), and Human Factors Program Plan items (NUREG-0985).

.( The listing of open generic safety issues is based on the Generic Issue ManagementControlSystems(GIMICS) report,firstquarterFY-89 update. These include issues identified as High or Medium priority and nearly resolved issues.

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WAPWR AMENDMENT 3 B788e:1d 5.0-3 AUGUST 1989 l

5.1 CATEGORY A ISSUES Nost safety issues identified in Category A are referred to today as.Unre-solved Safety Issues which are discussed in detail in Section 4.0. However, the NRC assignment of an issue to Category A does not necessarily mean that the issue is safety significant, and accordingly, all Category A issues do not involve Unresolved Safety Issues.

Likewise, Category B, C, and D issues have been reevaluated and priority established in 6:cordance with NUREG-0933. The following discussions pertain 3 to high or medium priority Task Plan issues in relation to the SP/90 design.

1. Issue A-1: Water Hammer This issue is identified as an Unresolved Safety Issue. Refer to Section 4.0 for a discussion of Unresolved Safety Issues.

2. Issue A-2: Asymmetric Blowdown Loads on Reactor Primary Coolant Systems This issue is identified as an Unresolved Safety Issue. Refer to Section 4.0 for a discussion of Unresolved Safety Issues.

3. Issue A-3: Westinghouse Steam Generator Tube Integrity This issue is identified as an Unresolved Safety Issue. Refer to Section 4.0 for a discussion of Unresolved Safety Issues.

4. Issue A-4: Combustion Engineering Steam Generator Tube Integrity

5. Issue A-5: Babcock and Wilcox Steam Generator Tube Integrity

6. Issue A-6: Mark I Short Term Program

7. Issue A-7: Mark I Long Term Program 1

O WAPWR 5.1-1 AMENDMENT 3 B788e:1d AUGUST 1989

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8. Issue A-8: Mark II Containment Pool Dynamic Loads These issues (A-4 through A-8) are not applicable to Westinghouse O

pressurized water reactor designs.

9. Issue A-9: Anticipated Transients Without Scram This issue is identified as an linresolved Safety Issue. Refer to Section 4.0 for a discussion of Unresc.1ved Safety Issues.

10. Issue A-10: BWR Feedwater Nozzle Cracking This issue is not applicable to Westinghouse pressurized water reactor designs.

11. Issue A-11: Reactor Vessel Materials Toughness This issue is identified as an Unresolved Safety Issue. Refer to Section 4.0 for a discussion of Unresolved Safety Issues.

12. Issue A-12: Fracture Toughness of Steam Generator and Reactor Coolant Pump Supports This issue is identified as an Unresolved Safety Issue. Refer to Section 4.0 for a discussion of Unresolved Safety Issues.

13. Issue A-13: Snubber Operability Assurance Discussion h Snubbers are utilized primarily as seismic and pipe whip restraints at nuclear power plants. Their safety function is to operate as rigid supports for restraining the motion of attached systems or components under rapidly applied load conditions such as earthquakes, pipe breaks, h

and severe hydraulic transients.

WAPWR 5.1-2 AMENDMENT 3 5788e:Id AUGUST 1989

p. .-

Operating experience reports have shown that a substantial number of snub-bars have leaked hydraulic- fluid and the rejection rate from functional testing and inspection has been high. This lead to an NRC and ACRS con-corn regarding the effect of snubber malfunctions on plant' safety.

The NRC considers this issue as being technically resolved for pressurized water reactors with the issuance of:

o Standard Technical Specification 3/4.7.9, " Snubbers."

s o Standard Review Plan 3.9.3, "ASME Code Class 1, 2, and 3 Compon-ents, Component Supports, and Core Support Structures."

o Draft Regulatory Guide and Value/ Impact Statement, Task SC-708-4,

" Qualification and Acceptance Tests for Snubbers Used in Systems Important to Safety."

The following is a brief summary of the NRC criteria contained in the three sources for technical resolution of this generic issue:

o Standard Technical Specification 3/4.7.9 All safety-related* snubbers must be listed in the technical spec-ifications. Safety-related snubbers must be visually inspected for operability at certain intervals depending on the number of ,

inoperable snubbers found in the prior inspection. In addition, safety-related snubber types must be functionally tested at least rsnee per 18 months during shutdown.

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• The Standard Technical Specifications currently use the term " safety-related". Indications are that the NRC really means "important to safety".

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WAPWR 5.1-3 AMENDMENT 3 ,

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s o ' Standard Review Plan 3.9.3 NRC acceptance criteria are provided for:

I (A) Structural analysis and systems evaluation (interaction of snubbers with the systems and components to which they are attached).

I (B) Characterization of mechanical properties (spring rates used

]

in analytical models).

(C) Design specifications.

(D) Installation and operability verification.

(E) Use of additional snubbers as a result ,of unanticipated piping vibration or interference problems during construction.

(F) Inspection and testing.

h (G) Classification and identification (safety analysis report documentation).

o Draft Regulatory Guide, Task SC-708-4.

(A) Functional specifications (in accordance with Appendix A of the guide) should be prepared for each snubber model and should be used as the basis to determine the acceptability test results.

-f gi (B) Snubbers should be constructed according to Subsection NF of Section III of the ASME Code.

g

( WAPWR AMENDMENT 3 l B788e:1d 5.1-4 AUGUST 1989

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! (C) Naterials used that are exempted from Subsection NF should be compatible with other materials of construction and the working environment.

(D) Snubbers should be qualified (in accordance with Appendix B of the guide).

(E) A completed snubber unit should be accepted from the produc-tion line only if it has successfully passed all the testing described in Appendix C of the guide.

(F) The quality assurance requirements of Appendix B to 10CFR Part 50 apply.

SP/90 Response Westinghouse will completely document and justify any deviations from the above mentioned NRC snubber acceptance criteria during the FDA application 3

for the SP/90 design.

14. Issue A-14: Flaw Detection 3

This issue was dropped and is no longer being reviewed.

15. Issue A-15: Primary Coolant System Decontamination and Steam Generator Chemical Cleaning Discussion Operation of a light water reactor results in slow corrosion of the inter-ior metal surfaces of the primary coolant system. The resulting corrosion products circulate through the reactor core and are activated by neutron flux from the fissioning reactor fuel. While some of these activated cor-rosion prodcuts are removed by the reactor's water chemistry system, a sull amount is continually deposited or plated out on the primary coolant WAPWR 5.1-5 AMENDMENT 3 B788e:1d AUGUST 1989

system's internal surfaces. Once activated corrosion products are depos-ited or plated out, they are not removed by the reactor water cleanup system and continue to accumulate.

h The presence of this accumulation of highly radioactive corrosion products adhering to the interior surfaces of the primary coolant system has, in some cases, prevented licensees from carrying out some of the less impor-h tant inservice inspections required by their technical specifications.

Because of the safety significance of the systems and components being inspected, the NRC believes an approach should be developed to permit these inspections while at the same time minimizing personnel radiation exposures. Several methods of decontamination to reduce radioactivity levels in the primary system are available to the nuclear industry for application in operating reactors. These include chemical decontamina-tion, electropolishing, mechanical and hydraulic decontamination. For example, NUREG/CR-1915. " Decontamination Processes for Restorative Opera-tions and as a Precursor to Decommissioning: A Literature Review," and similar documents are intended to give sufficient information to allow reasonable selections for decontamination processes for any given reactor.

This generic task involves an NRC review of existing an ongoing decontam-ination technology with the purpose of providing guidance to the NRC staff and industry relating to acceptable methods of decontamination of reactor primary coolant systems.

The NRC considers this issue resolved'with the issuance of NUREG/CR-2963,

" Planning Guidance for Nuclear Power Plant Decontamination Operations."

This NUREG provides generic guidance for planning, implementing, and mon-itoring restorative decontamination.

SP/90 Response One of the design objectives for the SP/90 is to minimize exposures to individuals associated with operation and maintenance through such methods as material selection, chemistry control, plant layout, high purification capability, plating of manways and other sealing surfaces, etc.

I WAPWR 5.1-6 AMENDMENT 3 l 378Be:Id AUGUST 1989

l 1

Westinghouse recommends that should decontamination of systems or compon-v ents be necessary for whatever reason, methods acceptable to the NRC and compatible with the particular system or component being decontaminated should be utilized by a SP/90 licensee.

16. Issue A-16: Steam Effects on BWR Core Spray Distribution This issue is not applicable to Westinghouse pressurized water reactor l j

designs.

17. Issue A-17: Systems Interactions in Nuclear Power Plants This issue is identified as an Unresolved Safety Issue. Refer to Section 4.0 for a discussion of Unresolved Safety issues.

18. Issue A-18: Pipe Rupture Design Criteria 3

This issue was dropped and is no longer under review.

O 19. Issue A-19: Digital Computer Protection Systems i 3

This task is not directed toward affecting the level of safety, but toward improving the efficiency of NRC licensing reviews and, therefore, has no hardware impact on the SP/90 design.

20. Issue A-20: Impacts of the Coal Fuel Cycle i

f

[3 This task is associated with an environmental proceedings issue that is  !

not applicable to Westinghouse in relation to the SP/90 design. l l

l 1

O l i

1 O WAPWR AMENDMENT 3 B788e:1d 5.1-7 AUGUST 1989 I

21. Issue A-21: Main Steam Line Break Inside Containment Evaluation of Environmental Conditions for Equipment Qualification Discussion Safety related equipment inside containment of a nuclear power plant is qualified for the most severe accident conditions under which it is expected to function. In a pressurized water reactor, this has for older generation plants been assumed to be the pressure and temperature that would accompany a loss-of-coolant accident resulting from the failure of g the largest pipe in the reactor primary system. However, for most plant W designs, calculations indicate that the failure of a main steam line inside containment results in a temperature that is higher than the tem-perature calculated for a loss-of-coolant accident and, therefore, possib-ly higher than the temperature for which the safety-related equipment is qualified. The purpose of this task is for the NRC to recommend accept-able methods of calculating environmental conditions that would result from a steam line failure within the containment for the purpose of quali-fying safety-related equipment.

Although calculations indicated that the temperature within the contain-mont following a steam line break are significantly higher than that foi-lowing a loss-of-coolant accident, the duration of the high temperature was calculated to be short. Because of the relatively low heat transfer rate in superheated steam and the heat capacity of the affected safety-related equipment, the equipment itself would not be expected to exceed the temperature for which it was qualified as a result of this short dura-tion peak in the temperature of the containment atmosphere. Therefore, the NRC believes that although this task may result in an improved basis for determining the environmental conditions for equipment qualification, g

it does not involve a major reduction in the degree of protection to the health and safety of the public.

O WAPWR AMENDMENT 3 B788e:1d 5.1-8 AUGUST 1989

SP/90 Response This issue and its ultimate resolution is really aimed at operating plant licensees that did not qualify safety-related equipment to steam line

(~ break conditions. Current NRC criteria imposed on recent plant designs

\ requires that all safety-related mechanical and electrical equipment shall be shown capable of performing their design safety functions under all normal, abnormal, accident, and post-accident environments. This criteria w includes a postulated steam line break inside containment and the SP/90 x design will meet this criteria. .

Current Westinghouse generic . environmental qualification programs are in accordance with this NRC criteria and the SP/90 design is not expected to be impacted by this issue. Westinghouse will demonstrate that the SP/90 design is enveloped by the generic qualification programs as discussed in Section 4.0, item 14 (Unresolved Safety Issue A-24, " Qualification of  ;

Class 1E Safety-Related Equipment"). The appropriate environment for equipment qualification will be determined as part of the mass and energy /

containment response analysis to be done as part of the normal design process.

,22. Issue A-22: PWR Main Steam Line Break -

Core, Reactor Vessel, and Containment Building Response This issue has been dropped and is no longer under review. 3 1

23. Issue A-23: Containment Leak Testing l O This issue has been classified as a regulatory impact issue and does not 3

require licensee resolution.

i

24. Issue A-24: Qualification of Class 1E Safety-Related Equipment This issue is identified as an Unresolved Safety Issue. Refer to j Section 4.0 for a discussion of Unresolved Safety Issues.

WAPWR AMENDMENT 3 B788e:1d 5.1-9 AUGUST 1989

s

25. Issue A-25: Non-Safety Loads on Class 1E Power Sources

~

Class 1E power sources are part of the onsite emergency power system and provide the electric power for the equipment and systems that are essen-tial to emergency reactor shutdown, containment isolation, reactor core cooling, containment and reactor heat removal or are otherwise essential in preventing a signficant release of radioactive material to the environ-ment. Past regulatory practice has allowed the connection of non safety loads in addition to the required safety loads to Class 1E power sources by imposing some restrictions. The purpose of this task is for the NRC to determine whether or not the reliability of the Class 1E power sources is l significantly afffected by the sharing of ssfety and non-safety loads.

The NRC considers this issue as technically resolved with the issuance of Revision 2 to' Regulatory Guide 1.75, " Physical Independence of Electric Systems." This regulatory guide basically endorses IEEE Standard 384-1974, "IEEE Trial-Use Standard Criteria for Separation of Class 1E Equipment and Circuits", (also designated ANSI N41.14), and still permits Class 1E power sources to share safety and non-safety loads with certain restrictions.

h A specific NRC concern related to this issue is discussed in Section 6.5 (item 18).

SP/90 Response Westinghouse will completely document and justify any deviations from the NRC Regulatory Guide 1.75 (and IEEE Standard 384-1974) positions during 3 the FDA application for the SP/90 design.

g

26. Issue A-26: Reactor Vessel Pressure Transient Protection This issue is identified as an Unresolved Safety Issue.

Section 4.0 for a discussion of Unresolved Safety Issues.

Refer to g WAPWR AMENDMENT 3 E788e:1d 5.1-10 AUGUST 1989

27. Issue A-27: Reload Applications This issue and its resolution deals with informational requirements neces-sary for the NRC reviews of reload applications and has no impact on the 3

.p SP/90 design.

28. Issue A-28: Increase in Spent Fuel Pool Storage Capacity With the present "no-reprocessing" posture throughout the nuclear power i O industry, a considerable increase in onsite spent fuel storage will be required in order to permit continued operation of many nuclear power plants. The NRC considers this issue resolved-with the issuance of a letter (dated April 14,1978) from B. Grimes (USNRC) to all power reactor licensees, "0T Position for Review and Acceptance of Spent Fuel Storage and Handling Applications."

SP/90 Response

. As stated above, the generic task actually deals with expanding the spent fuel storage capability at existing operating plants. In itself this issue does not impact the SP/90 design. However, due to the lack of sufficient "away-from-reactor" spent fuel storage capability, the SP/90 design will include spent fuel storage capability for approximately 5 cores. Possibly through the use of high density racks or spent fuel con-solidation, this capability may be increased to approximately 7 to 10 Cores.

29. Issue A-29: Nuclear Power Plant Design for the Reduction of Vulnerability to Industrial Sabotage Discussion The safety concern of this item deals with the consideration of alternatives to the basic design of nuclear power plants with the emphasis primarily on 3 reduction of the vulnerability of reactors to industrial sabotage.

O WAPWR AMENDMENT 3 1

E788e:1d 5.1-11 AUGUST 1989 i

s Extensive efforts and resources are expended in designing nuclear power plants to minimize .the risk to the public. health and safety from equipment or system  ;

malfunction or failure. However, reduction of the vulnerability of reactors to l industrial sabotage is currently treated as a plant physical security function and not as a plant design requirement. Although present reactor designs do  !

provide a' great deal of inherent protection against industrial sabotage, I extensive physical security measures are still required to provide an acceptable level of protection. An alternate approach would be to more fully consider reactor vulnerabilities to sabotage along with economy, operability, reliability, maintainability, and safety during the preliminary design phase.

Since emphasis is being placed on standardizing plants, it is especially important to consider measures which could reduce the vulnerability of reactors to sabotage. Of course, any design features to enhance physical protection must be consistent with present and future system safety requirements.

The design change assumed by NRC for the purpose of analyzing this safety issue is the addition of an independent hardened decay heat removal system which is designed to be only used in a sabotage incident or other extreme emergency as determined by plant operators. This proposed design change is based on considerations and recommendations in a Sandia report NUREG/CR-1345, g

" Nuclear Power Plant Design Concepts for Sabotage Protection completed for the NRC. Several other design changes were considered in the report.

The independent hardened decay heat removal system is assumed to be added only 3

to new PWRs and BWRs based on information in the Sandia Report.

SP/90 Response Several SP/90 plant features aid in reducing potential for sabotage by plant operating personnel. These include:

o Access to the huclear Power Block which contains all but one of the vital plant areas (the exception being the intake structure for the service water system) is from the service building only via two single WAPWR AMENDMENT 3 5788e:Id 5.1-12 AUGUST 1989

U

{

s doors, one to the " clean" and' one to the ' dirty" (or. radiation controlled) area. This type of arrangement allows close monitoring of personnel _ entering or leaving the Nuclear Power Block.

o _W ithin the Nuclear Power Block, the general approach .has been to design vital systems with redundant, non-interconnected trains,-and to locate these redundant trains into dedicated and separated safety areas. (It should be pointed out that this general approach was not only dictated by sabotage considerations; other .eventsL which could lead to common mode failure such as fire and flooding were of equal concern.) Within each vital area, equipment, piping, valves, etc.,

are located in individual rooms, which would normally be locked.

Access to these rooms would only be required occasionally (e.g.,

periodic testing, inspection maintenance) and can be strictly controlled.

These access-control and system design features complicate the task of a 3 would-be. saboteur; additional protection against sabotage is provided by redundancy and diversity inherent in the decay heat removal systems. Three totally separate systems are able to perform the decay heet removal function:

o The ~startup feedwater system is a control grade system consisting of a single pump taking suction from either the deaerating heater or the condensate storage tank and delivery to all four steam generators; autcoatic SG 1evel control is included. The single pump is not connected to the on-site emergency power supply.

o The emerge'ncy feedwater system is a dedicated safety system; it O consists of two non-interconnected subsystems, each containing one motor-driven pump, one turbine-driven pump, and one storage tank. The j motor-driven pumps are connected to the on-site emergency power supply. The turbine-driven pumps do not require any support system.

O Any one of the four pumps provides sufficient flow to remove the core i decay heat.

O WAPWR AMENDMENT 3 5788e:1d 5.1-13 AUGUST 1989

I o The integrated safeguards system, in conjunction with the pressurizer l power. operated relief valves (PORVs), allows feed and bleed operation; two of the four high head safety injection (HHSI) pumps are sufficient l in this case. The HHS! pumps are, of course, connected to the on-site emergency power supply. j The dependence of the above systems on the various support systems is provided in the Table A-29 at the end of this section.

With regard to layout, the above systems are located in separated areas of the plant, thereby further complicating the task of a potential saboteur. The l

startup feedwater system is primarily located in the turbine building. The '

two subsystems of the emergency feedwater system are located in the two dedicated ssfety areas in the clean part of the auxiliary building. The integrated safeguards system is located in the " dirty" part of the auxiliary ]

building with strict separation between the four subsystems. Finally, the 3

redundant trains of supporting systems are located in the two separated safety l areas in the clean area of the auxiliary building and are therefore protected l to the same degree as the primary mitigation systems.

In summary, the SP/90 decay heat removal systems incorporate substantial redundancy and diversity, both in terms of system design and plant layout; it

.would appear to be an extremely difficult task for an inside saboteur to disable all these systems and components without being detected.

The SP/90 basic system and layout design features, combined with security procedures and personnel screening measures to be developed by the utility, i provide the necessary protection against industrial sabotage; the concerns of Generic Safety Issue A-29 are, thus, adequately addressed.

l l

l WAPWR AMENDMENT 3 O I B788e:1d 5.1-14 AUGUST 1989 .

- _ - - _ _ _ _ _ _ _ _ = . ____

TABLE A-29 -- DECAY HEAT REMOVAL SUPPORT SYSTEMS DEPENDENCE Oc AC Power DC Power Cooling Water Off-Site On-Site Non-1E 1E Turbine Nuclear O Startup Feedwater X NA(1) X -

X -

Energency Feedwater o Notor Driven Pumps -

X -

X -

X o Turbine Driven Pumps - - - - - -

Bleed and Feed -

X -

X --

X 3

O I

t

(

(1) [Startup feedwater pump) is not connected to on-site power supplies.

WAPWR AMENDMENT 3 B788e:1d 5.1-15 AUGUST 1989 i I

30. Issue A-30: Adequacy of Safety-Related DC Power Supplies ,

3 This issue was subsumed in New Generic Issue 128 O

l l

31. Issue A-31: Residual Heat Removal Requirements This issue is identified as an Unresolved Safety Issue. Refer to l Section 4.0 for a discussion of Unresolved Safety Issues.

32. Issue A-32: Missile Effects g

This issue addresses three types of missiles for which impact effects on l

nuclear power plant structures, systems, and components important to safety must be evaluated. These missiles are also addressed in Issue A-38, " Tornado Nissiles;" Issue A-37, " Turbine Nissiles;" and for the most energetic accident-induced missile, Issue B-68, "PWR Pump Overspeed During LOCA." Refer to the discussion of these issues.

33. Issue A-33: NEPA Reviews of Accident Risks This issue and its resolution is associated with evaluating accidents in the context of environmental reviews of nuclear power plants and 3

. accordingly, would be addressed by each utility using the SP/90 design as part of its environmental impact statement, and as such it is not applicable to Westinghouse in relation to the SP/90 design.

34. Issue A-34: Instruments for Nonitoring Radiation and Process Variables During Accidents Discussion O

The purpose of this task was for the NRC to develop criteria and guidelines to be used by applicants, licensees, and NRC staff reviewers to WAPWR B788e:1d 5.1-16 AMENDMENT 3 AUGUST 1989 GI{

j i

l

h ,9 support' implementation of Regulatory Guide 1.97, Revision 1, " Instruments-

' tion for Light-Water-Cooled Nuclear Power Plants-- to Assess Plant Condi-tions During and Following an Accident."

The NRC considers this issue as technically resolved with the issuance of Revision 2 to Regulatory Guide 1.97.

SP/90 Response Regulatory Guide 1.97, Revision 2, .in relation to the SP/90 design is dis-3 cussed in Section 3.1 (item 23) 10CFR50.34(f)(2)(xix) - II.F.3.

35. Issue A-35: -Adequacy of Offsite Power Systems L Discussion The NRC requires that electric power for safety systems be comprised of two redundant and independent divisions, each capable of providing the O necessary plant protection functions during all normal operating con-ditions and following various design basis accidents. 'Each division includes an offsite AC power connection (the preferred power source), a.

standby emergency diesel generator AC power supply (capable of powering-essential safety systems should the offsite source be lost). and DC power sources.

Events at several plants involving the loss or degradation of the offsite power system or involving its connection to the emergency onsite power l

O- system have indicated that a reassessment of current NRC requirements was appropriate. This task was undertaken by the NRC to perform such an assessment and to determine the need, if any, for upgrading the offsite power. sources- and/or their interfaces with the onsite power system at nuclear power stations. The issue in relation to the Millstone Unit 2 event is discussed in Section 6.5 (item 20).

O l WAPWR 5.1-17 ANENDNENT 3

.E788e:Id AUGUST 1989

i s

The NRC considers this issue as technically resolved with the issuance of the Standard Review Plan 8.3.1, "A-C Power Systems (Onsite)," acceptance criteria.

g' l l SP/90 Response i Westinghouse will completely document and justify any deviations from the I NRC Standard Review Plan 8.3.1 acceptance criteria during the FDA 3

a lication for the SP/90 design.

36. Issue A-36: Control of Heavy Loads Near Spent Fuel This issue is identified as an Unresolved Safety Issue. Refer to Section 4.0 for a discussion of Unresolved Safety Issues.

37. Issue A-37: Turbine Missiles 3 This issue has been dropped and is no longer under review.

38. Issue A-38: Tornado Missiles h

~

Discussion General Design Criteria 2 and 4 of 10CFR Part 50, Appendix A, require in part that structures, systems, and components important to safety be designed to be able to withstand the effects of tornado missiles. A mis-sile generated by a tornado may be energetic enough to cause damage to improperly protected systems or components. This damage may ultimately result in the release of radioactivity to the environment. This design requirement imposed new demands on the practice of structural engineering, h

that is, for other types of facilities, tornadoes have always been con-sidered too rare an event to be included in the design basis. Consequent-ly, no body of design practice existed and design criteria for tornado resistance had to be developed. The first NRC requirements were published g

WAPWR AMENDMENT 3 5788e:1d 5.1-18 AUGUST 1989 I

s

) in Standard Review Plan 3.5.1.4, " Missiles Generated by Natural Phenom-ana," in 1975 and revised in 1976.

Since 1976 Standard Review Plans 3.3.2, " Tornado Loadings," and 3.5.1.4,

" Missiles Generated by Natural Phenomena," have been revised and Regula-tory Guides 1.76, " Design Basis Tornado for Nuclear Power Plants," and 1.117 " Tornado Design Classification," have been issued and/or revised.

In these documents the NRC details specific design acceptance criteria to meet the requirements of General Design Criteria 2 and 4 and recommends methods of satisfying the acceptance criteria.

The purpose of this task is not for the NRC to investigate new possibili-ties to increase plant safsty but to refine the spectrum of possible tor-nado missiles. The NRC's judgment was that postulated missile velocities, size, and orientation used in the plant safety analysis are more conserva-tive than tornado damage histories would warrant.

c The end product of this generic issue was to be a set of design basis mis-( siles that does not impose unnecessary design requirements on plant con-struction and for which a sound technical basis exists.

. SP/90 Response Current NRC regulations and regulatory guidance have been utilized in the i 3 SP/90 design in relation to tornado missiles. Westinghouse will complete-ly document and justify any deviations from the NRC acceptance criteria 3

during the FDA application for the SP/90 design.

b 39. Issue A-39: Determination of Safety Relief Valve Pool Dynamic Loads and Temperature Limits for BWR Containments This issue is not applicable to Westinghouse pressurized water reactor designs.

O WAPWR 5.1-19 AMENDMENT 3 B788e:1d AUGUST 1989 l

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40. Issue A-40: Seismic Design Criteria - Short Term Program This issue is identified as an Unresolved Safety Issue. Refer to O Section 4.0 for a discussion of Unresolved Safety Issues.

41. Issue A-41: Seismic De' sign Criteria - Long Term Program Discussion This issue involves long term research programs on seismic design. In g this rergard, the NRC has established a Seismic Safety Margins Research w Program which is basically intended to quantify how much seismic acrgin is available for various components in current operating plant designs. This quantification is intended to be used to develop probability models that j could assess the impact of seismic events much larger than the current j' safe shutdown earthquake design basis.

SP/90 Response The Westinghouse practice of generic seismic level qualification has, in general, resulted in additional seismic safety margins in Westinghouse h

equipment. Westinghouse anticipates no hardware impact on the SP/90 design as a result of this' issue.

42. Issue A-42: Pipe Cracks in Boiling Wcter Reactors This issue is not applicable to Westinghouse pressurized water reactor designs.

43. Issue A-43: Containment Emergency Sump Performance This issue is identified as an Unresolved Safety Issue. Refer to Section 4.0 for a discussion of Unresolved Safety Issues. g l

WAPWR AMENDMENT 3 B788e:1d 5.1-20 AUGUST 1989 l

I

44. Issue A-44: Station Blackout This issue is identified as an Unresolved Safety Issue. Refer to Section 4.0 for a discussion of Unresolved Safety Issues.

45. Issue A-45: Shutdown Decay Heat Removal Requirements This issue is identified as an Unresolved Safety Issue. Refer to

'Section 4.0 for a discussion of Unresolved Safety Issues.

46. Issue A-46: Seismic Qualification of Equipment in Operating Plants This issue is identified as an Unresolved Safety Issue. Refer to Section 4.0 for a discussion of Unresolved Safety Issues.

47. Issue A-47: Safety Implications of Control Systems This issue is irientified as an Unresolved Safety Issue. Refer to Section 4.0 for a discussion of Unresolved Safety Issues.

48. Issue A-48: Hydrogen Control Measures and Effects of Hydrogen Burns on Safety Equipment This issue is identified as an Unresolved Safety Issue. Refer to Section 4.0 for a discussion of Unresolved Safety Issues.

49. Issue A-49: Pressurized Thermal Shock O This issue is identified as an Unresolved Safety Issue. Refer to Section 4.0 for a discussion of Unresolved Safety Issues.

O WAPWR AMENDMENT 3 5788e:1d 5.1-21 AUGUST 1989

3 5.2~ CATEGORY B' ISSUES The following discussions: pertain to current Category B issues in relation to the SP/90 design. NRC discussions and descriptions of these issues are con-tained in NUREG-0471,"GenericTaskProblemDescriptions(CategoryB,C,andD O.

Tasks)"andNUREG-0933.-"APrioritizationofGenericSafetyIssues."

1. Issue B-1: Environmental Technical Specifications Discussion Current NRC regulations and practice require that certain ' operating requirements, Technical Specifications, be made part of each operating license. The nonradiological portion of Appendix B to the operating license . traditionally derives from information in the Final Environmental Statement and other relevant sources. Based on several years cf NRC experience with facility licensing and a' better_ understanding. of Environmental Protection Agency and NRC responsibilities in the' area of water quality regulation, it is believed that the development of Standardized Environmental Technical Specifications (SETS) is appropriate. SETS are intended to result in more efficient use of NRC 4*J applicant resources _ and more uniform requirements and performance standards for licensees.- The NRC intends that this task results in the development of Standardized Environmental Technical Specifications to be published as a NUREG report or as part of Regulatory Guide 4.8,

" Environmental Technical Specifications for Nuclear Power Plants." SETS are being prepared on a case-by-case basis. The NRC considers this issue resolved.

SP/90 Response This issue and its resolution is associated with site specific environmental technical specification guidance and accordingly, it is not applicable to Westinghouse in relation to the SP/90 design.

Environmental technical specifications are the responsibility of each utility utilizing the SP/90 design.

I WAPWR-RC 5.2-1 AMENDMENT 3 j 8830s:1d AUGUST 1989 l

___ _ _ _ _ - - - - . -. I

s l 2. Issue B-2: Forecasting Electricity Demand l l

Discussion 0i

)

Originally, this issue was directed at improving the NRC's capability to forecast electricity demand for the purpose of evaluating an applicant's need for power forecasts in individual licensing cases. i l

As discussed in some detail in Section 5.1 (item 20), the NRC has l recently revised their regulations to no longer require that the issue of ,

"need for power" be addrassed in operating license proceedings.

1

(

As a matter of policy, the Commission endorses placing substantial reliance on state assessments of need for power, energy conservation, and alternative energy source analyses to fulfill the NRC's National Environmental Policy Act responsibilities at the construction permit stage and has initiated the development of procedures for soliciting this input.

This Environmental issue has been resolved with the publication of the following documents: (A) Regulatory Guide 4.1, Rev. 2, Chapter 1 on

" Purpose of the Proposed Facility and Associated Transmission," July

. 1976; (B)NUREG/ CR-0022 on "Need for Power: Determination in the State Decision Making Process," March 1978; (C)NUREG/CR-0250 on " Regional Econometric Model for Forecasting Electricity Demand by Sector and State," September 1978; (D) Section 8 of NUREG-0555 on " Environmental Standard Review Plans for the Environmental Review of Construction Permit Applications for Nuclear Power Plants," May 1979; (E) Part III of March 1980; (F) ORNL/TM-7947 on "An Integrated System for Forecasting Electric Energy and Load for States and Utility Service Areas," May 1982; and h { '

(G) NUREG-0942 on " Conducting Need-for-Power Review for Nuclear Power i Plants: Guidelines to States," draft report of December 1982. I O\

l l

WAPWR-RC 5.2-2 AMENDMENT 3 I E83De:Id AUGUST 1989

f )

I SP/90 Response: .

L The need for ' power issue.is associated with an environmental procee' dings task that does not affect plant safety nor' impact Westinghouse in p relation-to the SP/90 design.

V '

3. Issus B-3:. Event Categorization. i This' issue.has been' dropped _ and is no longer under review. 3

4. Issue B-4: ECCS Reliability Discussion This issue has been subsumed into Item II.E.2.1 -of NUREG-0660, "NRC 3 Action Plan Developed as a Result of the TNI-2 Accident." . Refer' to ,

1 Section 3.3.2-(item 9) for a discussion of Item II.E.2.1 of NUREG-0660.

5. , Issue B-5: Ductility of. Two-Way Slabs and Shells and Buckling Behavior of Steel Containments Discussion Issue- B-5 Part II involves concern over the lack of a uniform, well-defined approach for design evaluation of steel containments.

If steel containment shells were to fail due to loading which my cause buckling, one of the plants' levels of defense would be lost and could O result in release of radioactivity to the environment. The loading would 3 L

have to be due to a high-energy source. A large LOCA or HELB near the L

containment wall could possibly provide such a load.

The- structural design of a steel containment vessel subjected to unsymmetrical dynamic loadings may be governed by the instability of the shell. For this type of loadirag, the current design verification O ' WAPWR-RC 5.2-3 AMENDMENT 3 E830s:1d- AUGUST 1989

4 methods, analytical techniques, and the acceptance criteria may not be as comprehensive as they should be.- Section III of the ASME Code does not provide detailed guidance on the treatment of buckling of steel g

containment vessels for such loading conditions. Moreover, this Code does not address the asymmetrical nature of the containment shell due to the presence of equipment hatch openings and other penetrations.

Regulatory Guide 1.57, " Design Limits and Loading Combinations for Metal Primary Reactor Containment System Components," recommends a minimum factor of safety of 2.0 against buckling for the worst loading condition provided a detailed rigorous analysis, considering inelastic behavior, is performed. On the other hand, the 1977 Summer Addenda of the ASME Code permits three alternate methods, but requires a factor of safety between  ;

2.0 and 3.0 against buckling depending upon the applicable service limits.

RES efforts in resolving this issue resulted in a proposed revision to SRP Section 3.8.2 that would be applicable to CP and OL applications filed after the effective date of the SRP Section revision. Operating 3

plants were not affected by the proposed SRP revision because there was a general staff consensus that existing steel containments had adequate design conservatism regarding buckling. The proposed SRP revision was a formal promulgation of the changed staff review practices since the first SRP had been published in 1975 and added guidance for the review of asymmetric containment designs. It included an interim set of criteria for evaluating steel containment buckling that had been developed several years earlier by the former Structural Engineering Branch of NRR and had been applied to plants undergoing operating license reviews.

During the review process, the Structural and Geosciences Branch of NRR identified two concerns with the proposed resolution: (1) although the proposed revision to SRP Section 3.8.2 reflected NRR practice on the most recent licensing reviews, NRR expressed the concern that it contained techni:a1 requirements that were overly conservative; and (2) there was a general consensus that existing plants with steel shell containments had acceptable margins regarding buckling. However, there was no readily g j available documentation to show this. I WAPWR-RC 5.2-4 AMENDMENT 3 .

B830e:1d AUGUST 1989 l

^ In March 1988, the Structural .and Geosciences Branch of NRR issued a memorandum that: (1)summarizedNRR'sconcernwiththeproposedrevision to'the SRP Section; (2) provided an evaluation that concluded that existing steel containments had adequate margins against buckling; and 3

(3) stated that it was NRR's judgment that the issue of steel containment O- buckling had very little safety impact and was not worth pursuing i

further, considering the staff resource constraints. Thus, the issue was RESOLVED and no new requirements were established.

SP/90 Response The interim criteria discussed above will be considered during the SP/90 containment design. In' addition, Westinghouse will completely document and justify any deviations from the NRC Standard Review Plan 6.2.1 acceptance criteria during the licensing process for the SP/90 design.

6. Issue B-6: Loads, Load Combinations, Stress Limits

. Discussion The designer of pressure vessels and piping system components must consider (A) the individual and combined loads that will act on each component due to normal operating conditions, system transients, and postulated low probability events (accidents and natural phenomena), and (B) the stress limits to be used in evaluating structural integrity and component operability when subject to these loads.

The work effort to investigate and establish a position on dynamic response combination methodology was completed and reported in NUREG-0484, Rev. 1, " Methodology for Combining Dynamic Responses". The conclusions in thic report have been incorporated into the latest version of SRP 3.9.3. Ln additt6n, work has been completed on an evaluation of the loads and load combinations for containment structures. The only work remaining is research into decoupling the LOCA and SSE events.

Reports on two investigations addressing this issue have been released as O WAPWR-RC 5.2-5 AMENDMENT 3 5830e:Id AUGUST 1989

s NUREG/CR-2136, " Effects of Postulated Event Devices on Normal Operation of Piping Systems in Nuclear Power Plants," and NUREG/CR-2189,

" Probability of Pipe Fracture in the Primary Coolant Loop of a PWR Plant."

g The purpose of this task is for the NRC to provide guidance on load combination methods and acceptable stress limits.

This issue has been subsumed into New Generic Safety Issue 119.1, which 3 has been identified as a regulatory impact issue. Therefore, no action is required for the SP/90 design.

7. Issue B-7: Secondary Accident Consequence Modeling 3 This issue has been dropped and is no longer under review.

8. Issue B-8: Locking Out of ECCS Power-Operated Valves 3 This issue has been dropped and is no longer under review.

9. Issue B-9: Electrical Cable Penetrations of Containment h D,iscussion The purpose of this task was for the NRC to reevaluate current licensing criteria for the design and qualification testing of electrical penetrations in the reactor containment in light of concerns raised by these failures. Some prototype electrical penetration failures occurred in both high- and low-voltage penetration modules at licensed facilities. It was originally postulated that the failures of the low-voltage penetration modules were due to electrical short circuits caused by collection of moisture in fissures (cracks) in the epoxy insulator-sealant. However, results of the laboratory analysis indicated that the failures were caused by heating of the conductors at the connection splices within the penetration module. The heating resulted from high contact resistance due to epoxy intrusion into an area of l

WAPWR-RC 5.2-6 AMENDMENT 3 5830e:Id AUGUST 1989

s connector splices that were not' insulated during the manufacturing Os process. . .The accumulation of carbon deposits over a period of time, resulting from the heating' process,- created a conductive path (short circuit) between adjacent conductors in the penetration modules.

Existing. requirements in IEEE Standard 317-1976, " Electrical Penetration Assemblies in Containment Structures for Nuclear Power Generating Stations," and Regulatory Guide 1.63, " Electric Penetration Assemblies in Containment Structures for Light-Water Cooled Nuclear Power Plants" provide adequate direction for the design of containment electrical cable penetrations. Thus, the NRC considers this issue technically resolved.

SP/90 Response l Westinghouse will completely document and justify any deviations from the NRC positions of Regulatory Guide 1.63 (which endorses IEEE Standard 317-1976) during the Final Design Approval licensing process for the 3 SP/90 design. .

' O 10. Issue B-10: Behavior of BWR Nark III Containment This issue is not applicable to Westinghouse pressurized water reactor designs.

11. Issue B-11: Subcompartment Standard Problems Discussion )

The calculations of differential pressures that occur in containment subcompartments from a loss-of-coolant event require a complex fluid dynamic analysis to assure that the rubcompartment design pressures are not exceeded. To check the various industry computer codes used for the analyses, the NRC has it. sued a standard problem to the reactor vendors and architect engineers so that their models and calculational methods can be evaluated. This task, now complete, involved the NRC review and ,

O WAPWR-RC 5.2-7 AMENDMENT 3 l

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)

evaluation of the subcompartment standard problem analyses supplied by l

vendors and architect engineers to determine the validity of their j g

models. Standard Review Plan 6.2.1.2, "Subcompartment Analysis," W provides current NRC acceptance criteria for containment subcompartment analyses.

This issue has been identified as an NRC Licensing issue and requires no 3

action on the part of the applicant. -

12. Issue B-12: ContainmentCoolingRequirements(Non-LOCA)

Discussion O

The rationale for normal and postaccident containment cooling has been reviewed by the NRC to determine the adequacy of the design requirements imposed on the containment ventilation systems. By reviewing typical designs the NRC developed a basic understanding of the consequences of a loss of normal containment cooling, including the impact, if any, on the operability of safety systems and control systems. Specifically, the purpose of this task was to establish whether or not (A) the normal l ventilation system is essential to achieve a safe cold shutdown, (B)a failure in the system could cause an accident, and (C) the system is N I required to mitigate accidents.

The NRC considers this issue as being technically resolved and their urrent acceptance criteria ar'e documented in Standard Review Plan 6.2.2,

" Containment Heat Removal Systems."

13. Issue B-13: Marviken Test Data Evaluations Discussion Test data from the Marviken containment tests have been obtained by the g NRC for the purpose of validating containment pressure codes currently w used for performing independent calculations related to licensing WAPWR-RC 5.2-8 AMENDMENT 3 5830e:1d AUGUST 1989

s. i

reviews. The Marviken data are. containment pressure responses from a f full-scale blowdown using a pressure suppression type containment. This task, now complete, correlated the Marviken data and compared the results with existing computer programs.-

The NRC considers this licensing issue as being technically resolved and Standard Review Plan 6.2.1.1.A. "PWR Dry Containments, Including i Subatmospheric Containments," provides- acceptance criteria for. the containment response (e.g., pressure and temperature) as a result of. a postulated loss-of-coolant- accident and secondary system steam and feedwater line breaks.

3

14. . Issue B-14:. Study of Hydrogen Mixing Capability in Containment Post-LOCA This issue was .susumed into Unresolved ' Safety Issue A-48, " Hydrogen l 3 Control Measures and Effects of- Hydrogen Burns on Safety Equipment" (refertoSection4.0, item 26).

15. . Issue B-15: CONTEMPT Computer Code Maintenance This issue has been dropped and is no longer under review. 3

16. Issue B-16: Protection Against Postulated Piping Failures in Fluid Systems Outside Containment This issue was subsumed into Issue A-18, " Pipe Rupture Design Criteria" 3 (refer to Section 5.1, item 18).

O 17. Issue B-17: Criteria for Safety-Related Operator Actions Discussion O Current plant designs are such that reliance on the operator to take action in response to certain transients is necessary. In addition, some current pressurized water reactor designs require manual operations to O 5.2-9 AMENDMENT 3 WAPWR-RC 8830e:1d AUGUST 1989 o_ _ __--__ ____-----_ _-----.____ - - - a

accomplish the switchover from the injection mode to the recirculation mode following a loss-of-coolant accident. The realignment operations must be accomplished before the inventory in the refueling water storage tank is depleted. ,

Issus' B-17 involves the development of a time criterion for safety-related operator actions including a determination of whether or

, not automatic actuation will be required. The evaluation of this issue  !

includes Issue 27, Manual vs. Automated Actions.

Development and implementation of criteria for safety related operator e1 actions would result in the automation of some actions currently performed by operators. The use of automated redundant safety grade controls in lieu of operator actions is expected to reduce the frequency of improper action during the response to or rec'overy from transients and accidents by removing the potential for operator error. This in turn could reduce the expected frequency of core damaging events and, therefore, reduce the public risk accordingly.

3 SP/90 Response 91 Specifically in relation to the development of the SP/90, Westinghouse has established (as a design objective) a 30-minute time period before the operator is assumed to take any safety-related action to mitigate the consequences of most design basis events.

Included in the MMI design process discussed in RESAR-SP/90 PDA Module 15, " Control Room / Human Factors Engineering," is a step titled g Man-Machine Allocation. During this step, as discussed in the module, W the MMI designs examine and match the attributes of humans and of  !

automata against the tasks (cognitive and physical) that are required for plant operation (safety as well as control) and which are identified as s, part of the Cognitive Task Analysis perfo med to initiate the SP/90 Centrol Room design process. The result is that during this step, any NRC requirement relative to Safety-Related Operator Actions will be l

WAPWR-RC 5.2-10 AMENDMENT 3 5830e:1d AUGUST 1989 1

l 1

o addressed or if nou exist, Westinghouse will use its own judgment and

+

experience and will consult the utility industry relative to relating the attributes of man and automata to the SP/90 plant design in order 'to determine the man-machine allocation.

O In ' addition, the basic design of the safeguards systems minimizes the 3

number of operator actions and automatic realignment required to perform.

desired safety functions. Also, these system designs minimize.the impact of inadvertent safety system actuation and place less stress on the operator (s) to -distinguish initiating events and/or terminate safety

system actuation.

18. Issue B-18: Vortex Suppression Requirements for Containment Sumps Discussion This issue was subsumed into Unresolved Safet> Issue A-43, " Containment .3 Emergency Sump Performance" (refer to Section 4.0, item 21).

19. Issue B-19: Thermal-Hydraulic Stability

. Discussion Demonstrating the thermal-hydraulic stability of a reactor is an essential element in the thermal-hydraulic design. Instabilities can result in fuel failures from premature departure from nucleate boiling or excessive hydraulic loads. This task involves the NRC development of the h analytical me'thods to perform independent calculations to check vendor analyses of thermal-hydraulic stability.

SP/90 Response O Westinghouse has successfully demonstrated the inherent thermal-hydraulic stability of open-channel fuel assemblies similar in configuration to the SP/90 fuel by testing and analysis.

O WAPWR-RC 5.2-11 AMENDMENT 3 5830e:1d AUGUST 1989

ll l

This issue applies to an ongoing NRC administrative activity. However, as part af the detailed FDA design and licensing process, Westinghouse j will demonstrate the thermal-hydraulic stability of the SP/90 reactor ,

I core by appropriate testing or analysis. ,

i

20. Issue B-20: Standard Problem Analysis Discussion j l

Nost vendors, in the conduct of internal audits of emergency core cooling performance computer codes, have discovered errors in coding and/or logic which have significant effects on the prediction results of approved models. This task involves the use of standard problems to evaluate the predictive accuracy of these complex computer codes and to detect errors to the extent that the errors affect the results of code predictions.

SP/90 Response Westinghouse emergency core cooling performance analyses for the SP/90 design (in accordance with 10CFR 50.46, " Acceptance Criteria Emergency Core Cooling Systems for Light Water Nuclear Power Reactors")

for h

will be performed using models approved by the NRC in accordance with Appendix K, "ECCS Evaluation Models," to 10CFR Part 50. j

21. Issue B-21: Core Physics I

3 This issue has been dropped and is no longer under review. l l

22. Issue B-22: LWR Fuel g Discussion Individual reactor fuel rods sometimes fail during normal operation, and many rods are calculated to fail during severe accidents releasing activity to the surroundings and providing a source for releases from the WAPWR-RC 5.2-12 AMENDMENT 3 B830e:1d AUGUST 1989 j 1

1 i

- - _ _ _ _ - _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _. - _ l

^

/ plant. Failures during some accidents could be severe enough to fragment .

the cladding and disperse fuel pellets into the coolant, but regulations require that the coolable rod-like geometry must be maintained.

Behavioral characteristics, such as rod bowing and densification, also p have. an effect on plant-limiting conditions. Thus, fuel behavior during

'V normal operation and postulated accidents must be predictable in order to set operating limits, to limit activity releases, and to ensure no more than acceptable degradation of the fuel system. The objective of this n task is to assure that such predictions are reliable.

U i This issue is to be prioritized in late 1989. Standard Review Plan 4.2, 3

" Fuel System Design," provides detailed NRC acceptance criteria for the design of fuel and core conponents. i SP/90 Response Westinghouse will continue to follow this issue through resolution and l 3 will completely document and justify any deviations from the NRC Standard Review Plan 4.2 acceptance criteria during the licensing process for the SP/90 design. There have been no deviations from SRP 4.2 identified for 3 the SP/90 PDA design.

23. Issue B-23: LMFBR Fuel This issue has been dropped and is no longer under review. 3

24. Issue B-24: Seismic Qualification of Electrical and Mechanical Components Discussion This issue was subsumed into Unresolved Safety Issue A-46," Seismic l3 Qualification of Equipment in Operating Plants" (refer to Section 4.0, item 24).

2 O WAPWR-RC 5.2-13 AMENDMENT 3 i

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25. Issue B-25: Piping Benchmark Problems Discussion O Applicants are requ. ired to provide confirmation of the adequacy of computer programs used in the structural analysis and design of piping systems and components. In the past this consisted of applicants i providing (and the NRC reviewing) brief descriptions of the computer programs used and solutions to simple textbook problems. In ordar to better provide assurance of the reliability of these programs, this task involved the NRC development of benchmark problems (and solutions to { '

these problems) for use in the review of applications for construction permits. l I

The results from this task were incorporated into Standard Review Plan

! 3.9.1, "Special Topics for Mechanical Components," which provides detailed acceptance criteria for demonstrating the applicability and validity of computer programs used in the structural analysis and design of piping systems and components.

This issue was determined to be a licensing impact issue and requires no O

3 action on the part of the applicant.

26. Issue B-26: Structural Integrity of Containment Penetrations Discussion Containment penetration assemblies provide a means to maintain the integrity of the containment pressure boundary and prevent overstressing of the penetration nozzle due to thermal stresses. A typical penetration h

assembly may consist of a flued head, a guard pipe, an expansion bellows and an impingement ring. The flued head may be fabricated from a forging which may be welded into the process line or may be welded to the outer surface of the process piping. This task involves a NRC evaluation to WAPWR-RC 5830e:1d 5.2-14 AMENDMENT 3 AUGUST 1989 e!'

1 4

( assess the adequacy of specific containment penetration designs from the  ;

\ point- of view of structural integrity and inservice inspection requirements.

Specifically, this NRC task involves two areas. The first (which is now considered complete) is an independently performed stress analysis by the '

NRC of the various penetrations produced as integral fittings and welded into the process line, or penetrations which are. welded to the outside circumference of the process line. The model considered the applicable

( requirements of Section III, Subsections NC and NE, of the ASME Code, NRC stress criteria, any existing fabrication residual stresses, and the mechanical loadings resulting from normal plant operation, from postulated pipe breaks, and from seismic events. The second area involves a determination that the configuration and accessibility of the welds in the proposed design and the procedures proposed for performing volumetric examination will permit the inservice examination requirements of Section XI of the ASME Code to be met.

SP/90 Response Standard Review Plans 6.2.1 through 6.2.7 specify NRC acceptance criteria for containment design. Westinghouse will completely document and justify any deviations from the NRC Standard Review Plan acceptance criteria during the licensing process for the SP/90 design, and shall 3

conform to ASME III, Subsection NF, 1986 Edition or later code of record.

27. Issue B-27: Implementation and Use of Subsection NF O Discussion Since the adoption by the ASME Code,Section III, of Subsection NF on

( component supports, NRC technical review has been limited to conformance of the information provided in the application and verification of a commitment by the applicants to component support design in accordance with the provisions in Subsection NF.

O WAPWR-RC 5.2-15 AMENDMENT 3 i

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[ l t Certain deficiencies in the use of Subsection NF, however, have been identified by the NRC. These include:

h l o The absence of definitive criteria to be used in defining the

! jurisdictional boundary between a load carrying building structure designed by AISC rules which do not contain inservice inspection requirements and an attached Subsection NF component support having inservice inspection requirements.

o As the design limits for Class 1 linear type component supports presently appear in the ASME Code, the allowable stresses exesed those permitted for other ASME Code designed components. If these limits are approached repeatedly in the component support, the support could fail by fatigue.

The NRC plans to develop a Branch Technical Position that will assess deficiencies for use by the NRC in case reviews of component supports.

This issue has been identified as a licensing impact issue and currently requires no action on the part of the applicant.

SP/90 Response The design of the SP/90 component supports (which will be documented during the licensing process for the SP/90 design) will be performed with due consideration of the above mentioned deficiencies.

28. Issue B-28i Radionuclides / Sediment Transport Program Discussion 1

As a result of Appendix I and the Liquid Pathway Generic Study I (NUREG-0440), the NRC is taking a more realistic look at the effects of sediment (surface waters) and aquifer materials (groundwater) on radionuclides transport through the hydrosphere. To accomplish this l

WAPWR-RC B830e:Id 5.2-16 AMENDMENT 3 ei '

AUGUST 1989 i i

4 m objective, itis necessary that the NRC have available for its use radionuclides / sediment transport models that have been field verified.

This task is intended to accomplish this objective through NRC radionuclides / sediment transport model development and verification.

The NRC considers this issue to be technically resolved with the issuance of NUREG/CR-2425 " Sediment and Radionuclides Transport in Rivers."

SP/90 Response This item is concerned with internal NRC radionuclides / sediment transport model development and, as such, has no impact on the SP/90 design.

29. Issue B-29: Effectiveness of Ultimate Heat Sinks Discussion This task involves the NRC confirmation of currently used mathematical i models for prediction of ultimate heat sink performance by comparing model performance with field data and development of better guidance '

regarding the criteria for weather record selection to define ultimate heat sink design basis meteorology.

The NRC considers this issue to be technically resolved with the publication of three reports. NUREG-0693, " Analysis of j Ultimate-Heat-Sink Cooling Ponds" and NUREG-0733, " Analysis of Ultimate Heat-Sink Spray Ponds," look at two sources of the ultimate-host-sink (UHS) in use today, identifying methods that may be used to select the most severe combinations of controlling meteorological parameters for cooling ponds of conventional design. NUREG-0858, " Comparison Between I Field Data and Ultimate Heat Sink Cooling-Pond and Spray-Pond Models" compares the results of the cooling pond and spray pond performances to A the NRC model predictions in the former two reports.

1 O WAPWR-RC 5.2-17 AMENDMENT 3 l 5830e:1d AUGUST 1989 '

This issue, which currently is considered a licensing impact issue, is 3

scheduled to be prioritized by late 1989.

h SP/90 Response The ultimate heat sink is plant specific and outside the scope of the SP/90 design. However, interface criteria for use by applicants in g

3 establishing their ultimate heat sink design has been provided in Subsection 9.2.5 of RESAR-SP/90 PDA Module 13 " Auxiliary Systems".

30. Issue B-30: Design Basis Floods and Probability Discussion The purpose of this task was for the NRC staff to prepere a paper for presentation to the Advisory Committee on Reactor Safeguards (ACRS) detailing the bases for design basis flood events used by the NRC staff in case reviews. Additionally, the task was to address the possible use of probability estimates for the principal flood producing events. This g ,

task has been completed and a report to the ACRS was issued in July W 1977. The report presents discussion and definitions of flood events which may be used as Design Basis Floods for review of nuclear power

, plants. It supports continued use by the NRC staff of a deterministic approach for identifying the Design Basis Flood events in preference to possible use of a probabilistic approach. The deterministic approach identifies the upper limit of flood potential physically possible. As indicated in the report, the NRC does not feel that a probabilistic approach is appropriate for use in licensing reviews at the present time because of the lack of confidence in estimates of extreme flooding events using current techniques.

h The preliminary results of the risk-based evaluation indicate that the probability of a flood-induced core meltdown accident at most sites is i

WAPWR-RC 5.2-18 AMENDMENT 3 B830e:1d AUGUST 1989

)

l very low. However, ongoing research efforts aimed at developing improved i 3 O methodological techniques for the probabilistic analysis of flooding are l being undertaken by the NRC Office of Nuclear Regulatory Research.

Standard Review Plan 2.4.2, " Floods," provides NRC acceptance criteria to 9 meet the hydraulic aspects of General Design Criterion 2, " Design Bases for Protection Against Natural Phenomena," and 10CFR Part 100, " Reactor Site Criteria." In addition, Regulatory Guide 1.29, " Seismic Design Classification," identifies the safety-related structures, systems, and G components and Regulatory Guide 1.102, " Flood Protection for Nuclear Power Plants," describes flood protection acceptable to the NRC to prevent the safety related facilities from being adversely affected.

This issue,. currently considered a licensing impact issue, is scheduled 3

for re prioritization.

SP/90 Response During the licensing process for the SP/90 design, Westinghouse will completely document and justify any deviations from the NRC regulatory positions and acceptance criteria of Regulatory Guides 1.29 and 1.102 and Star.dard Review Plan 2.4.2 for those safety-related facilities within the SP/90 scope. No deviations from SRP 2.4.2 have been identified for the 3

SP/90 PDA design.

31. Issue B-31: Dam Failure Model g Discussion During licensing reviews, the need has arisen on several occasions to have a NRC model to predict the failure discharge hydrograph due to erosional failures of earthen dams. No known model presently exists for such evaluations and, accordingly, the NRC and the applicants have been forced to conservatively postulate complete and instantaneous failure of the dam. This NRC task is intended to develop an analytical model, or O WAPWR-RC 5.2-19 AMENDMENT 3 B830e:1d AUGUST 1989

nomograph, to predict erosion rates and patterns of failure for an earthen . enhancement for a given initiating mode (e.g., overtopping, cracking).

g SP/90 Response This item deals with intarnal NRC efforts related to the development of a dam failure model and, as such, has no impact on Westinghouse in relation to the SP/90 design.

32. Issue B-32: Ice Effects on Safety-Related Water Supplies Discussion The operating experience during some severe winters has identified physical phenomena which might adversely impact the proper operation of safety-related systems (i.e., the ultimate heat sink) and impair the  !

ability to obtain sufficient cooling water to safely shut the plant down. Typical icing conditions (e.g., surface ice) appear less important than subsurface frazile ice as a flow blockage mechanism.

Pack ice on packed surface ice has, in the past, been assumed

, sufficiently porous to pass the relatively low flows necessary for ultimate heat sink operations. Frazile ice may not be as porous and may, under rare conditions, reduce the flow below acceptable levels. Also, '

forces produced by expanding ice sheets could damage safety related equipment and structures and impair the ability of the ultimate heat sink to function. The purpose of this NRC task is to ensure that operating reactors have the ability to circulate warm water to the intake (or have other processes) to limit ice buildup.

h Current NRC acceptance criteria are provided in Standard Review Plan 2.4.7, " Ice Effects." g i

WAPWR-RC 5.2-20 AMENDMENT 3 E830e:1d AUGUST 1989 l

I This issue, currently identified as a licensing impact issue, is ]

O scheduled for re prioritization.

3 i SP/90 Response The ultimate heat sink is plant specific and outside the scope of the SP/90 design. However, interface criteria for use by applicants in establishing their ultimate heat sink design has been provided in 3 Subsection 9.2.5 of RESAR-SP/90 PDA Module 13, " Auxiliary Systems".

33. Issus B-33: Dose Assessment Methodology Discussion This NRC task is considered a licensing impact issue and involves the 3 maintenance and improvement of calculational capabilities for assessing doses to individuals from radiation and radioactive effluents from normal operation and from radioactive releases from postulated accidents.

g Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10CFR Part 50, Appendix I," provides methods acceptable to the NRC for assessing public exposure to radioactive materials and effluents.

SP/90 Response Westinghouse will completely document and justify any deviations from the NRC Regulatory Guide 1.109 positions during the licensing process for the SP/90 design. No deviations from Regulatory Guide 1.109 have been identified for the SP/90 PDA design.

34. Issue B-34: Occupational Radiation Exposure Reduction Resolution of this issue will be accomplished through TMI Action Plan Item II.D.3.1, " Radiation Protection Plans.a 3 O WAPWR-RC 5.2-21 AMENDMENT 3 5830e:1d AUGUST 1989

[ 35. Issue B-35: Confirmation of Appendix I Nodels for " Calculations of Releases,of Radioactive Materials in Gaseous and Liquid Effluents from Light-Water-Cooled Power Reactors" gi l

Discussion This NRC task involves the revision of models for calculating releases of radioactive materials to improve the accuracy of current NRC models for IDCFR Part 50, Appendix I, calculations.

)

All research programs described in the action plan have been completed  !

3 except for the source term measurement program which was due to be completed in FY1983 or FY1984 depending upon the availability of funding to support the collection of additional data from selected operating reactors. NUREG-0017, Revision 1, " Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from PWRs", once issued, will document the results of NRC efforts related to model enhancement.

SP/90 Response This issue applies to an ongoing NRC administrative activity associated

, with their internal model development and has no impact on the SP/90 design.

36. Issue B-36: Develop Design, Testing, and Maintenance Criteria for Atmosphere Cleanup System Air Filtration and Adsorption Units for Engineered Safety Feature Systems and for Normal Ventilation Systems Discussjon O

This NRC task involved the development of revisions to current guidance and technical positions regarding engineered safety feature and normal ventilation system air filtration and adsorption units. The NRC considers this issue technically resolved with the issuance of Revision 2 WAPWR-RC 5.2-22 AMENDMENT 3 E830e:1d AUGUST 1989 i

4 7 . to Regulatory Guide 1.52, . " Design, Testing,'and Maintenance Criteria for

~( Post-Accident' Engineered-Safety-Feature Atmosphere Cleanup System Air Filtration and Adsorption Units of' Light-Water-Cooled. Nuclear Power Plants," and Revision 1-to Regulatory Guide 1.140, " Design, Testing, and Maintenance Criteria for Normal Ventilation Exhaust System Air' Filtration and Adsorption Units of Light-Water-cooled Nuclear Power Plants."

SP/90 Response Westinghouse will completely document and justify any deviations from the NRC Regulatory Guide 1.52 and 1.140 positions during the licensing process for the SP/90 design. There have been no deviations from these regulatory guides identified for the SP/90 PDA design.

37. Issue B-37: Chemical Discharges to Receiving Waters Discussion In accordance with NRC licensing responsibilities under the National Environmental Policy Act (NEPA), the NRC plans to assess the impact of discharges of chemicals to surface waters. The objective of this assessment is to afford a weighing of impacts of the proposed action and a comparison of alternative actions rather than to provide absolute protection to surface waters.

This task is intended to provide additional insight into the impact of chemical discharges and provide procedures for quantifying the magnitude l

of any such impacts. This improvement in NRC procedures for impact O assessment should provide a clearer division between NRC responsibilities under NEPA and EPA responsibilities under the FWPCA. l l

l O

O WAPWR-RC 5.2-23 AMENDMENT 3 l

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l 1

l There are three specific water quality effects which have been questioned and which will be studied initially by the NRC. These are o Environmental significance of condenser tube copper in cooling water I discharges.

1 o Impact of increased total dissolved solids in receiving waters.

o Significance of chlorinated organic compounds produced during condenser chlorination.

The NRC resolution of this task is expected to be documented in a revision to Regulatory Guide 4.2, " Preparation of Environmental Reports for Nuclear Power Stations."

SP/90 Response Plant specific environmental reports (including consideration of this issue) are the responsibility of each utility utilizing the SP/90 design.

38. Issue B-38: Reconnaissance Level Investigations 9

3 ,

This issue has been dropped and is no longer under review.

39. Issue B-39: Transmission Lines 3 This issue has been dropped and is no longer under review.

40. Issue B-40: Effects of Power Plant Entrainment on Plankton g 3 This issue has been dropped and is no longer under review.

41. Issue B-41: Impacts on Fisheries g 3 This issue has been dropped and is no longer under review. ,

I l

WAPhiR-RC 5.2-24 AMENDMENT 3 ES30e:1d AUGUST 1989

\

1

_ -_____O

< 42. Issue B-42: Socioeconomic Environmental Impacts Discussion l

e As part of the cost-benefit analysis of nucisar power plant licensing

,( applications the NRC is required to assess likely socioeconomic impacts of power plant construction and operation on local communities and the surrounding region. This task encompasses several studies to improve the g NRC's ability to forecast socioeconomic impacts for preparation of V environmental statements and hearing testimony. Areas to be studied include o Nuclear power station construction labor force mobility and residential choices. j o Visual change within a region due to alternative closed cycle cooling ]

systems and associated socioeconomic impacts. l I

l o Impacts of coastal and offshore nuclear generating stations on j recreational and tourist behavior at adjacent coastal sites. l The NRC considers this issue to be technically resolved with the

~

publication of NUREG/CR-2749," Socio-Economic Impacts of Nuclear Generating Stations," and NUREG/CR-2750, "Socio-Economic Impacts of Nuclear Generating Stations: Summary Report on the NRC Post-Licensing Studies."

i l

SP/90 Response 1

This task is associated with a site specific environmental proceedings I issue that is not applicable to Westinghouse in relation to the SP/90 design.

O.

O WAPWR-RC 5.2-25 AMENDMENT 3 1

5830e:1d AUGUST 1989

43. Issue B-43: Value of Aerial Photographs for Site Evaluation Discussion O I

The technique of aerial photography has a long established and proven utility for earth resource inventory and evaluation. Applicants for nuclear construction permits are becoming aware of this and are making increasing use of aerial photographs in their environmental reports. The uncertainties with the methodology at present relate to (A) photo interpretation techniques and the extent to which existing regulatory guidance can be met using this method, (B) fine tuning of the interplay between aerial photography and ground truthing needed to meet licensing '

requirements, (C) quantification of presumed cost advantages of this method, and (D) relative information return from different films, photographic scales, and seasons of coverage. The NRC plans to examine existing regulatory guidance and produce a list of items which might be fulfilled in whole or in part from aerial photographic information.

Field tests on actual sites are planned to be carried out to determine the information return from photographs in relation to regulatory requirements and in relation to conventional ground based data collection efforts. The results are intended to give the NRC a documentary basis for accepting aerial photographic inventories and resource evaluation in environmental reports and for revising existing guidance for making environmental surveys.

Work on this issue has resulted in the publication of NUREG/CF-2861,

" Image Analysis for Facility Siting: A Comparison. of Low and High Altitude Image'Interpretabilty for Land Use/ Land Cover Napping."

SP/90 Response O

This issue is not directed toward affecting the level of safety, but toward improving the efficiency of environmental licensing reviews and therefore, has no impact on Westinghouse in relation to the SP/90 design.

g WAPWR-RC 5.2-26 AMENDMENT 3 B830e:1d AUGUST 1989

44. Issue B-44:' Forecasts of Generating Costs of Coal and Nuclear Plants

~

4 D

Discussion In the performance of National Environmental Policy Act obligations to evaluate alternatives to the proposed action, the NRC .must reach a conclusion as to -the comparative costs of generating power among the feasible alternatives. While alternatives other than coal are treated in L the NRC's Analysis . coal represents by far the most feasible alternative and requires detailed cost comparison equivalent to those performed for nuclear. For several years, the NRC has used a computer code known as CONCEPT to obtain forecasts of plant capital costs. This task involves NRC maintenance of (and development of improvements to) the CONCEPT code so that it remains up-to-date for use in projections of power plant capital cost, front-ord cost, and generating cost forecasts.

The NRC considers this issue to be technically resolved with the publication of ORNL-5470, " Concept-5 User's Manual" and ORNL/TM-6467, "A Procedure for Estimating Non-Fuel Operation and Maintenance Costs for Large Steam-Electric Power Plants."

SP/90 Response This issue is associated with an ongoing NRC administrative activity in relation to environmental proceedings that does not affect plant safety nor impact Westinghouse in relation to the SP/90 design.

45. Issue B-45: Need for Power - Energy Conservation

';iscussion This issue was subsumed into Issus B-2, " Forecasting Electricity Demand" 3 O (refer to item 2 above).

O WAPWR-RC 5.2-27 AMENDMENT 3 5830e:1d AUGUST 1989

46. Issue B-46: Costs of Alternatives in Environmental Design 3 This issue has been dropped and is no longer under review.

O\

47. Issue B-47: Inservice Inspection Criteria for Supports and Bolting of Class 1, 2, and 3 and MC Components I

3 This issue has been dropped and is no longer under review. .

48. Issue B-48: BWR Control Rod Drive Mechanical Failure This issue is not applicable to Westinghouse pressurized water reactor designs.

49. Issue B-49: Inservice Inspection Criteria and Corrosion Prevention Criteria for Containments Discussion General Design Criterion 53, " Provisions for Containment Testing and Inspection," requires, in part, that the reactor containment be designed to permit (A) periodic inspection of all important areas, and (B) an

. appropriate surveillance program. 10CFR Part 50, Appendix J, " Primary Reactor Containment Leckage Testing for Water-Cooled Power Reactors,"

requires a general inspection of the surfaces of the containment prior to any Type A test to uncover any evidence of structural deterioration.

I Containment designs typically utilize any one of the following structural l

I meterials: steel, steel lined reinforced concrete, steel lined prestressed concrete. To date the only detailed criteria that have been devaloped for inservice inspection of containments relate to tendon surveillance for pre-stressed concrete containments. These criteria are contained in Regulatory Guides 1.35, " Inservice Inspection of Ungrouted Tendons in Prestressed Concrete Containments," and 1.90, " Inservice h

Inspection of Prestressed Concrete Containment Structures with Grouted WAPWR-RC 5.2-28 AMENDMENT 3 5830e:1d AUGUST 1989

s Tendons." These regulatory guides deal primarily with the prestressing l hardware;. no detailed inservice inspection criteria exist for the steel liner or other portions of the containment. Similarly, there are no I criteria for inservice inspection of steel containments or steel lined l reinforced concrete containments. In view of this, the NRC believes that detailed and comprehensive criteria need to be developed for performing inservice inspections of all types of containments.

In addition, the long-term corrosion problems of reinforcements and of the steel liner in contact with concrete in concrete containments, or the corrosion of the steel surface in contact with the water in boiling water reactor suppression chambers, have yet to be adequately analyzed. The NRC believes that long-term studies of these corrosion phenomena need to be undertaken to develop criteria and requirements to prevent corrosion in all types of containments.

1 This issue has been identified as a licensing impact issue and requires no action on the part of the applicant. 3

50. Issue B-50: Post Operating-Basis-Earthquake Inspection

. Discussion Section V(a)(2) of Appendix A, " Seismic and Geologic Siting Critaria for Nuclear Power Plants," to 10CFR Part 100 states that licensees will be required to shut down their plants in the event of an earthquake if vibratory ground motion exceeds that of the operating basis earthquake (DBE). Prior to restart the licensee must demonstrate to the NRC that no functional damage has occurred to those features necessary for continued operation without undue risk to the health and safety of the public. In order to determine the capability of a plant to resume operation following an OBE, an adequate inspection of the plant and site area must be performed. The requirements for this post-0BE inspection are also stated in Standard Review Plan 3.7.4, " Seismic Instrumentation."

O WAPWR-RC 5.2-29 AMENDNENT 3 5830e:Id AUGUST 1989

l 1

However, since neither the regulations nor Standard Review Plan 3.7.4 provide details on the extent of such inspections, this NRC task is intended to develop an acceptable inspection procedure.

SP/90 Response Procedures for performing post-0BE inspections are the responsibility of each utility utilizing the SP/90 design. q d

51. Issue B-51: Assessment of Inelastic Analysis Techniques for Equipment and Components This issue was subsumed into Unresolved Safety Issues A-40, " Seismic 3

DesignCriteriaShort-TermProgram,"(refertoSection4.0, item 19).

52. Issue B-52: Fuel Assembly Seismic and LOCA Responses

]

Discussion l 3 This issue was subsumed into Unresolved Safety Issue A-2, " Asymmetric Blowdown Loads on the Reacter Primary Coolant Systems" (refer to Section 4.0, item 2).

53. Issue B-53: Load Break Switch Discussion Plant designs which utilize generator load circuit breakers to satisfy the requirement for ar, immediate access circuit stated in General Design Criterion 17, " Electric Power Systems," must be prototype tested to h

demonstrate functional capability.

]

3 This task has been identified as a Regulatory Impact Issue and involves the preparation of a NRC position to clarify and document the prototype g

testing requirements for generator load circuit breakers and associated WAPWR-RC 8830a:1d 5.2-30 AMENDMENT 3 AUGUST 1989 O1 j

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_ _ _ _ . .____-___-_________a

circuitry used to provide an immediate access circuit. The NRC technical position.has been completed and has been incorporated into a revision to Standard Review Plan 8.2, "Offsite Power System" (FR 35201 dated August 3,1983).

SP/90 Response These systems are the responsibility of the plant specific applicant who 3 l n would be responsible for performing the prototypical tests.

U

54. Issue B-54: Ice Condanser Containments Discussion This task involves two NRC efforts associated with the ice condenser containment concept:

o Verification of the established design margin for ice condenser containments using the NRC CONTEMPT 4 code.

o Reviewing the surveillance programs for ice inventory and functional performance testing at operating facilities to determine whether the l

~

surveillance frequencies should be increased or other action should be taken.

SP/90 Response The design of the SP/90 does not include an ice condenser containment.

Therefore, this item is not applicable to the SP/90 design.

I

55. Issue B-55: Improved Reliability of Target Rock Safety-Relief Valves 1

This issue is specifically concerned with the failure of Target Rock '

safety relief valves in BWRs, and as such has no apparent impact on the SP/90 design.

O WAPWR-RC 5.2-31 AMENDMENT 3 5830e:1d AUGUST 1989

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56. Issue B-56: Diesel Reliability Discussion e An examination of licensee event reports by the NRC on the experience with diesel generators (1969 to 1975) indicated that the emergency onsite diesel generators at operating plants have an average reliability of about 0.94 compared with the NRC's reliability goal of 0.99, as expressed in Regulatory Guide 1.108. A study of the more significant causes of diesel generator unreliability has since been completed, and significant 3 causes and recommended corrective actions are identified in NUREG/CR-0660, " Enhancement of On-Site Emergency Diesel Generator Reliability."

The proposed program for operating plants establishes a graded set of requirements based on the reliability actually exhibited by diesel generators. The proposed program adopts a diesel generator startup reliability of 0.95/ demand as the minimum desired reliability and 0.9/ demand as the minimum acceptable level of reliability. At or below the minimum desired level, licensees would be required to improve their diesel generator reliability and document their program for doing so.

. Diesel reliability will also be a factor in the criteria associated with the resolution of Unresolved Safety Issue, A-44. The principal thrust for the resolution of issue B-56 is to develop guidelines for an acceptable Emergency Diesel Generator reliability program to ensure conformance with the Emergency Diesel Generator target reliability (0.95 3

to 0.975) identified is the proposed resolution of USI A-44. These guidelines will then be used to revise regulatory guides and SRP sections, h

SP/90 Response The SP/90 incorporates the conventional arrangement of two on-site emergency diesel generators; however, the SP/90 is expected to be less h

sensitive to diesel generator failure because of improved reliability of WAPWR-RC 5.2-32 AMENDMENT 3 5830e:1d AUGUST 1989

._______________________O

r AC-independent emergency feedwater (two vs. one turbine-driven emergency

~

feedwater. pump) and incorporation of AC-independent backup reactor coolant pump seal injection capability.

l O '

The present SP/90 emergency diesel generators are specified to have a standard 10 second start time. A recent review of SP/90 accident analyses has indicated that this start time is unnecessarily conservative; at the FDA stage, the start titae will be increased to at 3 least 20 seconds and possibly longer.

The SP/90 energency diesel generators will meet the intent of Regulatory Guide 1.108; as a means to improve diesel reliability, required start time will be extended to 20 seconds or more.

57.~ Issue B-57: Station Blackout This issue was subsumed into Unresolved Safety Issue A-44, " Station 3 Blackout" (refer to Section 4.0, item 22).

58. Issue B-58: Passive Mechanical Failures

, Discussion This NRC task involves a review of valve failure data in a more systematic manner to (A) confirm the NRC's present judgment regarding the likelihood of passive mechanical valve failures, (B) categorize these and other valve failures as to expected frequency. (C)specify acceptance criteria, and (D) determine if and how the results of this effort should be applied in licensing reviews.

SP/90 Response O The failure of passive mechanical valves will be considered in the design of the SP/90 fluid systems. That is, safety systems will be capable of withstanding a single active failure or a passive failure at any time O

WAPWR-RC 5.2-33 NAENDMENT 3 B830e:1d AUGUST 1989

following an initiating event. However, passive failures which are considered to have a low probability (e.g., check valve failing to open) may not be considered.

59. Issue B-59: (N-1) Loop Operation in BWRs and PWRs i Discussion The majority of operating boiling water reactors and pressurized water reactors are designed to operate with less than full reactor coolant flow. If a reactor coolant pump in a pressurized water reactor or a g

recirculation pump in a boiling water reactor becomes inoperative, the flow provided by the remaining (N-1) loops is sufficient for steady-state operation at a power level less than full power. Although safety analysis reports for the licensed plants present (N-1) loop calculations showing allowable power and protective system trip set points, the NRC has disallowed this mode of operation for most plants primarily due to insufficient emergency core cooling analyses.

The purpose of this NRC task is to develop a set of acceptance criteria andreviewguidelinesfor(N-1)loopauthorizationrequests.

SP/90 Response Westinghouse will perform accident analyses and establish technical specification requirements for all modes of operation to be licensed for I the SP/90 design. The SP/90 plant design at this time does not consider 3 operating at N-1 loop conditions. Each plant specific applicant would j l

have to apply for an N-1 loop operation license.

h ;

60. Issue B-60: Loose Parts Monitoring Systems Discussion The presence of a loose (i.e., disengaged and/or drifting) object in the I primary coolant system can be indicative of degraded reactor safety f l

WAPWR-RC 5.2-34 AMENDMENT 3 i 5830e:1d AUGUST 1989

s

.- resulting from failure or weakening of a safety related component. A  !

loose part, whether it be from a failed or weakened component or from an item inadvertently left in the primary system during construction,

' refueling, or maintenance procedures, can contribute to component damage

- and material wear by frequent impacting with other parts in the system.

A loose part can pose a serious threat of partial flow blockage with attendant departure from nucleate boiling which in turn could result "...

failure of fuel cladding. In addition, a loose part increases the potential for control rod jaming and for accumulation of increased levels of radioactive crud in the primary system.

The primary purpose of a loose part detection program is the early detection of loose metallic parts in the primary system. Early detection can provide the time required to avoid or mitigate safety-related damage to, or malfunction of, primary system components.

The NRC considers this issue as technically resolved with the issuance of Regulatory Guide 1.133, " Loose-Part Detection Program for the Primary System of Light-Water-Cooled Reactors."

SP/90 Response The SP/90 design includes a loose parts monitoring system. Westinghouse l 3 will completely document and justify any deviations from the NRC Regula-tory Guide 1.133 positions during the Final Design Approval licensing l 3 process for the SP/90 design.

61. Issue B-61: Allowable ECCS Equipment Outage Periods Discussion I

(

s Surveillance test intervals and allowable equipment outage periods in the technical specifications for safety-related systems are largely based on engineering judgment. This task involves the NRC development of analytically based criteria for use in confirming or modifying these l l

WAPWR-RC 5.2-35 AMENDMENT 3 B830e:1d AUGUST 1989

surveillance intervals and allowable equipment outage periods. Studies performed show the unavailability contribution to the ECI/ECCS systems from testing, maintenance, and allowed equipment outage time ranges from g

0.3 to 0.8 of the total unavailability. Optimization of the allowed outage period and the test and maintenance interval can significantly 3 reduce the equipment unavailability and in turn reduce public risk. -

Using techniques and methods currently available and the modeling from IREP and NREP programs, the optimum equipment test intervals and allowable equipment downtimes can be determined. Technical specifications rould have to be modified to conform to these findings.

SP/90 Response Westinghouse will use probabilistic risk assessment, statistical assessment of reliability and availability, and the Westinghouse statistical set point methodology to specify equipment outage times and surveillance intervals for the SP/90 design. Concerning equipment outage times and surveillance intervals, the Westinghouse objective is to optimize the relationship between outage times, surveillance intervals, reliability, availability, and safety. This optimization will ensure h

that safety needs are satisfied while maximizing plant availability and operability.

62. Issue B-52: Reexamination of Technical Bases for Establishing SLs, LSSSs, and Reactor Protection System Trip Functions 3 This issue has'been dropped and is no longer under review.

63. Issue B-63: Isolation of Low Pressure Systems Connected to the Reactor O

Coolant Pressure Boundary Discussion g There are several systems connected to the reactor coolant pressure boundary that have design pressures that are considerably below the WAPWR-RC 5.2-36 AMENDMENT 3 B830e:1d AUGUST 1989 L-----------_-_-----------------------------

s i:

/ reactor coolant system operating pressure. The NRC staff has required

,'\ that valves forming the interface between these high and low pressure systems have sufficient redundancy to assure that the low pressure systems are not subjected to pressures which exceed their design limits.

Recently, there has been discussions relative to the adequacy of the isolation of low pressure systems that are connected to the reactor coolant pressure boundary. Fast reviews have concentrated on ensuring isolation of the residual heat removal system, which is a low pressure O. system on almost all plants. Current reviews of license applications for new plants are based on NRC guidelines set forth in the Standard Review Plan (mainly Standard Review Plan 3.9.6, " Inservice Testing of Pumps and Valves").

This issue involves activities related to plants licensed prior to  ;

issuance of the NRC Standard Review Plan guidance. A related issue is discussed in Section 6.5 (item 8). ,

O se'ao a en -

Westinghouse will completely document and justify any deviations from the NRC Standard Review Plan 3.9.6 acceptance criteria during the Final 3

Design Approval licensing process for the SP/90 design.

64. Issue B-64: Decommissioning of Reactors Discussion O Decommissioning is defined as the orderly retirement of a nuclear facility from service and the safe disposition of the remaining radioactivity. NRC presently has under development new decommissioning rule's to supplement the present rules. Technical evaluations have been 3 completed and a draft rulemaking environmental impact statement has been prepared. RES has prepared proposed rule amendments for decommissioning O

WAPWR-RC 5.2-37 AMENDMENT 3 5830e:1d AUGUST 1989

which are intended to assure that decommissioning of all licensed facilities will be accomplished in a safe and timely manner and that adequate licensee funds will be available for this purpose.

g Potential changes to NRC regulations include 10CFR Parts 50 and 51 amendments, Regulatory Guide 1.86 revision, new regulatory guides on financial assurance and termination surveys, guidance on format and content for nuclear reactor decommissioning plans, and new standard 3 review plan for decommissioning review.

SP/90 Response The decommissioning concerns addressed in this issue are the responsibility of each utility utilizing the SP/90 design. SP/90 design features provided to limit exposures to ALARA will enhance plant decommissioning efforts.

65. Issue B-65: Iodine Spiking 3 This issue has been dropped and is no longer under review. h

66. Issue B-66: Control Room Infiltration Measurements Discussion A key parameter affecting control room habitability under the conditions  !

described in General Design Criterion 19 " Control Room," and Standard Review Plan 6.4, " Control Room Habitability System," is the magnitude of control room air infiltration rates. Estimates of these rates have been based on data relating to buildings that are substantially different than typical nuclear power plant control room buildings. This task involved the development of an improved data base.

The NRC considers this issue as being technically resolved and acceptance O criteria have been incorporated in Standard Review Plan 6.4, " Control Room Habitability".

WAPWR-P.C 5.2-38 AMENDMENT 3 O

5830e:1d AUGUST 1989 l

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SP/90 Response Westinghouse will completely document and justify any deviations from the NRC Standard Review Plan'6.4 Revision 2 acceptance criteria during the Final Design Approval licensing process for the' SP/90 design. The Control Room Habitability System is described in Subsei: tion 6.4 of 3 RESAR-SP/90 PDA Module 13 " Auxiliary Systems." No deviations from SRP 6.4 have been identified for the SP/90 PDA design.

67. Issue B-67: Effluent and Process Monitoring Instrumentation This issue was subsumed into TMI Action Plan III.D.2.1. Current NRC acceptance criteria are documented in the Standard Review Plan (i.e., 3 Sections 11.2,11.3,11.4,and11.5).

68. Issue B-68: Pump Overspeed During'a LOCA 3

This issue has been dropped and is no longer under review.

d 69. Issue B-69: ECCS Leakage Ex-containment This task was subsumed by TMI-2 lessons learned Item III.D.1.1 (refer to 3 Section3.1, item 26).

70. Issue B-70: Power Grid Frequency Degradation and Effect on Primary Coolant Pumps Discussion O  !

Offsite power system frequency decay, depending on the rate of decay, could provide an electrical brake on the reactor coolant pump motors that could slow the pumps faster than the assumed flywheel coastdown flow

( rates normally used in analyzing loss-of-flow accidents. Task A-35,

" Adequacy of Offsite Power Systems," (refer to Section 5.1, item 35) was used to determine the maximum credible frequency decay rate used by the O WAPWR-RC 5.2-39 AMENDMENT 3 I

5830e:1d AUGUST 1989

l NRC in this task. The NRC considers this issue as resolved with the determination that no additional measures (beyond those documented in Standard Review Plan 8.3.1, "A-C Power Systems (Onsite)") are necessary g

to protect against a frequency decay event.

SP/90 Response Westinghouse will completely document and justify any deviations from the NRC Standard Review Plan 8.3.1 acceptance criteria during the licensing process for the SP/90 design. There have been no deviations from Sar 8.3.11dentitied for the SP/90 PDA design.

71. Issue B-71: Incident Response This issue was subsumed into post-THI requirements for response to 3 incidents, covered in TNI Action Plan Item III.A.3.1, " Emergency Preparedness - NRC Rule in Responding to Nuclear Emergencies."

72. Issue B-72: Health Effects and Life Shortening from Uranium and Coal Fuel Cycles Discussion Current practice in health impact assessments is to convert radiation exposure estimates into estimates of health effects, such as cancer deaths, illness, and life-shortening. However, the models presently being used, such as those in WASH-1400, GESNO, current NRC case related testieony, and' EPA assessments, all suffer from similar weaknesses. A major common weakness, which appears amendable to solution, is related to the correct treatment of competing risks among populations with life h

expectancies, age, and sex distributions that vary with time. Since the NRC staff is currently attempting to assess health effects in the future (e.g., Year 2000 and beyond), it is reasonable to expect significant changes in current population statistics. To make such an assessment, a g

demographic model is required which extrapolates the current population WAPWR-RC 5.2-40 AMENDMENT 3 5830e:Id AUGUST 1989

k into the future, correctly all_owing for competing risks of mortality from s various causes (e.g., accidents, heart disease, and cancer). Failure; to do so results, for example, in hypothetical cancer deaths for people who would statistically die from other causes. In the absence of better predictive models, it is not possible to even evaluate the uncertainty associated with the use of the current simplified methods for estimating health effects and consequent life shortening. Uncertainties in the use -

of current models are v cally magnified when attempting to make i comparisons of health of/ects for the coal and nuclear fuel cycles. l O Current health effects models generally are used for estimating long-term impacts. Chronic exposure may be the primary determinant of the number of deaths for a given period for a given pollutant. However, in the case of nonradiological pollutants from the coal fuel cycle, short-term fluctuations leading to acute exposures may determine the time of death and consequent life-shortening. Current evaluations of the coal fuel cycle generally fail to account for short-term mortality, disease and illness. In addition, short-term effects from chemical pollutants are generally dependent on the prior history of chronic (long-term) exposure.

Current models generally assume linear dose-response relationships even when evidence exists for real or practical thresholds, or where experimental data support a nonlinear dose response relationship.

This task involves the development of models to address these problems so that health effects (morbidity and mortality) can be assessed for both the coal and uranium fuel cycles as completely as current data permit and on a comparable basis. Resolution of this issue will be done through O issue A-20, " Impacts of the Coal Fuel Cycle," (refer to Section 5.1, item 20).

'- SP/90 Response This issue applies to an ongoing NRC administrative code development activity and has no impact on Westinghouse in relation to the SP/90 design.

O WAPWR-RC 5.2-41 AMENDMENT 3 3830e:Id AUGUST 1989 L_________----__.--_

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[ '73. . Issue' B-73: Nonitoring for Excessive Vibration Inside the Reactor Pressure Jessel g I

This issue was subsumed into issue C-12, " Primary System Vibration

' Assessment," (refer to Section 5.3, Item 12).

u e

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WAPWR-RC 5.2-42 AMENDMENT 3 5830e:1d AUGUST 1989

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l5.3'. CATEGORY C ISSUES The following discussions pertain to current Category C issues -in relation' to the SP/90 Edesign. NRC discussions and descriptions of these issues'are con-tained in NUREG-0471,_" Generic Task Problem Descriptions (Category 8, C, and D

' Tasks),"andNUREG-0933,"APrioritizationofGenericSafetyIssues."

i

1. : Issue C-1: ' Assurance of Continuous Long-Term Capability of Hermetic Seals on Instrumentation and Electrical Equipment Discussion l l

Certain classes of instrumentation incorporate seals. When safety related components. within containment must function during post-LOCA cnnditions, their operability is sensitive to the ingress of steam or water. If the seals should become defective as a result of personnel errors in the maintenance of such equipment, such errors could lead to the loss of effective seals and the resultant loss of equipment operability. The NRC believes'that the establishment of a basis for confidence that sensitive equipment has a seal during the lifetime of the plant is needed.

The NRC considers this issue as being technically resolved with the issuance of current criteria for qualification of safety-related elec-trical equipment. This criteria is discussed in detail in Section 4.0, item 14.

SP/90 Response O Electrical components of the SP/90 plant design, which are located inside containment will be qualified in accordance with the requirements of 10CFR50.49, " Environmental Qualification of Electric Equipment Important to Safety for Nuclear Power Plants, and will follow the guidelines of  !

Regulatory Guide 1.89 and meet the criteria established in NUREG-0588,  !

" Interim Staff Position on Environmental Qualification of Safety-Related ,

Electrical Equipment."

O AMENDMENT 3 WAPWR-RC UO91e:1d 5.3-1 AUGUST 1989

1

2. Issue C-2: Study of Containment Depressurization by Inadvertent Spray i

. Operation to Determine Adequacy of Containment External Design Pressure Discussion Inadvertent operation of containment sprays can result in a rapid depres-O' surization of the containment building. Where containment external design pressure may be exceeded many plants have been provided with vacuum breakers or control system interlocks to prevent the containment external design pressure from being exceeded. The depressurization of the contain-  !

ment is a transient behavior and can take place in a short time period.

This NRC task involves the development of a code to be used for the analy- j sis of containment pressure response (both with and without the effects of vacuum breakers or control systems) for the inadvertent spray accident.

The NRC considers this issue as being technically resolved. Standard Review Plan Section 6.2.1.1 is used in reviewing licensee analys2s of '

containment depressurization due to inadvertent spray operation.

SP/90 Resoonse The containment spray function of the SP/90 plant is performed by the low head pumps of the integrated safeguards system (ISS). These pumps take suction from the emergency water storage tank (EWST)whichislocated inside containment. Since the water in this tank is at a temperature, which is close to the temperature of the containment atmosphere, the 3

transient following an inadvertent containment spray operation is much less severe than for a present day plant where the water temperature may I be close to 32*F. As part of the RESAR-SP/90 FDA application, l Westinghouse will perfore analyses in accordance with Standard P.eview Plan  !

6.2.1.1 to demonstrate that containment depressurization following l inadvertent spray operation is not a safety concern. j i

i i

WAPWR-RC AMENDMENT 3 UO91e:1d 5.3-2 AUGUST 1989 i

4

,gJ f3 3[IssueC-3:~ Insulation Usage Within Containment.

Discussion

.)

This issue is included as part of ' Unresolved Safety Issue A-43, . "Contain-ment-Emergency Sump Performance" (refer to Section 4.0, item 21)..

4. ' Issue C-4: Statistical Nethods for ECCS Analysis Discussion Appendix K, "ECCS Evaluation Models," to 10CFR .Part 50 specifies the- -

requirements for ECCS analysis. These requirements ~ presently call for specified. conservatism to be applied to certain models.and assumptions l

used in _the analysis to account for data uncertainties at the time Appendix K was written. The resulting conservatism in the calculated peak clad temperature, however,.has never been. thoroughly compared against .the uncertainty- in = peak clad temperature obtained from a realistically calcu-O lated (best. estimate) LOCA.

The purpose of.this issue was to aid the NRC in the review . of changes to

.- . vendor. ECCS models and in the performance of staff audit calculations of ECCS performance. Therefore, no significant change in public risk is attributed to this issue.

The staff. will allow voluntary use of statistical uncertainty analysis to justify relaxation of all but the required conservatism contained in current ECCS ' evaluation models. Additionally, the NRC is currently 3 1 preparing an ECCS rule change that will allow use of a best estimate model plus a statistical uncertainty to determine the peak cladding temperature.

Until the staff revisions to the Appendix K rule change are implemented, the staff proposed to accept the r.svised ECCS analysis methods for demonstrating conformance to the current Appendix K requirements. This a.

issue has been determined to be a resolved Regulatory Impact issue.

O WAPWR-RC AMENDMENT 3 UO91e:1d 5.3-3 AUGUST 1989

1 1

, l SP/90 Response 1 Best estimate models have not been used in the RESAR-SP/90 PDA 0\

a,p m ation. m hough the p,esent anal,ses, us4ng conse,vative methodology, indicate significant margins to Appendix K limits, Westinghouse may employ best estimate models at the FDA stage to demonstrate even larger margins.

5. Issue C-5: Decay Heat Update 4 Discussion This NRC task involves following the work of research groups in deter-mining best estimate decay heat data and associated uncertainties for use in LOCA calculations.

3 Westinghouse has been active in the ANS decay heat subcommittee (ANS-5.1) and has reviewed and concurred with their findings. Westinghouse has gone on record requesting that the Appendix K rule be more flexible to allow the impact of new experimental data including the new decay heat g

standards. Westinghouse will continue to press for this additional flexibility and will actively support NRC best estimate LOCA calculations which use the new decay heat standards.

The staff has determined that the 1979 ANSI /ANS Standard 5.1 is technically acceptable and has allowed the use of this data to justify 3

relaxation of non-required conservatism in current ECCS evaluation models. In addition, the proposed ECCS rule change being developed by the g

, NRC will allow use of this new data. This issue was determined to be a W f resolved Regulatory Impact issue.

O i

WAPWR-RC AMENDMENT 3 O

UO91e:1d 5.3-4 AUGUST 1989 l

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6. Issue C-6: LOCA Heat Sources Discussion The contributors to LOCA heat sources, along with their associated uncer-tainties, and the manner in which they are combined have an impact on LOCA calculations. An evaluation was made of the combined effect of power l 3 density, decay heat, stored energy, fission power decay, and their associated uncertainties with regard to calculations of LOCA heat sources.

No significant change in public risk was attributed to this issue. The NRC intends to allow the statistical combination of heat sources to justify the relaxation of non-required conservatism in ECCS evaluation models. Also, the proposed ECCS rule change being developed by the staff will allow the statistical combination of LOCA heat sources. This issue 3 was determined to be a resolved Regulatory Impact issue.

SP/90 Response O SP/90 LOCA analyses have been performed using conservative models for LOCA heat sources. As stated in the response to Issue C-4, best estimate models may be used for the RESAR-SP/90 FDA application.

7. Issue C-7: PWR System Piping Discussion Combinations of fabrication, stress and environment have resulted in instances of stress corrosion cracking of low pressure schedule 10 type 304 stainless steel piping systems. Although these systems are not part of the reactor coolant pressure boundary, they are safety related; e.g.,

the containment spray system. The incidence of cracking has been O restricted to thin wall, low pressure, low flow systems. These cracks have occurred adjacent to the weld zones of the thin-walled piping after approximately three to five years of service and were identified by O WAPWR-RC AMENDMENT 3 UO91e:1d 5.3-5 AUGUST 1989 i

volumetric examination, by leak detection systems, or by visual inspec- g tion. In each of the cracking events that have occurred to date, the W affected piping was determined to have been inadvertently exposed to j corrosive environments, such as thiosulfate and chlorides.

Current licensing criteria attempts to minimize the use of sensitized piping in safety-related piping systems and place increased emphasis on the use of corrosion-resistant material in such systems. The purpose of this task is to continue to evaluate operating experience to determine if 4 augmented inservice inspection requirements should be established to fur-ther enhance the reliability of such piping systems.

The WRC considers this issue as being technically resolved with the issuance of NUREG-0691, " Investigation and Evaluation of Cracking Inci- i dents in Piping in Pressurized Water Reactors."

l SP/90 Response 3

No schedule 10 type 304 stainless steel piping is used in the SP/90 plant for safety related systems. This issue is therefore not applicable to the g

RESAR-SP/90 PDA application.

'8. Issue C-8: Main Steam Line Leakage Control System Discussion i

This issue applies to the Main Steam System of a Boiling Water Reactor l 3 (BWR) and is therefore not applicable to a Pressurized Water Reactor such as the RESAR-SP/90 design.

l

9. Issue C-9: RHR Heat Exchanger Tube Failures This issue is not applicable to Westinghouse pressurized water reactor i designs. )

1 WAPWR-RC AMENDMENT 3 0 l UO91e:Id 5.3-6 AUGUST 1989

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10. Issue C-10: Effective Operation of Containment Sprays in a LOCA Discussion This NRC task is intended to respond to a concern of the ACRS about the O effectiveness of various containment sprays to remove airborne radioac-tive materials which could be present within the containment following a LOCA. This concern has been expanded to include the possible damage to equipment located inside containment due to an inadvertent actuation of O the sprays.

The NRC considers this issue as being technically resolved with the issu-ance of ANSI /ANS 56.5-1979, "PWR and BWR Containment Spray System Design Criteria," which is referenced in Standard Review Plan Section 6.5.2.

SP/90 Response The containment spray function of the SP/90 plant is included in the O 4 t or t e < a# re >>t - (iss)- 184- t 6- 8 4e e 4#

accordance with ANSI /ANS 56.5-1979 where applicable.

3

,11. Issue C-11: Assessment of Failure and Reliability of Pumps and Valves 1

This issue is included as part of Unresolved Safety Issue A-45, " Shutdown Decay Heat Removal Requirements" (refer to Section 4.0, item 23).

12. Issue C-12: Primary System Vibration Assessment O Discussion Structural damage to the primary system, including the reactor pressure vessel and internals, associated piping and steam generator tubing in pressurized water reactors can be caused by vibrations of sufficient magnitude. These vibrations can be either flow-induced or the result of operation of the pumps to which primary system piping is attached. There O AMENDMENT 3 WAPWR-RC l D091e:1d 5.3-7 AUGUST 1989

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have been a number of instances where components internal to the reactor coolant pressure boundary have come loose as the result of flow-induced g

vibration and been carried through the primary system by the coolant flow.

Excessive core barrel movement, caused by flow-induced vibration, may lead to many detrimental effects including damage to reactor internals and interference with control rod movement. Problems resulting from excessive core barrel movement have been encountered at Palisades and possibly other operating plants.

Structural damage due to flow-induced vibration of steam generator tubing has also been encountered. Anti-vibration bars are currently utilized to minimize tube vibration. However, fretting has occurred due to deficient design and material selection for the anti-vibration bars.

Piping systems are also susceptible to forced vibration as a result of pump vibration during operation. If a natural frequency of the connected piping is very nearly the same as the driving frequency of the pump there is then the possibility, depending on the amplitude of vibration, for fatigue failures in the system, particularly at the nozzle where the  !

stresses will be highest.

Preoperational testing of reactor internals, piping systems and mechani-cal equipment is conducted during startup functional testing to assure structural and functional integrity per Standard Review Plan 3.9.2,

" Dynamic Testing and Analysis of Systems, Components, and Equipment," and Regulatory Guide 1.20, " Comprehensive Vibration Assessment Program for g

Reactor Internals During Preoperational and Initial Startup Testing."

However, vibration frequency shifts are possible during operation as a result of component and/or component support wear or degradation. Also, vibration effects for the longterm may not have been properly assessed during startup testing. gl Inservice inspection during the life of the plant and possible visual and audible detection of vibration during plant operation may be necessary in i G WAPWR-RC AMENDMENT 3 l UO91e:1d 5.3-8 AUGUST 1989

s order to arrest structural damage already incurred or, if the vibration b- were to continue, might occur at some future time. This vibration assessment could lead to modifications in the design of systems compo-nonts~or' component support arrangements of system operation saquences.

3 The NRC considers this issue as being technically resolved. Current guidelines in SRP 3.9.2, combined with NRC positions on loose parts moni-toring' in Regulatory Guide 1.133 provide sufficient basis for considering

( ' this issue to be resolved..

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SP/90 Response Primary system vibration is addressed in the following manner for the SP/90 plant:

1. Reactor internals will undergo flow induced vibration testing in accordance with applicable guidelines of Regulatory Guide 1.20 for nonprototype plants.

2. The reactor coolant system is continuously monitored for the presence

- of loose metallic parts by a digital metal impact monitoring system 3

, (RESAR-SP/90 PDA Module 4, " Reactor Coolant System" Subsection 4.4.6.4).

3. Each of the four reactor coolant pumps is equipped for continuous monitoring of shaft and frame vibration levels (RESAR-SP/90 PDA Module 4,"ReactorCoolantSystem" Subsection 5.4.1.2.2).

O Westinghouse considers that based on these features, the SP/90 design complies with NRC Standard Review Plan 3.9.2 and the intent of Regulatory Guides 1.20 and 1.133. ,

13. Issue C-13: Non-Random Failures This issue is included as part of Unresolved Safety Issue A-17, " Systems Interactions in Nuclear Power Plants" (refer to Section 4.0, item 13).

l WAPWR-RC 5.3-9 AMENDMENT 3 UO91e:1d AUGUST 1989

14. Issue C-14: Storm Surge Model for Coastal Sites 3 l- Licensees are required to estimate the design basis water levels for each site. For coastal and estuarine sites, the design basis water level is often caused by a storm surge, which results from the wind and pressure fields of an intense storm acting on the water.

The primary tool used by the NRC for estimating storm surge has been the "bathystrophic" model as developed by the " S. Army Corps of Engineers, Coastal Engineering Research Center (CERC). This model is based on the bathystrophic approximation, relating sea surface slope to wind stress, bottom stress, and pressure gradient, with a correction for Corriolis force due to along-shore currents. The NRC considers this model to now be obsolete. Bigger and faster computers are now capable of solving multidimensional dynamic equations which account for many effects not included in the bathystrophic model. The multidimensional dynamic mathe-matical models can account for irregular shorelines, while the shape of the shoreline is not considered at all by the bathystrophic model.

True long wave dynamics are simulated by multidimensional dynamic mathe-matical models, but are completely neglected by the bathystrophic models.

This task called for the development of a replacement for the bathystrophic model so that the staff's evaluation of storm surge reflects state-of-the-art techniques. The storm surge model is applied at the CP stage and is possibly reviewed at the OL stage. Therefore, only future plants located at coastal or estuarine sites will be affected by the issue. I 3

This issue involves development or acquisition of a multidimensional model g which will reflect state-of-the-art mathematical techniques. It is W believed that a new multidimensional dynamic model would eliminate the need for initial estimates (required by the bathystrophic model) and would reduce the total required analysis time. Thus, this item is related to g increasing knowledge that would increase confidence in assessing levels of W safety and, therefore, is considered to be a Licensing Issue.

WAPWR-RC AMENDMENT 3 UO91e:1d 5.3-10 AUGUST 1989

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The staff believes that the existing bathystrophic model (SURGE) is b,n adequate for calculating design basis water levels at future nuclear plant sites. This model is very conservative and is still used by the CERC.

Its use is specified in SRP Section 2.4.5-3. Furthermore, as stated in the SRP, the use of other verified modes is not precluded. Thus, this licensing issue does not require any changes to be made by the staff and it is recommended it be dropped from further consideration.

f. SP/90 Response 3

G The RESAR-SP/90 plant is designed to withstand the design basis water level specified in the site parameters, RESAR-SP/90 PDA Nodule 3

" Introduction and Site." This issue applies to the site specific analysis that needs to be performed to demonstrate that the design basis water level assumed for the SP/90 plant is conservative.

15. Issue C-15: NUREG Report'for Liquid Tank Failure Analysis Discussion Standard Review Plan 15.7.3, " Postulated Radioactive Releases Due to Liquid-Containing Tank Failures," requires an analysis of the conse-

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quences of failure of tanks containing radioactive liquids outside con-tainment. This task involves the development of a NUREG report that is intended to describe a consistent and acceptable method for analyzing the

- effects of a fcilure of a radioactive liquid waste tank.

The current version of Standard Review Plan 15.7.3 does provide certain O' criteria for analyzfng the effects of a failure of radioactive liquid waste tanks. These criteria include:

(m o Limiting radionuclida concentrations to those specified in 10CFR ,

Part 20, " Standards for Protection Against Radiation,"

o Assuming 0.12 percent failed fuel.

O WAPWR-RC AMENDMENT 3 D091e:1d 5.3-11 AUGUST 1989

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1 o Assuming 80 percent volume in failed components. l J

o Credits in analyses that can and cannot be taken.

SP/90 Response Westinghouse will perform an analysis of the consequences of failure of tanks containing radioactive liquids outside containment in accordance with Standard Review Plan 15.7.3 as part of the FDA application for the SP/90 design.

16. Issue C-16: Assessment of Agricultural Land in Relation to Power Plant Siting and Cooling System Selection Discussion Interpretations of the National Environmental Policy Act (NEPA) require that environmental impact assessments include land use impacts and alter-natives in nuclear power plant licensing cases. The NRC has performed both economic and non-oconomic land resource assessments in compliance g

with these NEPA requirements. Some licensing cases have questioned the adequacy of the NRC's resource evaluative methods with respect to large 1ans areas required for sites and cooling lakes. The primary issue con- i corning the NRC's assessment is that neither economic analyses nor resource assessment as currently performed provides a convincing rationale for preemption of high quality land in view of continued population pressures, predicted impending lags in world-wide agricultural food production and probable increasing international demands on th9 United States for exports of agricultural products.

Food and fiber production and distribution rank with energy production and utilization as vital world problems now and for the foreseeable future, g These problems are inextricably linked since energy production facilities W can be consumers of large land areas while energy is a prime requirement-WAPWR-RC 5.3-12 AMENDMENT 3 O

UO91e:1d AUGUST 1989

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L o for evenr modest levels of agricul.tural production. Thus, land use is and b probably will remain a key siting issue in nuclear plant licensing. I This NRC task will involve the conduct of a confirmatory exploration ofl 3 new energy techniques to determine their suitability for application to O environmental licensing assessment under NEPA. A problem of immediate licensing concern to the NRC is the conflict in land use which occurs when power plants with large cooling lakes are sited in regions of prime agricultural land.

This environmental issue is addressed in environmental impact statements on a case-by-case basis. Thus, this item has been dropped from further '3 consideration.

17. Issue C-17: Interim Acceptance Criteria for Solidification Agents for Radioactive Solid Wastes Discussion O There are no current NRC criteria for acceptability of solidification agents. This NRC task involves the development of criteria for accepta-bility of radwaste solidification agents to properly implement a process control program for the packaging of diverse plant waste for shallow land burial.

The NRC considers this issue as technically resolved with the issuance of a proposed rule, " Licensing Requirements for Land Disposal of Radioactive Waste'(10CFR Pa'tr 61)."

SP/90 Response This issue and the associated proposed rule are related to requirements for land disposal of radioactive wastes which are not applicable to Westinghouse in relation to the SP/90 design.

lO WAPWR-RC UO91e:1d 5.3-13 AMENDMENT 3 AUGUST 1989 l E_

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5.4 CATEGORY D ISSUES The following discussions pertain to current Category D issues in relation to the SP/90 design. NRC discussions and descriptions of these issues are (q)

, contained in NUREG-0471, " Generic Task Problem Descriptions (Category B, C, andDTasks)"andNUREG-0933,"APrioritization'ofGenericSafetyIssues."

1. Issue D-1: Advisability of a Seismic Scram Discussion The ACRS has recommended that studies be made of techniques for seismic scram and of the potential safety advantages and potential disadvantages of prompt reactor scram in the event of strong seismic motion, say more than one-half the safe shutdown earthquake. Various suitable techniques l have been identified and exist, but thus far only limited studies have been reported on the pros and cons of seismic scram.

, Appendix A, " Seismic and Geologic Siting Criteria for Nuclear Pcwer Plants," of 10CFR Part 100 requires that suitable instrumentation shall be '

provided so that the seismic response of nuclear power plant features important to safety can be determined promptly to permit comparison of such response with that used as the design basis. Such a comparison is needed to decide whether the plant can continue to be operated safely and to permit such timely action as may be appropriate.

Regulatory Guide 1.12, " Instrumentation for Earthquakes," describes seis-mic instruments' tion acceptable to the NRC staff as satisfying the above stated requirements of Appendix A to 10CFR Part 100. Regulatory Guide 1.12 requires that one triaxial response spectrum recorder capable of providing signals for immediate control room indication be provided at the containment foundation.

O WAPWR-RC AMENDMENT 3 D091e:1d 5.4-1 AUGUST 1989  ;

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These criteria and regulatory guidance do not address the need for instru- g i mentation .that would automatically shutdown a nuclear power plant when an W earthquake occurs which exceeds a predetermined intensity. This issue involves considerations of the need for such instrumentation. 3 J

Westinghouse believes that the automatic shutdown of a nuclear power plant for an earthquake event with a magnitude less than or equal to the opera-ting basis earthquake does not seem necessary. For an operating basis earthquake occurrence the structural integrity of the plant is maintained ,

to the extent that the plant can continue to operate. Therefore, if immediate control room indication is provided in accordance with Regula-tory Guide 1.12, operator action and administrative procedures for plant t shutdown are sufficient for an earthquake less than or equal to the opera- f l ting basis earthquake. I t

SP/90 Response Westinghouse is not considering including a seismic scram in the SP/90 3

design. g

2. Issue D-2: Emergency Core Cooling System Capability for Future Plants j

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This issue is included as part of the Unresolved Safety Issue A-45, " Shut-down Decay Heat Removal Requirements" (refer to Section 4.0, item 23).

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3. Issue D-3: Control Rod Drop Accident (BWRs)

This issue is not applicable to Westinghouse pressurized water reactor designs. {

O WAPWR-RC AMENDMENT 3 O

UO91e:1d 5.4-2 AUGUST 1989

n 5.5 NEW GENERIC ISSUES L)

The NRC continuously evaluates the safety requirements used in its reviews against new information as it becomes available. Sections 5.1 through 5.4 e provide a discussion of NRC generic safety issues identified and categorized .

I i

( by the NRC in 1978. Since that time, new generic safety issues have been  !

identified as a result of licensee event reports, ACRS reports, and other NRC activities. Majorsourcesofnew generic safety issues include NUREG-0572,

" Review of Licensee Event Reports (1976-1978)," and NUREG-0705,

( " Identification of New Unresolved Safety Issues Relating to Nuclear Power Plants." Items under development from NUREG-0660, "NRC Action Plan Developed as a Result of the TMI-2 Accident," and NUREG-0985, "U.S. Nuclear Regulatory Commission Human Factors Program Plan" are also tracked as generic safety issues.

3 New generic safety issues have not been categorized by the NRC in the manner the previous safety issues were categorized (i.e., Category A, B, C, and D).

NUREG-0933 "A Prioritization of Generic Safety Issues," establishes priority rankings of High, Medium, Low and Drop for the current generic safety issues.

The following discussions pertain to these new generic safety issues in relation to the SP/90 design.  !

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1. Issue 1: Failures in Air-Monitoring, Air-Cleaning, and Ventilating Systems l This issue has been dropped and is no longer under review. 3

2. Issue 2: Failure of Protective Devices on Essential Equipment Discussion f The ACRS identified this potential safety concern in NUREG-0572. A large k number of licensee event reports have reported failure or incapacitation of essential equipment as a result of failure of fuses or other devices installed for the sole purpose of protecting that essential equipment or its services. The systems affected exist throughout the plant and include WAPWR-RC 5.5-1 AMENDMENT 3 5888e:1d AUGUST 1989

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the plant control system, the plant protection system, and the engineered safety features. Particularly vulnerable are actuators that require power g

in order to drive motors and operate valves. The failures are not limited to overcurrent protectors but occur in equipment such as torque limiters, overspeed protectors, and other interlocks and may be caused by improper applications or adjustments as well as component failures.

Safety implications arise because the expected failure rate of essential equipment may be overly optimistic because of not accounting for failure of pro',ctive devices. Where failures result from improper selection of fuse sizes or adjustment of protective devices, there is an increased probability of common mode failure of redundant vital services.

In the past, the corrective action has been to replace the failed fuse or readjust the adjustable devices. Where dist,bling of such equipment could remove or substantially degrade vital services, the NRC feels that the basic criteria for protecting the equipment should be reexamined. For example, the NRC believes the rules for protection of vital equipment should perhaps be different than current standard electrical practice, h SP/90 Response The design process for the SP/90 will investigate the above concerns, and the potential for problems will be minimized within existing practice. If necessary, consideration will also be given to the modification of "indus-try practice" for protection of equipment. For example, bypass of certain protective functions under accident conditions might provide a solution.

Any criteria ' modification will be undertaken with adequate consideration given to any increased probability of damage to equipment, the resulting g'

effect on utility financial risk, and within the risk / safety goal considerations.

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WAPWR-RC 5.5-2 AMENDMENT 3 5888e:1d AUGUST 1989 i

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/~j 3.. Issue 3: Set Point Drift in Instrumentation

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' Discussion This issue is identified in Appendix D of NUREG-0572 and is one of the key observations made after the ACRS requested its members and consultants 'to make . comprehensive reviews of all licensee event reports issued during the years 1976, 1977, and 1978.

Data collected over the 3 year period showed that 10 percent of all licensee event repor.ts were related to drift in the set points of instru-mentation beyond technical ' specification limits. This amounted to an average of 258 licensee event reports each year. The proportion of these events that resulted in simultaneous drifts in redundant channels was not established in NUREG-0572.

An unplanned change in the setpointofaninstrument(setpointdrift) will alter the actual'value of the measured parameter at which a particu-lar action is to occur. Excessive drift in an instrument's set point beyond technical specification limits could result in the instrument. not providing timely warning signals prior to or during an accident thereby  !

failing to perform its safety function. All safety instrumentation chan-nels 'are redundant but simultaneous drift of redundan't instruments beyond technical specification limits could affect plant safety. ,

For those instruments where set point drift is due to component failures, a possible solution is to make the necessary repair, recalibrates, and restore the instruments to service. For those instruments where the margin between the selected set point and the technical specification limit is not' sufficient to allow for normal instrument inaccuracy, a possible solution is to increase the margin between the selected set point and the technical specification limit to accommodate the inherent 4 instrument inaccuracy.

O WAPWR-RC 5.5-3 AMENDMENT 3 5888e:1d AUGUST 1989

i SP/90 Response There are two considerations which will reduce any problems with the SP/90 protective system to those associated with component failure.

3 By using the digital integrated protection system, the SP/90 design eliminates some of the problems associated with set point drift in the prutection system. The only portion of the instrumentation which will be subject to drift will be that analog portion from the sensor through the analog to digital converter. The redundant sensor selector will identify any sensor channels th,at have drifted outside of tolerance and will make this information available to the operators.

For the SP/90, the set points for the protection system will be determined 3 using the guidance of Regulatory Guide 1.105, Revision 2, " Instrument Setpoints," which endorses ISA 567.04, 1982 "Setpoints for Nuclear Safety-Related Instrumentation used in Nuclear Power Plants." Use of this guidance will eliminate all of the problems of drift beyond the technical g

specification limits except those associated with component failures.

3 The SP/90 protection system has been designed and set points will be selected using the Westinghouse setpoint methodology approved by the NRC in NUREG-0717, Supplement 4, dated August 1982.

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4. Issue 4: End-of-Life and Maintenance Criteria f Discussion This issue has been addressed as part of the NRC overall equipment qualification program. Existing and proposed requirements include both and-of-life and maintenance considerations. Available material aging information coupled with actual plant operating and maintenance experience, could be factw ed into the process of determining the end-of-life for various components as well as determining appropriate O

WAPWR-RC 5.5-4 AMENDMENT 3 5888e:1d AUGUST 1989

maintenance periodicity. The failure of safety-related components can lead to loss of reactor coolant pressure boundary integrity or loss of safety functions. Such failures possibly could be reduced by using end-of-life data and improved periodic maintenance criteria.

I NUREG-0588, " Interim Staff Position on Environmental Qualification of j Safety-Related Electrical Equipment," and the " Guidance for Evaluating  !

- Qualification of Class IE. Electrical Equipment in Operating Reactors" require that qualification programs for electrical equipment should iden-

\ tify materials susceptible to aging effects and establish a schedule for periodically replacing the equipment and/or materials.

Revision 1 to Regulatory Guide 1.89, " Environmental Qualification of 3

Electrical Equipment for Nuclear Power Plants," includes a number of specific positions on the subject of equipment end-of-life and maintenance. The Regulatory Guide positions are:  !

o If synergistic effects have been identified prior to the initiation of O, qualification, they should be accounted for in the qualification program. Synergistic effects known at this time are dose rate effects and effects resulting from the different sequence of applying

. radiation and (elevated) temperature.

o The expected operating temperature of the equipment under service, 3 conditions should be accounted for in thermal aging. The Arrhenius methodology is considered an acceptable method of addressing accelerated thermal aging within the limitation of state-of-the-art technology. Other aging methods will be evaluated on a case-by-case basis, o The aging acceleration rato and activation energies used during qualification testing and the basis upon which the rate and activation energy were established should be defined, justified, and documented.

O WAPWR-RC 5.5-5 AMENDMENT 3 5888e:1d AUGUST 1989

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o Periodic surveillance and testing programs are acceptable to account for uncertainties regarding age-related degradation that could affect i l the functional capability of equipment. Results of such programs will I be acceptable as ongoing qualification to modify designated life (or qualifiedlife) of equipment and should be incorporated into the maintenance and refurbishment / replacement schedules.

o Sections 6.4 and 6.5 of IEEE 323-1974 discuss qualification by ,

operating experience and by analysis, respectively. The adequacy of these methods should be evaluated on the basis of the quality and detail of the information available in support of the assumptions made. Operating experience and analysis based on test data may be used where testing is precluded by the physical size of the equipment or the state of the art of testing. When the analysis method is employed because of the physical size of the equipment, tests on vital components of the equipment should be provided.

The NRC is in the process of coding (refer to Section 6.1.2.3, item 5) the similar requirements for mechanical equipment.

h The NRC Standard Review Plan, Section 3.11., " Environmental Qualification of Nechanical and Electrical Equipment," includes requirements for main-tenance/ surveillance programs for equipment located in mild environments.

Specifically, it is required that "the maintenance / surveillance program data shall be reviewed periodically (not more than every 18 months) to ensure that the design qualified life has not suffered thermal or cyclic degradation resulting from the accumulated stress triggered by the abnormal environmental conditions and the normal wear due to its service condition. Engineering judgment shall be used to modify the replacement program and/or replace the equipment as deemed necessary."

SP/90 Response Ol l The SP/90 design will provide for design improvements in maintainability and an extension of the time between maintenance periods as practicable.

O WAPWR-RC 5.5-6 AMENDHENT 3 B888e:1d AUGUST 1989

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_ x In addition, Westinghouse will fully document the -level of conformance L

.with the_~ regulatory ' positions of Regulatory Guide'1.89 and the acceptance criteria of SRP-3.11 during the Final Design Approval- (FDA) Ifcensing process for the SP/90.

L (.)

5. ' Issue 5: Design Check and Audit of Balance-of-Plant Equipment Discussion nO This issue, which as subsumed into I.F.1, involves a potential 3 igrovement that might'be achieved by requirements for verification that the- balance-of plant "as-bailt" configuration satisfies _the design intent. Such action could improve 'the reliability of balance-of plant equipment .and reduce demands on safety equipment. This issue has arisen-1 because of failures of balance-of plant equipment to perform as intended for many reasons and as a result, place various demands on safety systems.

-The SP/90,- in moving toward a nuclear power block concept, has placed more O

i of the plant scope within the Westinghouse sphere of direct control. By so doing, portions' of the concern described here are of less importance

'because that portion of equipment which represents balance-of plant is further removed from the plant safety equipment.

Regardless of this consideration, some greater capability'to verify that as built conditions accurately reflect design needs will be required in the future. In - a one-step licensing process, there is a strong need to certify that the plant has been built'as licensed and the commitments made in the safety analysis report have been fulfilled. As part of the fulfillment of this verification, some consideration should be made to verify that balance-of plant systems adequately support the plant and will not unnecessarily increase the challenges to elements of the nuclear power block concept.  ;

-l SP/90 Response There is no direct impact on the SP/90 design posed by this issue. 3 l

WAPWR-RC 5.5-7 AMENDMENT 3 5888e:1d AUGUST 1989

6. Issue 6: Separation of Control Rod from its Drive and BWR High Rod Worth Events This issue is not applicable to Westinghouse pressurized water reactor designs.

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7. Issue 7: Failures Due to Flow-Induced Vibrations 3 This issue has been dropped and is no longer under review.

8. Issue 8: Inadvertent Actuation of Safety Injection in PWRs O

Discussion 3 This issue was subsumed into I.C.1 of the TMI Action Plan. Operator

- errors, instrument malfunction, and reactor tiaasients and trips have been reported as the cause of inadvertent actuation of the safety injection system. At least 40 cases of inadvertent actuation of safety injection have been identified in NUREG-0572. Approximately one-fourth of the events sampled were due to operator error. The problem is repetitive in g

nature; at several facilities the problem has a long history. The vast majority of events occurred in Westinghouse nuclear steam supply systems, whereas plants supplied by other vendors had few or no reported events.

Safety injection systems are required to operate during loss-of-coolant accidents and other severe transients that require berated water addition to the primary system. Inadvertent actuation of the system injects cold borated water into the reactor when it is not needed, subjecting injection nozzles to thermal stresses and requiring removal of boron from the pri-mary system before startup. The present number of occurrences is probably not significant with respect to the effects upon the primary system; however, operator response to an inadvertent safety injection involves termination of the injection and resetting of the injection signal. This generally occurs within 1 to 8 minutes following the start of injection 1

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WAPWR-RC 5.5-8 AMENDMENT 3 B888e:Id AUGUST 1989 i

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s and follows a check cf other. plant status instrumentation. Repeated

} 9 operator exposure to inadvertent safety injection and its termination may produce an unacceptable response in cases where the injection is required to provide core cooling water.

SP/90 Response The SP/90 design will be less likely to experience a spurious reactor trip or an inadvertent safety injection. The protection system objectives pro-vide for a reduced probability of spurious actuation due to the failure of any single component, or system and increased margin between the low pres-surizer pressure safety injection set point and the minimum pressurizer pressure reached following a reactor trip from full power. The control system and the human factors engineered control room will lower the 3 probability of putting the SP/90 into a state from which a reactor trip from full power would result in a safety injection. For example, improved steam generator feedwater control will prevent steam generator related inadvertent safety injection. The SP/90 control room will also make the g assessment of plant safety problems both more reliable and easier to make. Because of SP/90 control room related improvements there is a much 3

greater certainty that the operations personnel will recognize the need for a safety injection. Additionally, the sizing of reactor coolant system components will be performed with an objective of increasing the margin between the low pressurizer pressure safety injection set point and the minimum reached following a reactor trip from full power.

If the SP/90 does experience an inadvertent safety injection, the SP/90 3

control room in conjunction with plant procedures will aid in the 9 assessment of plant state. Additionally, any concerns over combir ed pressure and thermal stresses to injection nozzles will be reduced as the shut off head of the safety injection pumps will be such that injection will not occur following reactor trip.

The SP/90 design described above, particularly the protection system design and the sizing of reactor coolant system components, eliminate inadvertent safety injection as a problem in the SP/90.

WAPWR-RC 5.5-9 AMENDMENT 3 5888e:1d AUGUST 1989

9. Issue 9: Reevaluation of Reactor Coolant Pump Trip Criteria I

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Discussion 1

The issue of reevaluation of reactor coolant pump trip criteria involves l the potential improvement that might be achieved by establishing better criteria on when to allow the operation of reactor coolant pumps and when I to trip them. It was believed that better criteria might allow the use of reactor coolant pumps to aid in recover from certain transients while still ensuring that these pumps could be tripped during small-break LOCA.

1 3 This issue was subsumed into post-TMI licensing requirement II.K.3(5) )

(Section 3.3.1, Item 4 of this document) and is fully discussed in NRC Generic Letters83-10c and 83-10d (Section 6.4, Items 92 and 93 of this document). 2 1

SP/90 Response See the above referenced items for a complete discussion of this item and its relation to the SP/90.

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10. Issue 10: Surveillance and Maintenance of Tranversing Incore Probe Isolation Valves and Squib Charges This issue has been dropped and is no longer under review.

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11. Issue 11: Turbine Disc Cracking  !

Discussion 3 This issue was raised because of the discovery of stress corrosion cracking in the low pressure discs of Westinghouse-designed turbines. 1 This issue was subsumed into Issue A-37, "1urbine Missiles." Refer to j Section 5.1 (item 37) for a discussion of this issue.

9l1 WAPWR-RC 5.5-10 AMENDMENT 3 5888e:1d AUGUST 1989 i l

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12. Issue 12: BWR Jet Pump Integrity ,

< This issue is not applicable to Westinghouse pressurized water reactor designs.

' .13. Issue 13: Small Break LOCA frca Extended Overheating of Pressurizer Heaters This issue has been dropped and is no longer under review. 3

14. Issue 14: PWR Pipe Cracks ,

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Discussion Cracking has occurred in PWR piping systems as a result of stress corrosion, vibratory and thermal fatigue, and dynamic loading. However, to date, no cracking has been experienced in the primary system piping of PWRs. Thus far, all incidents of cracking have been. detected and corrective actions taken prior to any catastrophic failures.

Cracking in PWR nonprimary system piping could lead to a lessening of the system functional capability and possibly result in situations such as degraded core cooling. Cracking in PWR primary system piping has not been experienced, and the mechanisms and environmental conditions necessary to initiate and propagate the cracking in this piping are not known to exist. Tnerefore, the risk associated with PWR pipe cracks is negligible for the primary system and low for the other piping systems.

The third Pipe Crack Study Group was established in 1979. The charter of the PWR Pipe Crack Study Group included (A) the causes and safety signifi-cance of pipe cracks in PWR safety-related systems, (B) the ability of current inservice inspection and leak detection techniques to detect these cracks, and (C) recommendations for both upgrading the licensing process for plants in the operating license and construction permit stages and for implementation of new criteria on operating riants. In September 1980, the PWR Pipe Crack Study Group completed its investigation of this issue O l WAPWR-RC 5.5-11 AMENDMENT 3 AUGUST 1989 5888e:Id

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and published its findings as NUREG-0691 " Report of Investigations and Evaluations of Cracking Incidents in Piping in Pressurized Water g

Reactors." This report provides conclusions regarding systems safety and reconnends technical solutions to the issue. As a result of issuing NUREG-0691, the NRC considers this issue to be technically resolved.

SP/90 Response The SP/90 design will follow the recommendations of NUREG-0691 in minimizing the potential for cracking in SP/90 piping systems. The SP/90 analyses will also demonstrate that the criteria of NUREG-0691 are met.

Calculations as described in the NUREG-0691 will be performed to assure that safety systems, particularly safety injection, will perform acceptably under analyzed break situations.

15. Issue 15: Radiation Effects on Reactor Vessel Supports Discussion This issue was first identified in June 1978 when Virginia Electric and O

Power Company filed a notification for its North Anna plant in accordance

, with 10CFR Part 21, " Reporting of Defects and Noncompliants."

Reactor pressure vocsel external steel support structures may become embrittled by neutron radiation to the point where their structural integrity may be impaired by virtue of reduced fracture resistance. The theory is that neutrons with less than 1 NeV of ~ energy can induce significant damage te supports because of their relative abundance.

Additionally, compared to the reactor vessel, supports operate at low g

temperatures thereby making concurrent annealing during operation very small. Structural steels vary widely throughout the industry and the problem could be quite' severe at some plants.

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WAPWR-RC 5.5-12 AMENDMENT 3 8888e:1d AUGUST 1989 l

v Thus, embrittlement damage to reactor vessel supports can result in their O' failure to adequately support the reactor vessel under large load condi-tions such as an earthquake or a loss-of-coolant accident.

SP/90 Response The SP/90. design is less prone to embrittlement of reactor vessel 3 supports.- Improvement of the core baffle / reflector region to provide p increased shielding of the reactor vessel also reduces the affect on the d fracture toughness of supports.

The SP/90 ' design and safety analysis demonstrates that support structures ,

for vital equipment are'adequato under design basis loading conditions.

l The supports for the reactor pressure vessel have been evaluated for.their 3 l adequacy under appropriate loading combinations. This includes a demonstration that the reactor pressure vessel steel support structures will not become embrittled by neutron irradiation to the point - where their fracture resistance is reduced to a level which yields unacceptable results under design basis loads. 3

16. Issue 16: BWR Main Steam Isolation Valve Leakage Control Systems This issue is not applicable to Westinghouse pressurized water reactor designs.

17. Issue 17: Loss of Offsite Power Subsequent to a LOCA This issue has been dropped and is no longer under review. 3

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O i WAPWR-RC 5.5-13 AMENDMENT 3 5888e:1d AUGUST 1989

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18. Issue 18: Steamline Break with Consequential Small LOCA 1

Discussion i

This issue can be broken down into twc issues:

o Steamline break with a subsequent small LOCA resulting from )

failure of a partially degraded steam generator tube (s). I o Steamline break with a subsequent small LOCA (other than a steam generator tube rupture) resulting from a stuck-open power-operated relief valve or safety valve actuated during the primary system transient or resulting from pipe whip or jet impingement from the broken steam line.

In PWRs, the potential exists for steamline breaks consequently leading to a small primary system LOCA. NRC analysis has indicated that the primary pressure and the pressurizer level may chenge qualitatively in the same way during a combined LOCA compared to a primary break, a steamline break, or a steam generator tube rupture. For the primary temperature and h

i secondary pressure, a combined LOCA behaves qualitatively like a steamline break. For these latter two parameters, a primary rupture or steam l generator tube rupture appear clearly distinct from the behavior of a combined LOCA.

Thus, two concerns have been identified which could increase the risk associated with these issues. These are (1) the possibility of primary side LOCAs ma'y be increased thrcugh the consideration of new initiating mechanisms, and (2) the symptoms of a combined primary / secondary blowdown g

may increase the possibility for operator error through misinterpretation and improper action.

The conclusion reached is that operator mi4 interpretation could supply the greatest contribution to the probability of an accident.

O WAPWR-RC 5.5-14 AMENDMENT 3 BS88e:Id AUGUST 1989 i

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i SP/90 Response '

The SP/90 design incorporates steam generator design. improvements .which reduce the problems associated with tube degradation. -This in turn decreases the probability of a steam generator tube rupture following a steam 'line break. Additionally, the criteria for plugging of tubes has been reviewed in order to minimize the probability of this event, and 3 instrumentation to address Reguhtory Guide 1.97, Revision 2 (refer to p

v Section 3.1, item 23) .has been provided to ensure capability 'to detect a tube rupture following a steam line break.

i 3

In a similar sense, the ~ work done in conjunction with the EPRI valve testing ~ program (described in' some detail in Section 3.1, item 15) provides a greater assurance that the safety and relief valves of the 3 SP/90 will close when required. Instrumentation (described in some detail in Section 3.1, item.16) has also been provided which will positively indicate the positions c these valves and allow appropriate procedures to be implemented to bring the plant to a safe shutdown.

Besides the work described above, which provides assurance that this issue is of little concern for the SP/90 design, procedures will be written to

, address these events following a steam line break.

19. Issue 19: Safety Implications of Nonsafety Instrument and Control Power Supply Bus Discussion O The issue was subsumed into Unresolved Safety Issue A-47, " Safety 3 Implications of Control Systems" (refer to Section 4.0, item 25).

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20. Issue 20: Effects of Electromagnetic Pulse on Nuclear Power Plants I Discussion j l

The electromagnetic pulse (EMP) from a high altitude nuclear weapon '

detonation will induce electrical transients in the instrumentation, control and power lines of nuclear power plants. The extent to which these EMP transients may cause critical plant electrical and electronic systems to fail or malfunction and ultimately result in damage to the reactor is being investigated. A single EMP could affect most of the nuclear power plants in the continental United States. EMP-like effects can also be simulated locally using truck-transportable land based generators. The NRC regulations (10CFR 50.13) state that license appli-cants are not required to provide design features or other measures for the specific purpose of protection against the effects of (A) attacks and destructive acts; including sabotage, directed against the facility by an enemy of the United States, whether a foreign government or other person, or (B) use or deployment of weapons incident to U.S. defense activities.

The present NRC investigation was initiated as a result of informal staff discussions with five Commissioners in 1979. Subsequently, Commissioner '

Ahearne (then Chairman) instructed the staff to plan and carry out this investigation. The objectives of the investigations are (A) to determine the vulnerability of selected safe shutdown systems of a specific nuclear plant to EMP effects due to nuclear weapon detonations and non-nuclear '

generators, (B) to determine how those safe shutdown systems vulnerable to EMP may best be hardened against EMP, and (C) to characterize to the i extent possible the effects of EMP on nuclear plants in general based on the study of specific systems of the subject plant. The overall objec-tive is to provide the Commission with a basis for considering the need for amending the regulations to include design requirements for the protection of nuclear power plants against effects of EMP.

O WAPWR-RC 5.5-16 AMENDMENT 3 l i

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(~'i A technical assistance program with Sandia National Laboratory (SNL) was' initiated in August 1980 to implement the investigation. The Watts Bar plant was selected for the study. The program includes EMP coupling analysis, evaluation of failure thresnold of selected safety equipment,

., and an onsite test program to obtain data for confirmation of the results of analyses. The preliminary conclusion is that the safe shutdown systems at Watts Bar would not be damaged by EMP. The major work remaining to be completed is the extension of these results to nuclear power plants in p

V general, and the preparation (by Sandia) of the interim report and the draft final report. An NRC staff report is planned for late summer, 1982.

EMP concerns during the peacetime operation of nuclear power plants derive from EMP which could be produw d by terrorist actions involving nuclear l weapon detonations or notic.uclear generators, or which could result from accidents involving U.S. or foreign weapons systems. The determination of the probability of occurrence of these types of EMP events is not within the scope of the current EMP investigation. However, consideration of effects due to nonnuclear generators is included in the investigation.

O, The NRC preliminary conclusion is that significant threat does not exist from nonnuclear generators because of the difficulty of deploying and

. operating such equipment in the vicinity of a plant without being detected, and because the effects of this type of equipment are low level and highly localized.

The NRC considers this issue to be technically resolved with the issuance of the final report, NUREG/CR-3069, " Interaction of Electromagnetic Pulse with Commercial Nuclear Power Plant Systems," and is included in the report to the staff on EMP, SECY-82-367. The results indicate that commercial nuclear . power plants are invulnerable to EMP and that there is nothing affected that impacts any systems required for safe shutdown of

( the plant.

O WAPWR-RC 5.5-17 AMENDMENT 3 B888e:1d AUGUST 1989

4 SP/90 Response Given the above resolutions, this issue has no impact on the SP/90 design.

21. Issue 21: Vibrat' ion Qualification of Equipment 3 This issue has been dropped and is no longer under review.

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22. Issue 22: Inadvertent Boron Dilution Events Discussion Many pressurized water reactors have no positive means of detecting boron dilution during cold shutdown. Some operations carried out during outages (e.g., steam generator decontamination) reduce the reactor coolant system volume, thus speeding up dilution. Boron dilution has taken place during such operations although, thus far, criticality has not occurred.

The fix is to install instrumentation to detect the event and stop the dilution either automatically or, if the detection is sufficiently early, by alerting the operator.

SP/90 Response The SP/90 protection system design incorporates boron dilution mitigation logic. The system generates a signal based on flux doubling; the flux doubling algorithm receives its input from the source range neutron flux detectors.

g 3

Administrative controls in the chemical and volume control system ensure j that during shutdowns the only potential source of low boron water would '

be through the volume control tank to the charging pumps. Upon generation i of a flux doubling signal, the valves at the outlet of the volume control tank close, thus stopping the dilution event. An alternate suction path O

WAPWR-RC 5.5-18 AMENDMENT 3 E888e:1d AUGUST 1989

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is automatically aligned to the charging ~ pumps from the spent fuel pit;  !

5 g

this source is assured to be at the refueling concentration of boric _ acid, 3 j and so boration of the reactor coolant system will automatically start.

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23. Issue 23: Reactor Coolant Pump Seal Failures

• O I Discussion This issue deals with an unexpectedly high rate of failures of reactor O coolant pump seals in pressurized water reactors. A seal failure results M in a primary coolant leak (i.e., a very small LOCA).  !

The results reported : in' WASH-1400, " Reactor Safety Study.- An Assessment of Accident Risk is U.S. Commercial Nuclear Power Plants," indicated that a break in- the reactor coolant pressure boundary having an equivalent diameter in the range of 0.5 to 2 inches was a significant cause of a core- l melt. Since- the' current study shows that comparable break flow rates have  !

resulted from reactor coolant pump seal failures at a frequency about an  !

order of magnitude greater than the pipe break frequency used in  ;

WASH-1400, the overall probability of core melt due to these small-size breaks could 'be dominated by events such as pump seal failures if the l

W ASH-1400 assessment is correct. Using the current estimates of seal i failure rates and WASH-1400 scenarios for core melts induced by small ,

LOCAs,'the NRC estimates a core melt frequency of approximately 10'4 pr reactor year.

For ranking purposes, NRC is interested primarily in the frequency of seal failures which result in the release of radioactivity. Seal failure is i involved in many accident sequences, which lead to a spectrum of releases.

Possible solutions to this issue include special detectors that signal i high leakage, more frequent seal replacement, new seal designs, and more smoothly running pumps that take longer to mechanically degrade the seals. f i

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O l WAPWR-RC 5.5-19 AMENDMENT 3 l B888e:1d AUGUST 1989

This issue also is incorporated in the probability considerations of Unresolved Safety Issue A-44, " Station Blackout".

SP/90 Response Over the years, a number of improvements in Reactor Coolant Pump (RCP) l seal design and supporting systems configuration have been introduced in a continuing effort to reduce the incidence of RCP seal failures. RCP seal improvements include:

o Replacement of the aluminum oxide No. 1 seal faceplate with O

silicon nitride; this reduces the potential for damage to the seal surface in case of slight rubbing at low pressure conditions.

o Upgrading of the sealing surfaces on the No. 2 and No. 3 seals from aluminum oxide coating to chrome carbide; this alleviates the concern that the aluminum oxide coating might degrade under certain reactor system conditions.

3 o Incorporation of chrome carbide on the balance diameters of all inserts and housings; this coating improves the effective wear life of these components.

o Optimization of the No. 2 seal to include chrome carbide faced inserts, chrome plated anti-rotation pins and lower durometer 0-rings; these improvements increase wear life and prevent seal hang-up with resulting excessive leakage.

Support system improvements include:

o Incorporation of a temperature instrument in the No. I seal injection line with remote indication and high temperature alarm in the Main Control Room; this will provide added protection against damage to the seals and related components caused by excessive seal injection temperature.

O i WAPWR-RC 5.5-20 AMENDMENT 3 j B888e:1d AUGUST 1989 -

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o. ' Upgrading::of: .the existing flow meter.in the No. I seal injection line to provide an expanded range', and remote. indication and alarm in the Main' Control Room; this will help to ensure that sufficient injection flow is present prior. to RCP startup, as well as to aid in the verification of positive labyrinth flow.

o Addition. of a check val've to the No. 2 seal leakoff line to prevent reverse flow and crud from entering the No..2 and No. 3 i seals; this protects each RCP from many of the effects of abnormalities in other'RCP's. i o Implementation of an-interlock to prevent RCP startup if the No. I seal differential: pressure is less than the minimum limit; this l will protect against damage during the critical startup operation.

Incorporation of another interlock to prevent RCP startup if the

~

o 3 oil. lift system supply pressure is inadequate; this will help prevent damage to the thrust bearing and other pump parts, which could occur if the pump were started with excessively low oil lift system pressure.

A number of operating plants include-several of the above features; this has contributed to the continuing improvement in RCP operation as measured by RCP contribution to nuclear plant unavailability (Figure 23). The SP/90 plant will include all of the above improvements and can therefore be expected to further improve ori this record.

The foregoing addresses RCP reliability during normal cenditions, i.e.,

startup and power operation. The other major issue is RCP integrity )

during loss-of-all-AC or station blackout conditions, when normal RCP seal support system (seal injection and thermal barrier cooling) are no longer available. _

1 O l WAPWR-RC 5.5-21 AMENDMENT 3 i B888e:1d AUGUST 1989 l

w It should be noted that the severity of this issue has diminished since the results of a test program in France have become available; this program has shown that RCP leakage that develops following loss of support systems is considerably less than originally assumed. This will extend the time to onset of core uncovery, and thus increase the probability that either off-site or on-site power can be restored.

Nevertheless, to further reduce the probability of this sequence of events developing, a backup seal injection system is included in the SP/90 plant. This is a non redundant control grade system which is independent of normal off-site and emergency on-site power supplies, and which is actuated on loss of normal seal injection; it is part of the Chemical and Volume Control System (CVCS) which is contained in Section 9.3.4 of RESAR-SP/90 PDA Module 13, " Auxiliary Systems." The Probabilistic Safety 3 Study as contained in RESAR-SP/90 PDA Module 16 has shown that this AC independent seal injection capability is effective in reducing the  !

contribution of loss-of-all-AC events to plant risk.

The SP/90 plant adequately addresses the safety concerns contained in Generic Safety Issue 23 " Reactor Coolant Pump Seel Failures" by:

g o Upgrading existing RCP seal and associated supporting systems designs to enhance reliability during startup and power operations, o Adding a backup seal injection system to prevent RCP failures during loss-of-all-AC condition.

o Determining by test that leak rates following RCP seal failure are low, such that early core uncovery will not occur.

Westinghouse will review the p7oposed resolution of this Generic Safety Issue once it becomes available, but does not believe that design changes will be required in the future.

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O WAPWR-RC 5.5-22 AMENDMENT 3 B888e:Id AUGUST 1989 l

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I

I s

O I

-i

3. 0 -

O ,,

2.5 x 2.0 -

3

. 1.5 -

t O' 1.0 -

.5-1

0. , , , i e i e i, i i 75 77 79 31 83 SS O

O Figure 23. Westinghouse Reactor Coolant Pump Unavailability O

WAPWR-RC 5.5-23 AMENDMEN1 3 5888e:1d AUGUST 1989 a-______-_____________-_________

T ry 24. Issue 24: Automatic Emergency Core Cooling System Switch to Recirculation l O

Discussion g3 The emergency core cooling system (ECCS) operation has two different

(,) phases, the injection phase and the recirculation phase. The first phase i

(injection) involves initial cooling of the reactor core and replenishment {

of the primary coolant following a LOCA, while the second phase (recircu-g lation) provides long-term cooling during the accident recovery period.

([ Switchover from the injection phase to the recirculation phase includes alignment of a number of valves to the recirculation position. Switchover can be achieved by a number of manual actions, by automating these actions or by automatic realignment of certain valves and manual completion of the switchover process. This last option is referred to as the semiautomatic option. The three switchover options (manual, automatic, and semiauto-matic) are vulnerable with varying degrees to human errors, hardware fail-ures as well as common cause failures. Moreover, an automatic system designed to control the whole switchover process or a portion of it can

] reduce the impact of operator error in executing the switchover. However, '

automatic systems may be subject to spurious actuation. Spurious switch-over of ECCS and containment spray pump suction to a dry containment sump can result in pump damage and possible loss of rafety function resulting in potentially unacceptable safety consequences. Review of past reactor experience indicated the existence of a significant number of ECCS spuri-ous actuations and, in particular, four ECCS spurious automatic switch-over actuations occurred in 1980 at Davis-Besse Nuclear Plant, Unit 1.

Subject to the limitation of certain human factors assumptions, the auto-p matic option provides minimum risk to the public. Moreover, it is this V option which is apparently current practice in newer plants. Thus, unless Issue B-17 " Criteria for Safety-Related Operator Actions," (refer to ,

I Section 5.2, item 17) yields new information which modifies the human-q factors assumptions, the NRC believes this issue can be considered V resolved from a generic standpoint.

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V WAPWR-RC 5.5-24 AMENDMENT 3 BBBBe:Id AUGUST 1989

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[] In the design of current plants, the switchover described above must be performed whether manually or automatically. For the SP/90 the problem of

'switchover for ECCS is eliminated. The emergency water -storage tank inside containment establishes a continuous circulation path for safety injection with no actions either manual or automatic.

SP/90 2esponse For the SP/90 there is no impact as a result of this issue as discussed

(\ above.

25. Issue 25: Automatic Air Header Dump on BWR Scram System This issue is not applicable to Westinghouse pressurized water reactor designs.

26. Issue 26: Diesel Generator Loading Problems Related to SIS Reset on Loss of Offsite Power O /

This issue was subsumed into "New" Generic Issue 17, which has since been 3

dropped.

27. Issue 27: Manual Versus Automated Actions Discussion Plant design reviews and emergency operating procedures reviews have

( raised questions as to whether certain safety actions have to be accomplished automatically or whether manual operator action would be acceptable. There are no generally accepted criteria for safety-related operator actions and guidelines in current use are too ill-defined to form a basis for criteria. ANS-58.8 (ANSI N660), " Time Response Design Criteria for Safety-Related Operator Actions," is intended to fill this void and to serve as a basis for future designs.

WAPWR-RC 5.5-25 AMENDMENT 3 B888e:1d AUGUST 1989 l-

w This issue was subsumed into Generic Safety Issue B-17 " Criteria for g 3

Safety-Related Operator Action" (refer to Section 5.2, item 17). W

28. Issue 28: Pressurized Thermal Shock Discussion This issue was subsumed into Unresolved Safety Issue A-49, " Pressurized ThermalShock(refertoSection4.0, item 27).

Bolting Degradation or Failure in Nuclear Power Plants O

29. Issue 29:

Discussion Issue 29 includes all safety related bolting, with the emphasis on reactor coolant pressure boundary bolting degradation or failure.

There are numerous bolting applications in nuclear power plants.' The most crucial bolting applications are those constituting an integral part of the primary pressure boundary such as closure studs and bolts on reactor vessels, reactor coolant pumps, and steam generators. Failure of these bolts or studs could result in the loss of reactor coolant and thus jeop-3 ardize the safe operation of nuclear power plants. Other bolting applica-tions such as component support and embedded anchor bolts or studs are essential for withstanding transient loads created during abnormal or i accidental conditions.

In recent yea'rs, the number of bolting related incidents reported by the licensees of operating reactors and reactors under construction has increased. A large number of the reported belting incidents are related to primary pressure boundary applications and major component support structures. Therefore, there is increasing concern regarding the integrity of the primary pressure boundary in operating nuclear power l plants and the reliability of the component support structures following a l

LOCA or earthquake.

O' WAPWR-RC 5.5-26 AMENDMENT 3 j 5888e:1d AUGUST 1989 i

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I Most of the bolting incidents- reported to date were discovered either )

during refueling. outages or scheduled inservice inspections or maintenance / repair outages. Therefore, such reported incidents have no immediate impact on public health and safety and the bolting incidents so far have not resulted in accidents. Degradation or failure of such studs b and bolts constitutes a reduction in the integrity of the primary pressure boundary. Concern is compounded- by the fact that there is currently no reliable NDE method to detect the cracking or degradation of such bolts or studs resulting from the principal modes of failure which are stress O corrosion, fatigue, erosion corrosion, and boric acid corrosion.

Visual Examination is curre'ntly the only reliable method to discover degradation by boric acid or erosion corrosion. In almost all cases this i requires disassembly of the component in order to inspect the bolts or studs. If there is not clear evidence of boric acid leakage to the 3

surroundings, bolting degradation by boric acid corrosion can potentially be undetected until the bolts or studs completely fail. Under the present inservice inspection program, visual inspection of bolts is not a manda-tory requirement and UT inspection is not required on pressure retaining bolts or studs with diameters less than 2 inches. A major accident such as a LOCA could conceivably occur due to undetected extensive bolting

. failure of the primary pressure boundary.

Issue 29 had previously been limited in scope to primary pressure boundary bolting. However, when the results of broader industry initiatives were reviewed, this issu'e was expanded to address as large a scope of safety related bolting as can be justified on the basis of regulatory analysis.

Current plans call for the staff to incorporate all safety related bolting under GI-29. Review is underway of the regulatory analysis which will lead to revisions to the SRP and for some type of surveillance program.

CRGR review is planned for September P 38 with final implementation planned for the March 1990 timeframe.

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SP/90 Response l The objective of the SP/90 design is to minimize the n mber of bolts.

Westinghouse will follow this issue and consider any recommendations which 3

result from this effort and factor them into the SP/90 design as

, appropriate.

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30. Issue 30: Potential Generator Missiles - Generator Rotor 3 This issue has been dropped and is no longer under review.

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31. Issue 31: Natural Circulation Cooldown Discussion This issue arose as a result of an incident that occurred at an operating pressurized water reactor a few years ago. While operating at full power on 6/11/80, one of the two containment isolation valves in the component cooling water (CCW) return line from the reactor coolant pumps (RCPs) at Saint Lucie failed closed causing a simultaneous loss of component cooling water to all reactor coolant pumps.

LP/90 Response This issue, which was subsumed into I.C.1 of the TMI Action Plan and its 3 impact on the SP/90 design, is fully discussed in NRC Generic Letter 81-21 (see Section 6.4, Item 21).

32. Issue 32: Flow Blockage in Essential Equipment Caused by Corbicula This issue was subsumed into "New" Generic Issue 51, discussed later in 3

this section.

O WAPWR-RC 5.5-28 AMENDMENT 3 BB8Be:1d AUGUST 1989

, 33. Issue 133L Connecting Atmospheric Dump. Valve Opening Upon. Loss of-

. Integrated Control System Power

.This issue was subsumed into USI A-47, Section 4.0 of this module.- 3

34. Issue'34: Reactor' Coolant System Leak 3

This issue has been dropped and is no longer under review.

35. Issue 35: . Degradation of Internal Appurtenances in LWRs Discussion  ;

This issue deals with loose parts in the primary system. From time .to l time, loose parts have .been transported through a portion of the primary side' system only to become lodged in some unidentified location before causing any damage. In the event of .a steamline break, the resulting pressure transient in the primary side could cause a loose part to become dislodged. travel to the steam generator and cause a small break LOCA.

Internal ' appurtenances such as flow straighteners, orifices, thermal-sleeves, screens, etc. have the potential to break loose and become " loose

~

. parts" in the fluid system.

This issue has been assigned.a low priority by the NRC. 3 SP/90 Response As this issue evolves, Westinghouse will consider and factor into the SP/90 design any NRC recommendations which result, as deemed appropriate.

.36. Issue 36: Loss of Service Water O Discussion Calvert Cliffs Unit 1, experienced a loss of both redundant trains of service water when the system became air-bound as a result of the failure WAPWR-RC 5.5-29 AMENDNENT 3 3888e:1d AUGUST 1989

4 1

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of a non-safety-related instrument air compressor aftercooler. The j significance 'of this event lies in the fact that it involved two l l

fundamental aspects in the design of safety-related systems: (1) interac-tion between safety and non-safety-related systems and components, and (2) common cause failure of redundant safety systems.

All but one generic concern and one plant-specific matter raised by the AE00 case study on the Calvert Cliffs loss of service water have been or will be adequately addressed as part of USI A-45 or Issue 67.5.2. The generic concern was resolved with the issuance of SRP Sections 9.2.1, Rev. 4, and 9.2.2, Rev. 3, in June 1986.

These revisions did not incorporate any new guidelines or requirements. The remaining plant-specific matter concerning Calvert Cliffs has been brought to the attention of the DL for appropriate action. Thus, this issue has been RESOLVED and no new requirements were established.

SP/90 Response The essential service water system (ESWS) of the SP/90 plant will be designed to be independent of non-safety related systems such as instrument air. Nevertheless, common cause failure of the ESWS has been considered in the design of the plant in that the backup seel injection pumps and one of the two turbine-driven emergency feedwater pumps will be 3

able to maintain the plant in a safe condition for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. These features are independent of offsite and onsite AC power, as well as of the essential service and component cooling water systems. In essence, therefore, these features provide protection against both station blackout and loss-of-cooling scenarios.

37. Issue 37: Steam Generator Overfill and Combined Primary and Secondary Blowdown Discussion I

This issue was subsumed into USI A-47, which is addressed in Section 4.0 l I

3 of this module. The issue of steam generator overfill"is discussed in NRC i Generic Letter 81-28. (Section 6.4, Item 29).

WAPWR-PC 5.5-30 AMENDMENT 3 B838e:1d AUGUST 1989 I

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38. Issue 38: Potential Recirculation System Failure as a Consequence of Injection of Containment Paint Flakes or Other Fire Debris This issue has not been defined or prioritized by the NRC. However, this issue may be'related to Section 4.0, item 21.

39. Issue 39: Potential for Unacceptable Interaction Between the Control Rod Drive System and Nonessential Control Air System This issue is not applicable to Westinghouse pressurized water reactor I designs. .

40. Issue 40: Breaks in the BWR Scram System l

This issue is not applicable to Westinghouse pressurized water reactor designs.

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41. Issue 41: BWR Scram Discharge Volume Systems

(

This issue is not applicable to Westinghouse pressurized water reactor designs.

42. Issue 42: Combination Primary / Secondary LOCA This issue was subsumed into Generic Issue 18 and I.C.1 of post-TMI 3 requirements.

43. Issue 43: Contamination of Instrument Air Lines Discussion This issue has been initiated in response to regarding dessicant contamination of instrument air lines. This concern was prompted by an 3 incident at Rancho Seco in which the slow closure o'f a containment .

WAPWR-RC 5.5-31 AMENDMENT 3 588Be:1d AUGUST 1989 w _ _ _ _ _ _ _ _ - . _ _ _ _ _ _ _ _ _ - _ _ _ .

isolation valve resulted from the presence of desiccant particles in the valve operator. Desiccant cc Tiination of the plant instrument air i system (IAS) was also found to bu ;ne of the contributing causes of the f loss of the salt water cooling system at San Onofre in March,1980.

The principle concern of this issue is that contamination of instrument air lines can result and has resulted in reactor transients and scrams.

Compressed air may be contaminated from several sources including: (1) the ambient air, (2) the compressor itself, (3) drying equipment, and (4) corrosion products in the piping systems. Thus, compressed air must be cleaned and dried for many applications.

h A compressed air system is provided for instrumentation and valve operators, both of which are required for plant control. The function of the compressed air system is to reliably provide required air of suitable quality and pressure for instruments, controls, maintenance, and general power plant uses and operations.

3 The compressed air system is generally divided into two subsystems: the ser- 'ce air system (SAS)andtheIAS. The compressed gas system (air and nitrogen) is not classified ac safety grade except for those portions of the distribution system that penetrate the containment. In some cases, a

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separate and independent system called the containment instrument air system (CIAS) is located entirely within the containment structure to preclude any pressurization of the containment structure.

The IAS is designed so that the instrument air shall be available under (

l all normal and abnermal operating conditions. The SAS is designed to back j up the IAS during abnormal unit operations and has in a few instances been shown to introduce contaminants to the IAS when celled upon as a backup system due to lower que !ity air requirements of the SAS. All essential I systems requiring air during or after an accident are self-supporting, and g after an accident the air system is reestablished. W 1

Ol WAPWR-RC 5.5-32 AMENDMENT 3 5888e:1d AUGUST 1989

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1 Operation of the IAS is' not required to initiate operation of engineered safeguards equipment. However, scenarios can be developed where, after the storage accumulators are exhausted, failure of the IAS can be shown to i influence performance of equipment in other service groups which,. after I their subsequent failure, can then adversely affect the performance of yet other equipment in engineered safeguards systems. The probability. of such .

a common cause failure happening is very low. j 1

'l In December 1987, the staff issued Information Notice- No. 87-28, Supplement 1, to inform OLs and cps of the publication of NUREG-1275, Volume 2. This report indicated that the performance of the air-operated safety-related components may not be in accordance with their intended safety function because of inadequacies in the design, installation, and 3

maintenance of the instrument air system. The report also indicated that anticipated transient and system recovery procedures were frequently inadequate and that operators were not well-trained for coping with loss of instrument air conditions.

In August 1988, Generic Letter 88-14 was issued to request that each l licensee / applicant review NUREG-1275. Volume 2, and perform a design and operations verification of instrument air systems. In addition, all licensees / applicants were requested to provide a discussion of their program for maintaining proper instrument air quality. Thus, this issue was RESOLVED and requirements were established.

SP/90 Response q The SP/90 air s'ystems, including the instrument air system, are control

(/ grade systems whose availability is not required for operation of any of the safety related systems. Additionally, the design of these systems is the responsibility of each plant specific applicant and should be reviewed 3 against SRP Section 9.3.1, " Compressed Air' in each applicant's FDA.

O i

O WAPWR-RC 5.5-33 AMENDMENT 3 i 8888e:1d AUGUST 1989

l Failure of the Instrument Air Systems may cause a transient to develop; however, such transients are no more severe than those normally analyzed and are, therefore, enveloped by existing analyses.

3 The concerns addressed in Generic Safety Issue 43 are not applicable to the SP/93 plant. At the FDA stage, an evaluation will be perforc,ed to demonstrate that air system failures are enveloped by existing analyses.

h

44. Issue 44: Failure of Saltwater Cooling System l 3 This issue was subsumed into Generic Issue 43 above.

l

45. Issue 45: Inoperability of Instrumentation Due to Extreme Cold Weather Discussion i

This issue involves an assessment of the measures taken to protect  !

instrumentation from severe weather and to verify the condition and opera- l bility of heat tracing systems and other measures taken to protect plant equipment from severe weather.

Bulletin No. 79-24, regarding frozen lines, was issued on September 27, 1979. Issuance of the bulletin was prompted by LERs which had revealed many events involving frozen instrument, sampling, and processing lines.

All licensees and CP holders were requested to " review their plants to determine the adequate protective measures have been taken to assure that safety related process, instrument and sampling lines do not freeze during extremely cold weather."

O At Arkansas Nuclear.One, Unit 2, all four RWST instrumentation channels were lost when the level transmitters froze. The system heat-tracing circuit was de-energized because the main line fuse was removed. This i situatior. would have prevented the automatic change over of the ECC from l injection to recirculation mode under LOCA conditions (i.e.,

loss-of-safetyfunction).

Ol WAPWR-RC 5.5-34 AMENDMENT 3 5838e:Id AUGUST 1989 l

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4 In an August 14, 1981 memorandum to AE00, NRR advised that a BTP on freeze s protection of safety related instrument lines was being -developed and j would be included in the appropriate SRP Section following its review and approval. NRR further advised that DIE proposed to amend the Inspection

-and Enforcement- Manual to include a module which would ' set forth requirements fcr inspection of systems and measures for-protection against cold weather.- This inspection module would require that regional inspectors perform plant site visits prior to the beginning of the cold season to verify the condition of heat-tracing systems and measures taken to protect plant equipment from cold weather conditions. An amendment to the Inspection and Enf.orcement Manual (Procedure No. 71714) was issued by OIE on January 1, 1982, thus completing the OIE portion of the resolution of this issue. Acceptance criteria for the design of protective measures against freezing in instrument lines of safety-related systems were included in Draft _ Regulatory Guide 1.151, " Instrument Sensing Lines."

With inclusion of the criteria in the Draft Regulatory Guide, further work on a BTP was terminated. The Draft Regulatory Guide was issued for comment ~in . March, 1982. Comments were collected and dispositioned and the Regulatory Guide was published in July, 1983. Notice of the issuance of

-Regulatory Guide 1.151 was published in the Federal Register on August 8, 1983. Implementation of the Guide is limited to all cps issued after September 1, 1983. However, other licensees or applicants may adopt the

~

use of the Guide on a voluntary basis. As atated in the value/ impact statement for the Guide, no backfitting of requirements for freeze prota.: tion and alarms is to be accomplished other than those changes effected by IE Bulletin 79-24 (an'd the inspection requirements added to the OIE Inspection Manual).  :

l O In February,1984, the following SRP Sections were revised to incorporate i

q the changes associated with the resolution of this issue: (1)Section 7.1, Rev. 3; (2) Section 7.1, Appendix A. Rev. 1; (3)Section7.5, Rev. 3; )

and (4) Section 7.7, Rev. 3.

This issue has been RESOLVED and requirements were issued.

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WAPWR-RC 5.5-35 AMENDNENT 3 8888e:1d AUGUST 1989 t'

4 SP/90 Response The SP/90 design does not include any safety related equipment which is located outside the Seismic Category 1 buildings. Specifically, the 3 expergency water storage tank (EWST) and the emergency feedwater storage tanks '(EFWST) are located in Seismic Category 1 structures, and freezing of any instrumentation associated with these tanks is thereby precluded.

46. Issue 46: Loss of 125 Volt D.C. Bus Discussion 3

This issue has been subsumed into Generic Issue 76 which has not been prioritized by the NRC.

47. Issue 47: Loss of Offsite Power Although this particular issue has not been defined by the NRC, it has 3 been integrated into Unresolved Safety Issue A-44, " Station Blackout" (Section4.0, Item 22).

h

,48. Issue 48: LCO for Class IE Vital Instrument Buses in Operating Reactors This issue has been subsumed into Generic Issue 128 which is addressed 3

later in this section.

49. Issue 49: Interlocks and LCOs for Redundant Class IE Tie Breakers This issue has been subsumed into Generic Issue 128 which is addressed 3

later in this section.

50. Issue 50: Reactor Vessel Level Instrumentation in BWRs g i This issue is not applicable to Westinghouse pressurized water reactor designs.

i WAPWR-RC 5.5-36 AMENDMENT 3 1 E888e:1d AUGUST 1989 1

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51. Issue 51: Proposed Requirements for Improving Reliability of Open Cycle i Service Water System Discussion O This issue addresses the subject of service water system (SWS) fouling at operating plants primarily by aquatic bivalves. The following is a summary of reported events of serious fouling in open cycle water systems:

O 1. Arkansas Nuclear One, Unit 1 (ANO-1) failed a technical specification surveillance test of a containment fan cooler unit due to buildup of Asiatic clams (corbicula).

2. Brunswick 1 and 2 reported that 3 of the 4 RHR heat exchangers had experienced baffic plate displacement due to a buildup of oysters.

3. Pilgrim reported that the baffle plate of a component cooling water heat exchanger was displaced by a buildup of mussels (mytilus).

3

4. San Onofre 1 reported that a buildup of barnacles prevented proper

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cooling of a component cooling water heat exchanger.

5. Rancho Seco reported that a buildup of corrosion products prevented proper cooling of a diesel generator lubc oil cooler.

l

6. Sequoyah Unit I reported flow blockage in the emergency raw cooling' water system due to Asiatic clams.

I As a result of the NRC concern for the effects on safety of open cycle water system fouling, IE Bulletin 81-03 was issued. Responses to this O bulletin revealed that bivalves were observed at approximately 45% of all sites.

O WAPWR-RC 5.5-37 AMENDMENT 3 B888e:1d AUGUST 1989

4 The following related issues have been combined with the issue of whether or not the staff should develop requirements for improving the reliability of open cycle water systems: Issue 32, " Flow Blockage in Essential Equipment Caused by Corbicula," and Issue 52, "SSW Flow Blockage by Blue Mussels."

The SWS is the ultimate heat sink that, during an accident or transient,

• l cools the reactor building component cooling water hoat exchangers, which '

in turn cool the RHR heat exchangers as well as provide cooling for ,

safety-related pumps and area cooling coils. Fouling of the )

safety-related SWS either by mud, silt, corrosion products, or aquatic )

bivalves has led to plant 'hutdowns, s reduced power operation for repairs, j and modifications and degraded modes of operation. j 3

. I The scope of this task will be to evaluate the adequacy of surveillance

]

and control methods and to propose the use of those methods which will l assure the continued reliability of open cycle service water systems against various forms of fouling.

SP/90 Response The service water system is plant specific and outside the scope of the SP/90 design. However, Westinghouse has provided a brief description of the required functions and required components of the SWS in Subsection l 9.2.1 of RESAR-SP/90 PDA Module 13, " Auxiliary Systems." Additionally, I

this section provides design critoria required to assure the SWS is compatible with the systems included in Westinghouse's nuclear power block scope. Surveillance and control methods to ensure the reliability of open cycle service water systems will be established by the utility.

52. Issue 52: SWS Flow Blockage by Blue Mussels 3 This issue hac been subsumed into Generic Issue 51 above.

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WAPWR-RC 5.5-38 AMENDMENT 3 5888e:1d AUGUST 1989 i

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53. Issue 53: Consequences of a Postulated Flow- Blockage Incident in' a Bolling Water Reactor This issue was not applicable to Westinghouse pressurized water reactor 3

designs, but has been dropped and is no longer under review.

-54. Issue 54:

Survey of Valve Operator Related Events Occurring During 1978,

, 1979, and 1980 3

This issue has been subsumed into TNI Action Plan item II.E.6.1.

1

55. Issue 55: Failure of Class IE Safety Related Switchgear Circuit -Breakers to Close on Demand Discussion This issue has been dropped and is no longer under review. 3

56. Issue 56: An Analysis of the Abnormal Transient Operating Guidelines Discussion This issue has been subsumed into USI A-45 and THI Action Plan Item I.D.1, 3

57. Issue 57: Effects of Fire Protection System Actuation on Safety Related Equipment Discussion -

This issue is concer.ned with fire protection system (FPS) actuations which have resulted in adverse interactions with safety-related equipment at operating nuclear power plants. Events have shown that safety-related 0

equipment subjected to FPS water spray enuld be rendered inoperable. The events also indicated numerous spurious actuations of the FPS initiated by operator testir.g errors or by maintenance activities (e.g., welding),

steam, or high humidity in the vicinity of FPS detectors.

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WAPWR-RO 5.5-39 AMENDMENT 3 8888e:1d AUGUST 1989 L _ - - _ _ . - . . - _ - - - _ . - - ._

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On June 22, 1983, IE Information Notice 83-41 was issued to alert  :

licensees and provided examples of recent experiences in which actuation of fire suppression systems caused damage or i. operability of systems l important to safety. The IE Notice indicated that the plant Fire Hazards Analysis required by Appendix R to 10CFR50 and by the related NRR BTP requires, not only consideration of the consequ3nces of a postulated fire, but also consideration of the effects of fire-fighting activities. The IE Notice stated that a properly conducted Fire Hazards Analysis in conjunc-tion with a ph sical walk-down of plant areas would have identified l instances where minor modifications such as shielding equipment and sealing conduit ends suld have reduced equipment water damago fron h

inadvertent FPS operat'ix . The IE Notice indicated that none of the reported events resultad in a serious impact on the functional cepability of a plant to protect public health and safety. However, examples were given where it would not be difficult to extrapolate actual occurrences into a sequence of events that could lead to more serious consequences.

FPS actuations which result in adverse interaction with plant safety systems reduce the availability of such safety systems needed to achieve p . safe plant shutdown or to mitigate a postulated accident.

3 A possible solution is to follow up IE Information Notice 83-41 with an IE Bulletin which would require licensees to reevaluate their implementation of the FPS system guidelines regarding adverse interactions with safety systems to assure that safety-related equipment, not damaged by fire itself, can perform its intended function during and following an FPS actuation.

There already exists a mechanism for mandatory review of the Fire Protection Program and implementing procedures for each plar.t operating without requiring an IE Bulletin. The Technical Specifications for each plant requires under Administrative Controls:

a. An independent fire protectiot. and loss prevention inspection and audit annually utilizing either qualified offsite licensee personnel  ;

or an outside protection firm.  !

WAPWR-RC 5.5-40 AMENDMENT 3  ;

588Be:1d AUGUST 1989  !

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.; Ib. Audit > of' tho' Fire Program and implonanting procedures at least once

)

, per 24 months. ]

c. An inspection and audit of the fire protection and -loss prevention

< program by an outside' qualified fire consultant at intervals no

! greater-than 3 years.

The issuance of 'IE. Notice 83-41 and periodic .INPO Significant Event Reports provide ti m ly operations feedback to enable licensees to consider ,

-applicability of such avsnts in the periodic program reviews indicated I above. Several plants have documented -instances where their periodic 3 l reviews in conjunction with the feedback information identified an area where there was potential for FPS interaction with part of a safety system {

and corrective action was taken.

The NRC has assigned a MEDIUM priority to this issue.

SP/90 Response O

d' It is the intent of the SP/90 design to preclude water damage to safety related equipment from inadvertent operation of the fire protection system to the extent possible. This issue will be addressed in detail in the SP/90 FDA application.

58. Issue 58: Inadvertent Containment Flooding This issue has been dropped and is no longer under review. 3

59. Issue 59: Technical Specification Requirements for Plant Shutdown When Equipment for Safety Shutdown is Degraded or Inoperable Discussion This issue is concerned with equipment failure resulting in impairment of the capability to take the plant to a shutdown condition where the Tech-nical Specifications required that the plant be shutdown in a short time period.

I WAPWR-RC 5.5-41 AMENDMENT 3 B888e:1d AUGUST 1989

The NRC has concluded this is a Regulatory Impact issue to be addressed by  ;

3 the Technical Specification Improvement Project. g .

60. Issue 60: Lamellar Tearing of Reactor Systems Structural Supports l Lamellar tearing results in almost all cases from limitations in steel plate introduced during manufacture.

gj I l

This issue has been subsumed into Unrasolved Safety Issue (USI) A-12 3

(Section4.0, Item 12). ,

4

61. Issue 61: SRV Line Break Inside the BWR Wetwell Airspace of Mark I and II Containments This issue is not applicable to Westinghouse pressurized water reactor designs.

62. Issue 62: Reactor System Bolting Applications Discussion This issue was raised in November 1981 and was based on the concern that 3

NRC provides no control regulations or guides for bolting other than for the reactor vessel head.

In December 1981, Issue 29, " Bolting Degradation or Failure in Nuclear l Power Plants," was determined to be of high priority. Resolution of  ;

Issue 29 will include all safety-related bolting, with emphasis on reactor coolant pressure boundary bolting degradation or failure. Thus, Issue 29 was broadened to cover Issue 62. The tasks in Issue 29 include:

h (1) development of t'he technical bases for bolting application require-i ments; (2) review of licensee responses to IE Bulletin 82-02; (3) draft of a staff recommendation for proposed criteria / guidelines to be incorporated into the SRP; and (4) development of a proposed implementation plan for h

management consideration. Therefore, this issue was integrated into the resolution of Issue 29. )

O WAPWR-RC 5.5-42 AMENDMENT 3 588Be:1d AUGUST 1989 l

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.63. Issue 63: Use of Equipment Not Classified.as Essential to Safety in BWR G Transient'Anr. lyses 1

This issue is not applicable to Westinghouse pressurized water reactor j designs.

64. Issue 64: Identification of Protection System Instrument Sensing Lines Discussion v.

This issue has not been defined or prioritized by the NRC.

65. Issue 65: Probability of Core Melt Due to Component Cooling Water System Failures l

Discussion This issue has been subsumed into Generic Issue 23 addressed earlier in 3

this section.

66. Issue 66: Steam Generator Requirements Discussion The issue was established to assess proposed generic requirements stemming from USI's A-3, A-4, and A-5 regarding steam generator tube integrity.

O Following the steam generator tube rupture (SGTR) event at Ginna in U January 1982, the staff proceeded to develop generic steam generator requirements which. would help mitigate or reduce steam generator tube 3 degradations and ruptures.

t As approved by the commission, originally proposed generic requirements were issued as recommendations to all PWR licensess and applicants in GL 85-02.

WAPWR-RC 5.5-43 AMENDMENT 3 5888e:Id AUGUST 1989 1

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+ l Given that the risk to the public from a SGTR is not high, the proposed .

staff requirements were assessed for potential reductions in occupational radiological exposure, potential reductions in SGTR frequency and tube degradation, and potential reduction frequency in forced plant outages. l The staff's assessment of licensee respenses to GL 85-02 was provided to the commission in SECY-86-77, dated 03/24/86.

Completion of issue 66 will include issuing plant-specific letters describing staff assessment of response to GL 85-02 and revising SRP and

, STS to be consistent with GL 85-02.

3 SP/90 Response The proposed generic steam generator requirements found in Generic Safety Issue 66 are essentially the same as the requirements proposed for resolu-tion of Unresolved Safety Issue A-3, A-4, and A-5. Severe 1 of these requirements were dropped or determined to require additional Staff action. The remainder were classified as Staff Recommended Actions. The disposition of the proposed steam generator generic requirements was documented in NUREG-0844 and the Staff Recommended Actions also issued as Generic Letter 85-02. The manner in which the SP/90 design addresses these recommendations is found in the discussion of USI A-3.

67. Issue 67: Steam Generator Staff Actions Discussion Following a SGTR event at Ginna on January 25, 1982, increased staff effort was placed on developing means to mitigate and reduce steam genera-tor tube degradations and ruptures. To meet these objectives, a dual 3 approach was taken. The first approach was to develop staff require-ments to be implemented by the licensees. The proposed staff requirements are evaluated in Issue 66. In addition to these proposed requirements, the staff identified and recommended certain staff actions.

O WAPWR-RC 5.5-44 AMENDMENT 3 5888e:1d AUGUST 1989

4 The' issue has been separated into sub-items as listed in Table- 3.67-1 of-NUREG-0933, "A prioritization of Generic Safety Issues."- Many of the sub-issues have been dropped, subsumed or-. identified as NRC tasks. The only sub-item which requires resolution'by Westinghouse is 67.03.3.

ITEM 67.03.3: Improved Accident Monitoring This issue calls. for the staff to address the accident monitoring weaknesses of the type observed at Ginna by implementation of Regulatory Guide 1.97 and the Safety Parameter Display System-(SPDS).

During the event at Ginna, several weaknesses in accident monitoring were apparent. These include: (1)non-redundant monitoring of RCS pressure;.

(2) failure of the position indication for the steam generator relief and safety valves; and (3) the limited range of the charging pump flow indicator for monitoring charging flow during accidents. These conditions make it more difficult for correct operator action in response to such events. 3 A

U Had. Regulatory Guide 1.97 been implemented at Ginna before the January 1982 event, the monitoring of. the event would have been substantially improved and there would have been more assurance of correct operator actions. Improved accident monitoring would also have improved the NRC's ability .to assess the plant status and. the appropriateness of the licensee's actions and recommendations.

The above recommendation was issued as resolution of this sub-item by Supplement 1 to NUREG-0737 (Gensric letter No. 82-33).

SP/90 Response The SP/90 control room design, as discussed in Section 18.0 of RESAR-SP/90 l Os PDA Module 15 " Control Room / Human Factors Engineering," includes a human factors engineering process that results in an improved man-machine O

WAPWR-RC 5.5-45 AMENDMENT 3 5888e:1d AUGUST 1989

interface to maximize the rel.iability of the human / automatic roles in preventing and mitigating against the effects of all transient or accident conditions. ,

l i

The SP/90 alarm system, described in Subsection 18.3.2 of RESAR-SP/90 PDA l Module 15, includes the attributes of a safety parameter display system (3PDS) as required by NUREG-0696. The design meets the requirements of Regulatory Guide 1.97.

68. Issue 68: Postulated Loss of Auxiliary Feedwater System Resulting from Turbine-Driven Auxiliary Feedwater Pump Steam Supply Line Rupture l

This issue was subsumed into Item 124 discussed later in this section, j 69. Issue 69: Make-up Nozzle Cracking in B&W Plants l

3 This issue is not applicable to the Westinghouse pressurized water reactor designs.

g

70. Issue 70: PORV and Block Valve Reliability Discussion This issue involves assessing the need for improving the reliability of PORV and block valves.

Both the PORVs and block valves were originally designed as non-safety components in the reactor pressure control system for use only when at power operation. The block valves were installed because of expected leakage from the PORVs. Neither the PORVs nor the block valves were required to safely shut down the plant or mitigate the consequences of g accidents. However, it has recently been determined that PORVs are, in W fact, relied upon to mitigate a design-basis Steam Generator Tube Rupture (SGTR). The acceptability of relying on non safety grade PORVs to mitigate a design-basis accident, SGTR, is of concern.

WAPWR-RC 5.5-46 AMENDMENT 3 5888e:1d AUGUST 1989

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In most plants, the low temperature overpressure protection (LTOP) system is designed to use the PORVs. For this mode of operatier., the valves are  ;

typically set to open at 500 psig rather than the high pressure (~2300 psig)setpointusedatpower. LTOP systems, as specified in SRP Sectio. 1 5.2.2, are to be single failure proof, testable, designed to quality standards, and operable from emergency power. Full implementation of IEEE Standard-279 to withstand an SSE is not specified, but the OBE is.

(

NUREG-0737, Clarification of TNI Action Plan Requirements, Item II.D.1,

\ set forth functional requirements for both PORVs and block valves. All plants were required' to demonstrate the functionability of these valves for all expected flow conditions during operating and accident j conditions. It was further required that the block valves be capable of closing to ensure that a stuck open relief valve can be isolated, thereby terminating a small loss of coolant accident. ,

When PORVs are used for high point vents in some plants under Item II.B.1 3

of NUREG-0737, both PORVs and block valves are required to meet seismic and environmental requirements for safety-related equipment.

There was, and still is, no Technical Specification requirements that

- these components be operational when the plant is at power. Continued operation at power with inoperable PORVs and block valves is permitted by the TS if the block valve is closed and power to the block valve (s) is removed.

It is noted that the NRC safety evaluations for Item II.K.3(2) determined that an automatic PDRV isolation system is not necessary. However, the possible need to improve PORY reliability was recognized.

PORVS and block valves are used in various modes of plant operation. When these valves have been demanded to operate, they have stuck open on a I number of occasions. Such malfunctions have led to significant plant transients and aggravated others in the past.

fs WAPWR-RC 5.5-47 AMENDMENT 3 5888e:Id AUGUST 1989

4 In other cases, when the PORVs have leakage problems and the PORVs are blocked, this could cause the safety valves to be challenged and, if they stick open, could result in an unisolable SBLOCA.

SP/90 Response The SP/90 plant includes safety grade pressurizer PORVs (power operated relief valves) and block valves. Their primary function is partial depressurization of the Reactor Coolant System during one of the following events.

o Safety grade cold shutdown o Steam generator tube rupture o Feed and bleed operation Operation of both the PORVs and the block valves in these cases is under manual control. The PORVs can also be used to depressurize the RCS during a degraded core scenario in order to prevent failure of the primary boundary at high pressure; for this reason, the power supplies for the PORVs are independent of both off-site and emergency on-site AC power supplies.

The PORVs are also used to limit RCS pressure for all design load reduction transients up to and including a full load rejection. This mode of operation is fully automatic; as a safeguard against spurious operation of the PORVs, coincident high pressure signals derived from any two of the four pressurizer pressure transmitters are required to open these valves.

When the PORVs'are in this automatic mode of operation, the block valves will close automatically if the PORVs fail to close when the low pressure g

setpoint is reached. Note that this automatic mode will cause the PORVs to open during a loss-of-load or ATWS scenario. )

The PORVs are not utilized for low pressure temperature overpressure protection (LTOP); this function will be performed by the safety valves in j the residual heat removal (RHR) system. l O !

l 1 WAPWR-RC 5.5-48 AMENDMENT 3 B888e:1d f AUGUST 1989 I

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, ('T Positive position indication (open or closed) is provided in the main control room for the PORVs; temperatures in the PORV discharge lines are measured and alarmed in the main control room. )

4

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V Finally, the PORVs and block valves will be qualified to function under all expected flor conditions.

As stated in our March 1989 response to Staff Question 730.1, the p following commitments, which for the most part are currently reflected in V the PDA, will be included in the SP/90 FDA submittal:

o There will be a minimum of two PORVs and block valves including redundant and diverse control systems.

e The PORVs and block valves will be designed to Safety Grade, Seismic Category 1 requirements. 3 o The design will allow for environmental qualification of the PORVs and i block valves.

o The design of the PORVs and block valves will allow inservice testing in accordance with Article IWV-3400 of ASME Section XI.

o The PORVs and block valves will be included in the final safety related 0-lirt of Table 3.2-1 of the RESAR-SP/90 FDA.

o Surveillance requirements will be included in the FDA Standard Technical Specifications to ensure PORV and block valve operability.

The SP/90 plant adequately addresses the concerns of Generic Safety Issue 70.

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WAPWR-RC 5.5-49 AMENDMENT 3 E88Be:1d AUGUST 1989

I s

71. Issue 71: Failure of Resin Domineralizer Systems and their Effects on Nuclear Power Plant Safety l

Discussion I I

This issue arises from a need to provide additionsl a licensing attention for certain ancillary power plant equipment. Available information showed l that failures of resin bed type domineralizer sub-systems have occurred I within the process systens (both nuclear and non-nuclear) of nuclear power facilities. These process systems, by definition, do not directly perform any reactor protection or engineered safeguards features functions, yet their failure has seriously impaired the capability of these systems to perform by rendering their redundant trains inoperable.

SP/90 Response This issue has not been further defined or prioritized by the NRC.

3

72. Issue 72: Control Rod Drive Guide Tube Support Pin Failures g, Discussion This issue involves the failures of the cupport pins that are attached to the bottom of the CRD guide tubes in reactors designed by Westinghouse.

The support pins align the bottom of the CRD guide tube assembly with the top of the upper core plate in a manner that provides lateral support and accommodates thermal expansion of the guide tube relative to the core plate.

Some of the safety concerns include failure to SCRAM, SCRAM system performance during design-basis accidents, and potential damage to safety systems and cc,mponents due to loose parts in the reactor coolant system.

O WAPWR-RC 5.5-50 AMENDMENT 3 EB8Be:Id AUGUST 1989

L SP/90 Response hO u This issue has not been further defined or prioritized by the NRC.

73. Issue 73: Detached Thermal Sleeves Discussion The issue involves the discovery of metal pieces at the bottom of reactor

/ vessels in B&W and Westinghouse reactors. At Trojan nuclear plant, further inspection revealed another metal fragment between the lower core plate and the core 'suppor't plate. These pieces were subsequently identified as thermal sleeve pieces initially installed in the safety injection accumulator piping r.ozzle connections to the reactor coolant system cold leg piping. Confinnation that the 10-inch thermal sleeves were missing from the four safety injection piping nozzle connections was obtained shortly thereafter.

Fatigue failure problems connected with nozzle-thermal sleeve assemblies Os have been identified in piping systems of both BWRs and PWRs.

. 3 SP/90 Response This issue has not been further defined or prioritized by the NRC.

74. Issue 74: Reactor Coolant Activity Limits for Operating Reactors 1

This issue has been dropped and is no longer under review.

O 75. Issue 75: Generic Implications of ATWS Events at the Salem Nuclear Plant - QA O  !

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O l WAPWR-RC 5.5-51 AMENDMENT 3 5888e:1d AUGUST 1989

Discussion Failure to scrum (also commonly referred to as ATWS) could rupture the RCS or distort ECCS valves- such that a core-melt would result (see USI A-9, "ATWS").

On two occasions (February 22 and 25, 1983), Salem Unit 1 failed to scram automatically due to ~ failure of both reactor trip breakers to open on receipt of an actuation signal. In both cases, the unit was successfully tripped by manual action. The failure of the breakers has been attributed to excF'a wear due to improper maintenance of the undervoltage relays which receive the trip signal from the protection system and cause mechanical action to open the breakers.

Three sepcrate NRC actions were initiated to address this problem. One was plant specific and was to address those things which needed to be done in order for Salam to restart. This was completed in the " Salem Restart Evaluation." The second action was an investigation into the Salem events and the circumstances leading to them. This was reported in NUREG-0977. g 3 The third action was the formation of an NRC task force to study the W overall generic implications of this event. The Salem Generic Implications Task Force published NUREG-1000 Volume 1, " Generic Implications of ATWS Event at the Salem Nuclear Power Plant, which provided their findings.

The required actions for licensees and applicants as well as the internal NRC staff actions were published in Aigust 1983 as NUREG-1000, Volume 2. In addition, a number of issues have bein raised within NRC which are closely related to the design and testing of the reactor protection system.

The resolution of quality assurance concerns related to the Salem ATWS O

events is being addressed in a revision to Regulatory Guide 1.33 " Quality Assurance Program Requirements (Operations)" currently under preparation.

The Regulatory Guide revision under consideration will contain detailed guidance for operational QA programs.

more h

O WAPWR-RC 5.5-52 AMENDMENT 3 5888e:1d AUGUST 1989

il SP/90 Response Y  ;

Quality Assurance Program Requirements for operations are not within the scope .of the SP/90 design effort. Therefore this issue is not applicable

, to the current SP/90 design.

76. Issue 76: Instrumentation and Control Power Interactions Discussion This issue involves a number of concerns regarding DC power systems that were raised in a memorandum which provided comments on the proposed resolution of Item A-30, " Adequacy of Safety-Related DC Power Supplies."

The main concerns were presented as follows:

(1) A control and instrumentation power supply fault can cause a 3

critical challenge to standby ESFs, i.e., cases including trips, loss of main feedwater, loss of offsite power, and/or small LOCA.

(q/

(2)The same C&I power supply fault could defeat some of the ESFs

~

called upon to mitigate the initiating event, both core cooling

- systems and containment cooling systems.

(3)The same C&I power supply fault could blind or partially blind the operators to the status of the plant.

The memorandum in which the concerns were raised does acknowledge that there are existing programs which consider and address part of the concerns. However, it is concluded that there is no one program that appears to address all of the aspects of the possible significance to safety of accident sequences as precipitated by instrumentation and control power supply failures.

O l WAPWR-RC 5.5-53 AMENDMENT 3 588Be:1d AUGUST 1989

)

SP/90 Response This issue has not been further defined or prioritized by the NRC.

77. Issue 77: Flooding of Safety Equipment Compartments by Backflow Through g' Floor Drains W  !

This issue was subsumed into USI A-17 addressed in Section 4 of this module.

i

78. Issue 78: Nonitoring of Fatigue Transient Limits for Reactor Coolant System Discussion l

This issue originated as MPA-B-70. However, a letter to licensees was l 3 never issued for the collection of information.

CONCLUSION This issue has not been further defined or prioritized by the NRC.

79. Issue 79: Unanalyzed Reactor Vessel Thermal Stress During Natural Convection Cooldown Discussion This issue addresses a concern of potential generic safety significance relating to an unanalyzed reactor vessel thermal stress that could occur g

during natural convection cooldown of PWR reactors. The concern emerged from a preliminary B&W evaluation of the voiding event that had occurred in the upper head of the St. Lucie reactor on June 11, 1980. Based on several conservative assumptions, B&W tentatively concluded that during natural convection cooling there could develop axial temperature gradients O

WAPWR-RC 5.5-54 AMENDMENT 3 BBBBe:1d AUGUST 1989 l

,- between 150'F to 200*F in the vessel flange area which could produce thermal stresses in the flange area or in the studs that might exceed code allowables when added to the stresses already considered (bolt-up leads, pressureloads,etc.)B&W acknowledged the preliminary nature of their

, analysis and noted that, if their conservatively calculated cooling rate of the stagnant coolant in the vessel head (2'F/hr) were to be on the order of 20*F/hr, then the estimated vessel stresses would not be excessive.

The safety significance of this unanalyzed reactor vessel thermal stress is that, when added, to the existing stresses, the stresses in the flange area or studs may exceed the allowable stress. Moreover, the cycling of these temperature gradients over the life of the plant may cause a reduction in the fatigue margin or usage factor of the vessel over the life of the plant. In addition, depending upon the vessel temperature distribution, there is a possibility of vessel fracture under these 3

circumstances. These factors could cause vessel cracking leading to unacceptable vessel failure during the life of the vessel. However, it is assumed in this analysis that sufficient water is available to prevent dry-out of the steam generators. Otherwise, the consequences could be l more serious than is presently estimated.

If these unanalyzed thormal stresses do cause a reduction of fatigue life or lead to vessel stresses that exceed the allowable stresses for the ,

vessel, the solution is assumed to be a slower cooldown rate than the presently allowable rate of 100*F/hr.

i p SP/90 Response  !

Y Stress analyses will be performed in accordance with ASME Section III requirements. Should additional analyses be required as a result of the resolution of this issue, such analysis will be performed for the FDA (q_/ submittal.

WAPWR-RC 5.5-55 AMENDMENT 3 E888e:1d AUGUST 1989

_- _ _ _ _ - _ _ _ _ - _ _ _ - _ _ _ _ _ _ _ _ _ _ . _. _ _ _ n

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80. Issue 80: Pipe Break Effects oo Control Rod Drive Hydraulic Lines in the g l.

Drywells of BWR Nark I and II Containments W l-1 This issue is not applicable to the Westinghouse pressurized water reactor designs.

81. Issue 81: Impact of Locked Doors and Barriers on Plant and Personnel ,

Safety This issue has been dropped and is no longer under review.

82. Issue 82: Beyond Design Basis Accidents in Spent Fuel Pools Discussion The risks of beyond design basis accidents in the spent fuel storage pool were examined in WASH-1400. It was concluded that these risks were orders

3. of magnitude below those involving the reactor core. The besic reason for this is the simplicity of the spent fuel storage pool--the coolant is at atmospheric pressure, the spent fuel is always suberitical and the heat source is low, there is no piping which can drain the pool, and there are no anticipated operational transients that could interrupt cooling or cause criticality.

The reasons for re-examination of spent fuel storage pool accidents are two-fold. First, spent fuel is being stored instead of reprocessed. This had led to the expansion of onsite fuel storage by means of high density storage racks, which results in a larger inventory of fission products in I the pool, a gretter heat load on the pool cooling system, and less distance between adjacent fuel assemblies. Second, some laboratory studies have provided evidence of the possibility of fire propagation between assemblies in an air-cooled environment. These two reasons, put together, provide the basis for an accident scenario which was not i previously considered.

O WAPWR-RC 5.5-56 AMENDMENT 3 5888e:1d AUGUST 1989

( A typical spent fuel'~ storage pool with high density storage racks can hold roughly five timer, the fuel in the core. However, since reloads typically discharge one third of a core, much of the spent fuel stored in the pool will have had considerable decay . time. .This reduces the radioactive p ' inventory somewhat. More importantly, after roughly three years of A/ storage, spent fuel can be air-cooled, i.e., such fuel need not be sub-merged to prevent melting. (Submersions still desirable for shielding and to reduce airborne activity, however.)

If the pool were to be drained of-water, the' discharged fuel from the last two refuelings would still be " fresh" enough to melt 'under decay heat.

However, the zircaley' cladding of this fuel could be ignited during the heatup. The resulting fire, in a pool equipped with high density storage racks, would. probably spread to most or all of the fuel in the pool. The heat of combustion, in combination with decay heat, would certainly release considerable gap activity from the fuel and would probably drive

" borderline aged" fuel into a molten condition. Moreover, if the fire 3 i becomes oxygen-starved (quite probable for a fire located in the bottom of a pit such as this), the hot zirconium would rob oxygen from the uranium dioxide fuel, forming a liquid mixture of metallic uranium, zirconium, oxidized zirconium, and dissolved uranium dioxide. This would cause a

~

release of fission products from the fuel matrix quite comparable to that of molten fuel. In addition, although confined, spent fuel pools are i almost always located outside of the primary containment. Thus, release

]

to the atmosphere is more likely than for comparable accidents involving the reactor core.

I LWR spent fuel storage pools do not differ greatly. None are equipped with drains; a portable pump must be brought in when it is desired to empty the pool. .The cooling systems are provided with anti-siphoning ,

i devices (checkvalvesand/oranti-siphoningholes)sothat pipe breaks in l- the cooling system will not drain the pool. All are seismic Category I.

One difference does exist: PWR pools are generally below grade (often on bedrock) while BWR pools are considerably above grade. Thus, even a hole O

WAPWR-RC 5.5-57 AMENDMENT 3 5888e:1d AUGUST 1989

in the bottom of the pool will not rapidly drain a PWR pool. This

. priority determination, therefore, is concentrated on a BWR pool because of its (somewhat) greater vulnerability.

l SP/90 Response &

1 L

This issue involves the somewhat remote possibility of zircaloy claddi g l being ignited in the event relatively fresh spent fuel is uncovered due to pool failure. The circumstances of ignition and propagation, and the consequences and likelihood of a severe spent fuel pool accident are to be assessed by the staff. It has already been recognized that the scenarios leading to such an event are related to a greater degree to BWRs than to PWRs.

Furthermore, BNL and LLNL work completed to date indicates that the 3 likelihood of a pool drainage event may be acceptably low, in combination with the consequences evaluation, to consider a regulatory analysis which

. justifies no action. ,

Based on the foregoing, there are strong indications that this issue will have no impact on the Westinghouse SP/90 design. However, should a regulatory requirements package that affects the design of the Westinghouse SP/90 be issued, Westinghouse will meet the requirements. .

The manner in which the requirements will be met cannot be given until these requirements are specified.

83. Issue 83: Control Room Habitability Discussion This issue addresses concerns identified by the ACRS regarding (1) I' deficiencies in the maintenance and testing of engineered safety features g designed to maintain control room habitability; (2) design and installa- W 7 tion errors, including inadvertent degradation of control room leak }

O WAPWR-RC 5.5-58 AMENDMENT 3 BB8Be:1d AUGUST 1989

l

-4

tightness; and, (3) the shortage of NRC and licensee personnel knowledge-V: able about HVAC . systems and' nuclear air-cleaning technology. These ACRS 1 concerns' encompassed both plant-licensing review and operations / inspection activities.

Loss 'of control room habitability following an accident release of external airborne toxic or radioactive material or smoke can impair or cause loss of the control room operators' capability to safely control the

.g. reactor and could lead to a core damaging accident. Use of the remote i shutdown station outside' the control room following such events is unreliable since this. station has no emergency habitability or radiation protection provisions similar to the control room.

Three other issues are related to this issue:- B-36, " Develop Design Testing and Maintenance Criteria for Atmospheric Cleanup System Air Filtration and Absorption Limits for Engineere'd Safety Feature Systems and for Normal Ventilation Systems"; B-66, " Control Room Infiltration Measure-3 monts"; and TMI Action Plan Item III.D.3.4, " Control Room Habitability."

O SP/90 Response The control room habitability system is described in Subsection 6.4 of RESAR-SP/90 PDA Module 13 " Auxiliary Systems." Inspection and testing requirements will- be addressed in the FDA submittal and will consider any

~a dditional requirements resulting from the resolution of this issue.

84. Issue 84: CE PORVs Discussion Following the TNI-2 eccident, the purpose and use of PORVs has been the subject of considerable analyses and discussions. The original purpose O for which PORVs were installed was to prevent challenges to the spring-operated safety valves. However some plants rely on the PORV to depres-surize the plant in certain design basis events such as a steam generator O

WAPWR-RC 5.5-59 AMENDMENT 3 8888e:1d AUGUST 1989

I w

1 tube rupture. Another use of the PORVs is to provide low temperature overpressure protection. A more in-depth discussion on the use of the PORVs in various modes of plant operations is provided in Issue 70.

This issue addresses potential safety concerns associated with the absence of PORVs in CE designed plants.

l I

SP/90 Response j i

This issue is not applicable to Westinghouse pressurized water reactor designs.

)

85. Issue 85: Reliability of Vacuum Breakers Connected to Steam Discharge Lines Inside BWR Containments This issue is not applicable to Westinghouse pressurized water reactor designs.

3

86. Issue 86: Long Range Plan for Dealing with Stress Corrosion Cracking in BWR Piping Discussion l

This issue is not applicable to Westinghouse pressurized water reactor designs.

87. Issue 87: Failure of HPCI Steam Line Without Isolation Discussion This issue concerns a potential break in the High Pressure Coolant Injection steam supply line (or Reactor Water Cleanup, Reactor Core Isolation Cooling, or other systems) where valves might not be qualified to close in the event of a downstream break. Failure to close could result in a LOCA outside containment for BWR plants. This issue is not applicable to the Westinghouse pressurized water reactor design.

WAPWR-RC 5.5-60 AMENDMENT 3 E888e:1d AUGUST 1989

N 88.. Issue 88: Earthquakes and Emergency Planning

-V Discussion

- This issue was initiated to address concerns raised by the Union of ,

Concerned Scientists. The purposes for including this issue as a generic issue are to: (1) provide brief background information that summarizes the history of the issue; (2) reduce the probability of resurrecting the {

iscue and duplicating effort; and (3) identify the final disposition of {

the issue.

Recent PRAs have indicated that earthquakes (and other external events) can cause severe reactor accidents which are comparable with internally initiated accident event sequences. The results argued for a reexamination of the emergency response measures to ascertain whether they are adequate to protect the health and safety of the public. The issue effects all operating and planned nuclear power plants.

3 O On August 19, 1980, the Commission published its rule on emergency planning establishing 16 planning standards (see 10 CFR 50.47(b)] which must be generally met by both onsite and offsite emergency response plans

~

for nuclear power plants. The planning standards are addressed by specific evaluation criteria in NUREG-0654, Revision 1. Thus, the NRC 1 emergency planning requirements and guidance reflect coordinated efforts with the Federal Emergency Management Agency (FEMA). Both the NRC and FEMA shared the view that the required emergency response plans have considerable flexibility to respond to a wide variety of adverse conditions, including those resulting from an earthquake.

This issue was RESOLVED and no new requirements were established.

( SP/90 Response 1

The emergency response guidelines are essentially the responsibility of each plant specific applicant, and they should consider the adverse O

WAPWR-RC 5.5-61 AMENDMENT 3 5888e:1d AUGUST 1989 l

l L_- _ _-- - - _ -

i

'4

, conditions of a worst case earthquake in the development of their emergency response plans.

Westinghouse.

This issue requires no action on the part of h'

(.

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89. Issue 89: Stiff Pipe Clamps This issue has not been further defined or prioritized by the NRC.

O

90. Issue 90: Technical Specifications for Anticipatory Trips Discussion O

This issue has been assigned a low priority. Additionally, there are no technical specifications for the SP/90 at the PDA stage. During the Final Design this issue will be reviewad as it stands at that time.

91. Issue 91: Nain Crankshaft Failures in Transamerica Delaval Emergency 3

Diesel Generators Discussion-L On August 12, 1983, one of the three 6mergency diesel generators (EDG) at the Shoreham Plant failed during overload testing as a result of a fractured crankshaft. The failure occurred in EDG-102 and similar crankshaft cracks were discovered in EDG-103 and EDG-101 on August 22 and 23, 1983, respectively. In addition to the crankshaft cracks, 4 of 24 connecting rod bearings were found to contain cracks in the bearing shells. All 3 EDGs were supplied by Transamerica Delaval, Inc. (TDI) and were Model DSR-48 diesels.

As a result of the EDG failure at Shoreham, a TDI Project Group was established by NRR on January 16, 1984. On January 25, 1984, the staff provided the Commission with a status report in SECY-84-34. In order to more clearly define the issue and to determine remedial action, the staff issued a letter to TDI on February 14, 1984 requesting more information.

O WAPWR-RC 5.5-62 AMENDMENT 3 B888e:1d AUGUST 1989

i f,. '1 4

W In March 1984, the TDI Diesel Generators Owners' Group submitted to the i V NRC its program for addressing the issue.

r 1 The~ staff's overall finding was that the OGPP incorporates the essential p elements needed to resolve the outstanding concerns relating to the d reliability of the' TDI diesel generators for nuclear service, and to ensure that the TDI diesel engines comply with GDC 1 and GDC 17. These corrective actions include: (1) resolution of known generic problems

. (Phase I), (2) systematic DR/QR of all components important to reliability

\ and operability of the engines (Phase II), (3) appropriate engine inspections and testing as identified by the results of Phases I and II, and (4) appropriate maintenance and surveillance programs as indicated by the results of Phases I and II.

After licensees complete Phases I and II of the OGPP, the licensing basis will be reviewed by the staff to determine what modifications to the license conditions will be required. A final SER will be issued for each 3 of the plants that are being licensed or restarted on an interim basis.

These are expected to include: Shoreham, Grand Gulf, San Onofre, Catawba, and Comanche Peak. For plants where Phases I and II are scheduled to be completed sufficiently ahead of licensing or restart, a final TDI Diesel SER will be developed that encompasses the results of Phases I and II and the operational history of an engine.

In the event of loss of offsite power, the power to operate the equipment necessary to maintain core cooling is provided in most plants by EDGs.

Although to varying degrees, plants can withstand the loss of both offsite  !

and onsite AC power (and further requirements are being proposed in USI A-44), EDG unreliability is a significant contributor to the estimated frequency of core damage events. The question of diesel generator reliability in general is addressed in Item B-56, " Diesel Reliability."

Issue 91 applies to the design and operation of the 16 plants which have O or have not ordered TDI diesel generators.

O WAPWR-RC 5.5-63 AMENDMENT 3 8888e:1d AUGUST 1989

The possible solutions to this issue are considered to be the three elements of the TDI OGPP:

g (a)PhaseI: Resolution of 16 identified generic problem areas intended (by the Owners' Group) to serve as a basis for the licensing of plants during the period prior to completion and implementation of the OGPP. 4 (b)PhaseII: A design review / quality revalidation of a larger set of important engine components to assure that their design and manufacture (including specifications, quality control, quality assurance, operational surveillance, and maintenance) are adequate.

(c) Identification of any needed additional engine testing or inspections based on findings from Phases I and II.

In response to the problems raised in this issue, the Owners' Group 3 performed extensive design reviews of all key engine components and developed recommendations to be implemented by the individual owners concerning needed component replacements and modifications, component inspections to validate the "as-manufactured" and "as-assembled" quality of key engine components, engine testing, and an enhanced engine

, maintenance and surveillance program.

The staff's evaluation of the Owners' Group program, documented in NUREG-1216, concluded that implementation of the Owners' Group recommendations, plus additional actions identified, will establish the adequacy of the TDI diesel generators for nuclear standby service as required by GDC 17 of 10CFR50, Appendix A. The staff further concluded that these actions will ensure that the design and manufacturing quality of the TDI engines is within the range normally assumed for diesel engines designed and manufactured in accordance with 10CFR50, Appendix B.

Continued reliability and operability of the TDI engines for the life of the facilities will be ensured by implementation of the l maintenance / surveillance program described in NUREG-1216. Thus, this I

issue was RESOLVED and no new requirements were established.

O WAPWR-RC 5.5-64 AMENDMENT 3 B888e:1d AUGUST 1989

l; -SP/90 Response a .

l This issue deals with the failure of a specific vendor's emergency diesel ]

generators, and the licensing of operating or near-term operating license J applications utilizing that design, and has no direct impact on the SP/90 design. The reliability and operability of the subject EDG design can be q ensured by . implementation of a maintenance / surveillance program as l described in NUREG-1216. Use of the EDG models discussed here in an SP/90 O plant design would require adherence to those maintenance requirements.

V

92. Issue 92: Fuel Crumbling During a LOCA This issue has been assigned a low priority and will be addressed at the Final Design stage to evaluate any impact on the SP/90 design.

93. Issue 93: Steam Binding of Auxiliary Feedwater Pumps 3

Discussion This issue was identified after a review of vapor binding of the AFW pumps at H. B. Robinson Unit 2.

The H. B. Robinson report discusses thirteen occurrences reported in 1983 of steam binding of one or more AFW pungs resulting from the leakage of heated main feedwater into the AFW system. The systems are isolated by various combinations of check valves and control valves. The back-leakage occurred through several valves in series. The heated main feedwater, leaking into the AFW system, flashed to steam in the pumps the AFW discharge lines and resulted in steam binding of the AFW pumps.

Operating experience to date includes 22 events of reported back-leakage in 6 operating PWRs in the USA and at 1 foreign reactor. In other cases, back-leakage has been observed but was not considered as reportable occurrences.

O WAPWR-RC 5.5-65 AMENDMENT 3 5888e:1d AUGUST 1989

i The potential for common mode failure is present whenever one pump is g steam-bound because the pumps are connected to common piping with only a T single check valve to prevent back-leakage of hot water to the second or l third pump. Steam binding of more than one pump was reported to occur in l

3 of the 13 events reported 1983.

O The back-leakage of steam represents a potential common cause failure for f the AFW system could result in the loss of its safety function.

)

SP/90 Response

'In the SP/90 plant, the traditional Auxiliary feedwater (AFW) System has been replaced by two separate systems, i.e.:  ;

o A control grade Startup Feedwater System (SFWS), which is used during startup and shutdown conditions, as well as following a 3

reactor trip.

o A safety grade Emergency Feedwater System (LFWS), which is used whenever the SFWS is unavailable as well as during Condition III g

and IV events.

This Generic Safety Issue applies to the latter system, which is described in Section 10.4.9 of RESAR-SP/90 PDA Module 6/8, " Secondary Side Safeguards System / Steam and Power Conversion System." Specific features addressing the steam binding issue include:

o The stHet separation into two totally independent subsystems in general reduces the probability of common mode failures, e.g. steam binding.

o Each pump is protected against backflow by at least two check g valves in series; temperature sensors with indication and alarm in W the Main Control Room are provided to alert the operator to failure of the check valve nearest the steam generator.

O l WAPWR-RC 5.5-66 AMENDMENT 3 588Be:1d AUGUST 1989 i

l

t. A o If failure of a check valve is detected, the operator has the f3

v' optien to close the motor operated valves 9956A and 9957A (or 9956B and 99578), thereby preventing the backflow from affecting more.than one EFW pump (note that closure of these valves does not )

affect EFWS reliability, but does reduce the likelihood of feeding j all steam generators.)

(

o Individual suction lines to each pump eliminate the possibility that back-leakage through one pump could affect the other pump in ~

that subsystem.

Furthermore, the SP/90 plant 'is designed to allow decay heat removal with the EFWS unavailable; both the SFWS previously identified and feed and i bleed operation are able to provide this function.

The SP/90 plant adequately addresses the concerns contained in Generic Safety Issue 93 by: 3 O o Incorporating design feature to reduce potential for steam binding O of EFW pumps.

o Including appropriate instrumentation to detect potential onset of

~

steam binding.

o Providing alternative decay heat removal capability -in the unlikely event that the EFWS is unavailable as a result of steam binding, or indeed any other common : node failure.

94. Issue 94: Additional Low Temperature Overpressure Protection for Light Water Reactors Discust. ion Major overpressurization of the RCS, if combined with a critical size crack, could result in a brittle failure of the reactor vessel. Failure

>O WAPWR-RC 5.5-67 AMENDMENT 3 588Be:1d AUGUST 1989

l of the reactor vessel could make it impossible to provide adequate coolant g to the reactor and result in a major core damage or core-melt accident. W l This issue applies to the design and operation of all PWRs.

Low temperature overpressurization (LTOP) was previously identified as USI A-26 and was resolved in September 1978 with a revision to SRP 4 Section 5.2. The resolution of USI A-26 affected all operating and future PWRs and required PWR licensees to implement procedures to reduce the potential for overpressure events and install equipment modifications to mitigate such events.

' Subsequent to the reso'lution of US! A-26, additional pressure transients were reported, including two events at Turkey Point Unit 4. The overpressurization transients at Turkey Point exceeded the TS limits and were identified to Congress as Abnormal Occurrences which indicate that the events involved a major reduction in the degree of protection to the 3 public health or safety.

The continuation of overpressure transient events and the two instances at Turkey Point may indicate potential weaknesses in the present overpressure g

protection criteria or its implementation that warrant further consideration. The two overpressure transient occurrences at Turkey Point resulted from one overpressure mitigation system (OMS) channel being out for maintenance and the other (redundant) channsi being disabled by undetected errors during the first event and from undetected equipment malfunctions during the second event.

The NRC work scope for resolution of this generic issue includes evaluation of current operating data and consideration of all or some of the following proposed new requirements:

a. Amend the STS and the SRP to require each licensee to identify the criteria used to determine if and when the LTOP system setpoints need to be adjusted to acccunt for the irradiation-induced embrittlement of the reactor vessel.

O WAPWR-RC 5.5-68 AMENDMENT 3 B88Be:1d AUGUST 1989

f~ b. Nake more use of the' relief valkes' in the RHR f or LTOP by raising A? the setpoint for the auto-closure of the isolation valves.

c. Amend the STS to allow no plant operation in the " water-solid" condition with either. train of the LTOP system out of service.

i

d. ' Amend the STS to allow no plant to operate in the " water solid" j I

condition with an SI pump in service.

e. Require the LTOP system to be fully safety grade.

f. Require all operating reactors to upgrade their TS to the STS for LTOPs.

SP/90 Response Subsection 5.2.2.10 of RESAR-SP/90 PDA Module 4, " Reactor Coolant System," 3 details the LTOP features provided in the SP/90 design to preclude a breach of the reactor vessel due to overpres:urization of the reactor coolant system during low temperature operation. Standard Technical Specifications will be part of the FDA subaittal and will include administrative controls to ensure the availability of the LTOP system.

95. Issue 95: Loss of Effective Volume for Containment Recirculation This issue has not been further defined or prioritized by the NRC.

96. Issue 96: RHR Suction Valve Testing This issue has not been further defined or prioritized by the NRC.

97. Issue 97: PWR Reactor Cavity Uncontrolled Exposures This issue was subsumed into TMI Action Plan Item III.D.3.1.

O WAPWR-RC 5.5-69 AMENDMENT 3 5888e:1d AUGUST 1989

s

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98. Issue 98: CRD Accumulator Check Valve Leakage This issue has been dropped and is no longer under review.

99. Issue 99: RCS/RHR Suction Line Valve Interlock on PWRs  !

Discussion O'

Interlocks are provided to assure that there is a double barrier (two closed valves) between the RCS and RHR systems when a plant is at normal operating conditions, i.e., pressurized and not in the RHR cooling mode.

A related issue (Issue 96) addresses the concern of assuring that both j series RHR isolation valves are closed during normal power operation. '

Issue 99 is concerned with the inadvertent closing of these valves when ,

the RHR system is in use.

3 Two basic features are incorporated in the interlock design: (1) an automatic closure signal on high RCS pressure (typically 600 psig), and (2) a block of the manual open signal at a lower RCS pressure (typically 425 psig). The autoclosure setpoint is generally set higher than the design pressure of the RHR system. However, overpressure protection of  !

the RHR system during RHR cooling is provided by relief valves and not by the slow-acting RHR suction valves. The block setpoint is lower than the RHR system design pressure to preclude opening of either RHR suction valve when the RCS is at a higher pressure.  !

In the W design, two interlock channels are provided such that one channel is used to interlock the operation of one RHR suction valve and the other <

channel is used for the other valve. The same interlock configuration is  !

used in }{ plants for designs that have one or two RHR drop lines from the  ;

RCS. When either channel is in a tripped state, its associated suction valve will automatically close if it is open. Since the relays used for this interlock are deenergized to initiate valve closure, a loss of the [

instrument bus used for either channel will result in e loss of RHR ,

cooling due to inadvertent closure of one of the suction valves.  !

9; WAPWR-RC 5.5-70 AMENDMENT 3 B888e:1d AUGUST 19,89

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() The loss of one instrument bus or disablement of one logic channel will result in the automatic closure of one of the RHR suction line isolation l valves. In the RHR cooling mode, such closure gives rise to the potential

- for RHR pump damage and loss of decay heat removal by the RHR system.

I This safety concern appli6s to all W reactors. 3 I

Operating experience shows that a change in the autoclosure feature of the l RCS/RHRS isolation valve interlocks is needed to eliminate a leading cause i of the loss of residual heat removal during plant outage operations.  !

Additional loss-of-RHR events have resulted from inadequacies in the RCS water level measurement and related monitoring by the operator as well as by other human error related inadequacies in plant outage procedures, f training, and administrative controls.

The scope of Issue 99, initially directed solely at the autoclosure interlock (ACI)-related mode of RHR failure, was broadened in June 1986 to include the less frequent but higher risk mode of failure associated with 3

OV mid-loop operations. The risk aspects of cold shutdown loss of cooling l were studied by BNL and the results, published in NUREG/CR-5015 in May 1988, were applied in the preparation of the regulatory impact analysis for this issue. In a related development following the Diablo Canyon 2

~

loss of-RHR event of April 1987, NRR undertook to define the needs for licensee short- and long-term regulatory actions in regard to the perceived deficiencies in the conduct of PWR mid-loop operations. The regulatory requirements for improved licensee operations recommended by NRR were in overall substantive agreement with those proposed by RES in p the regulatory analysis for Issue 99. NRR's issuance of Generic Letter

' 88-17 on October 17, 1988, requesting PWR licensees and applicants to implement plant improvements pertinent to the concerns of Issue 99 provided the resolution of this issue.

The staff concluded that the proposed requirements for improved instrumentation, procedures, and administrative controls were highly cost-beneficial in reducing the estimated baseline core damage frequency l

WAPWR-RC 5.5-71 AMENDMENT 3 5888e:1d AUGUST 1989 l

1

l i

(CDF)byafactoroften. The value/ impact results also supported the proposed requirement for closure of the containment during mid-loop operations, at least pending the appropriate implementation of the CDF-reduction recommendations. Finally, the cost / benefit evaluation of a proposal to remove the ACI to obtain an addition minor reduction in CDF suggested that removal of the ACI be recommended, but not required, for h !

plant implementation. Thus, this issue was RESOLVED and requirements were established.

SP/90 Response 1

Westinghouse Owners Group is currently reviewing the use of autoclosure interlocks on RCS/RHR isolation valves and the manner in which RCS water Itvels are measured in an attempt to reduce the loss of residual heat removal capability during plant outages. The current SP/90 design does not include autoclosure interlocks and is therefore not effected by this 3

issue. The SP/90 FDA design will, however, consider any additional regulatory requirements as well as the recommendations upon conclusion of g the WOG finding on this issue. W 100. Issue 100: OTSG Level This issue has not been further defined or prioritized by the NRC.

101. Issue 101: Break Plus Single Failure in BWR Water Level Instrumentation Discussion This issue is concerned with BWR water level instrumentation, specifically with the concern related to a break in an instrument line in conjunction with the worst single failure. This issue is not applicable to the Westinghouse pressurized water reactor design.

l Ol l

WAPWR-RC 5.5-72 AMENDMENT 3 BB88e:1d AUGUST 1989 i

i

! t .

102. Issue 102: Human Error in Events Involving Wrong Unit or Wrong Train Discussion In January 1984, NRC issued a special study report describing _the number of. events that resulted from human error in removing equipment from service or rtistoring equipment. This study focused on Licensee Event reports issued during 1981, 1982, and part of 1983.

l Although the scope of its study was narrow, it was found.that 19 out of 1 27 events identified.resulted from human error during maintenance and surveillance testing;. 16 of these occurred while the plants were at power. Although most of the events had limited safety significance because of the short duration of the condition and/or because redundant systems were operable and available, NRC considered them to be examples of events that could have high safety significance under other circum-stances. -As a result, it was concluded that the above statistic was 3 evidence that human errors in maintenance and testing operations are major contributors to loss of safety system events.

A The possible solutions to this issue include:

~

o . Consider the need for further clarification and/or guidance on what constitutes an acceptable independent verification program, o Review the wrong unit / train events and develop appropriate guidance to minimize such events.

o As part of the Maintenance and Surveillance Program Plan (MSPP),

consider the high proportion of events that were due to human error in maintenance and testing operations at power.

O Although the concerns outlined above are to be addressed in the MSPP, (see issue HF08) NRC has decided to pursue resolution of Issue 102 separately from Issue HF08.

O WAPWR-RC 5.5-73 AMENDMENT 3 B888e:1d AUGUST 1989

--- _ _ . _ . _ _ . _ _ _ _ _ _ _ _ _ _ _ . _ _ _ - - m ___-__-__._.--.___________._______-_.______________________.__o_

In addressing this issue, the staff reviewed operating experience, conducted site visits, and interviewed licensee personnel to determine the nature and root causes of 35 wrong unit and wrong train events. The results indicated that the primary causes of these events were inadequate labeling of areas, equipment, and components, inadequate personnel g training and experience, and inadequate procedures; these results were W reported in NUREG-1192. In June 1987, Information Notice No. 87-25 was issued and reiterated the primary causes of the subject events and called for the industry to increase its attention in this area. Wrong unit / wrong train cot 9enent concerns are being addressed in the broader context of the Emergency Operating Procedure Inspection program, Detailed Control Room Design Reviews, and future integrated inspections which include assessment of local control stations and maintenance programs.

During discussions with INPO, subsequent to publication of NUREG-1192, the staff learned that INPD reviews licensee actions to resolve this issue as part of their planc evaluations and INPO expects to continue 3 such reviews in the future.

Based on the above staff actions and industry initiatives, NRR concluded that no further staff action was warranted. Thus, this issue was RESOLVED and no new requirements were established.

SP/90 Response The concerns relt.ted to this issue fall more under the responsibility of the plant specific applicant (utility) rather than the plant systems' supplier. However, specific SP/90 design features should greatly reduce the concerns of this issue. The wrong-unit concern is not an issue in SP/90 because there is no sharing of systems between units. And, because the SP/90 plant . features simplified system designs and an improved layout, the added separation and redundancy of the SP/90 will reduce the probability of a wrong trsin event occurring.

The online testing capability further reduces the chance of this type of human error from occurring.

O WAPWR-RC 5.5-74 AMENDMENT 3 B8BSe:1d l AUGUST 1989

103. Issue 103: Design for Probable Maximum Precipitation Discussion Improper drainage at reactor sites during heavy rainfalls can lead to f -

flooding that can render safety-related equipment inoperable. Issue 103 concerns procedures for determining probable maximum precipitation (PMP). The PMP values are used in estimating design flood levels at reactor sites. Specifically, the most recent NOAA Hydrometeorological

( Report (HMR) in general results in higher flood levels than those obtained using earlier reports, and cited in SRP Section 2.4.2. This expansion extends the precipitation duration from 48 to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and increases the drainage areas from 1,000 to 20,000 square miles. In addition to other provisions, HMR-52 provides techniques for analyzing PMP for drainage areas of I square mile and durations of one hour and less. ,

3 GDC-2 requires that design bases for flood reflect consideration of the most severe historical data with sufficient margin for the limited accuracy, quantity, and period of time in which data have been accumu-lated. Guidance on what constitutes sufficient margin is contained in

. Regulatory Guides 1.59 and 1.102. These documents state that the appro-priate design basis for precipitation-induced flooding is the probable maximum flood (PMF) as developed by the U.S. Army Corps of Engineers.

This PMF criterion has been used by NRC since 1970. Thus, in the case of floods, the PNF is the criterion that has been used to meet GDC-2.

Procedures for estimating PMFs are given in Appendices A and B of Regulatory Guide 1.59 (Appendix A has since been superseded by ANSI N170-1976). ANSI N170-1976 defines PMF as a hypothetical flood that is considered to be the most severe reasonably possible, based on comprehen- ,

sive hydromet'enrological application of PMP and other hydrologic factors favorable for maximum flood runoff. Thus, PMP is an integral component of PMF determination. Section 5.2 of ANSI N170-1976 states that PHP estimates for the U.S. are available in generalized studies prepared by

\

l WAPWR-RC 5.5-75 AMENDMENT 3 E888e:Id AUGUST 1989

l l

the National Weather Service (NWS), these estimates are presented in g i varying degrees of completeness. Specific PHP estimates for areas not W adequately covered by these studies may be made by using techniques l similar to those employed by NWS.

Recognizing the importance of using the most recent engineering technology in evaluating the potential impacts on reactor site safety,

{

SRP Section 2.4.2 was written to allow "... improvements in calculational methods ..." With the publication of HMR-51 and HMR-52, OL applicants were requested by the staff to assess the effects of their use on plant safety.

~

I In all cases reviewed by the staff against HMR-51 and HMR-52, the issue has been resolved by the applicants taking the following actions: (1) site drainage has been designed to handle the increased design basis precipitation, (2) commitments were made to develop procedures to assure 3 that critical entrances to buildings will be closed, and (3) curbs were installed at critical entrances. In order to clarify the staff's position and remove ambiguities from the SRP, it will be necessary to revise SRP Sections 2.4.2 and 2.4.3. This solution will be a forward-fit g

and will incorporate the most recent technical advances for determining PMP that are known at the time that the SRP revision is made. Future

~

technical advances in the determination of PMP will also require revisions to the SRP. i l

SP/90 Response The SP/90 plant is designed in accordance with the site envelope specified. Applicability of the site envelope will be verified for each plant utilizing the SP/90 design. j 104. Issue 104: Reduction of Boron Dilutien Requirements This issue has been dropped and is no longer under review.

O WAPWR-RC 5.5-76 AMENDMENT 3 l BBBBe:1d AUGUST 1989

s 105. Issue 105: Interfacing Systems LGCA at LWRs Discussion This issue concerns the suitability of leak test and . operability test requirements for valves that isolate the low pressure systems that are connected to the RCS and outside the containment.

This issue was originally limited to pressure isolation valve testing concerns on BWRs. Recent BWR operating experience indicates that the isolation valves between the RCS and low pressure interfacing systems (including related test and maintenance requirements) may not adequately protect against overpressurization d low pressure systems. There have been three reported failures of ths boundary between the RCS and low pressure injection systems. Two of the events were the result of maintenance errors which left the testable isolation check valve in the open position. The third was the result of personnel errors (improper 3 combination of surveillance tests) and a stuck open failure of an isolation check valve. In all three of these cases, there was a degrada-tion of the pressure isolation valves due to personnel errors. None of these plants was required to leak test pressure isolation valves.

Overpressurization of low pressure piping systems due to RCS boundary isolation failure could result in rupture of the low pressure piping.

This, if combined with failures in the ECI and/or the DHR systems, would result in a core-melt accident with an energetic release outside the containment building causing significant offsite radiation release.

Generic Safety Issue 105 has since been expanded to include concerns with interfacing LOCAs on PWRs. This issue will be resolved with the pressure isolation valve portion of Generic Safety Issue II.E.6.1, Test Adequacy Study.

O WAPWR-RC 5.5-77 AMENDMENT 3 B888e:1d AUGUST 1989

l 4 j SP/90 Response The principal contribution to core melt frequency from interfacing systems LOCA originates from the four residual heat removal (RHR) suction '

linesoftheintegratedsafeguardssystem(ISS). An initiating frequency per year was established in RESAR-SP/90 PDA Wedule 16 of 1.0E-6

'Probabilistic Safety Study" for this event. These lines connect the RCS hot legs to the RHR pumps suction and therefore penetrate the containment boundary; the low pressure portions of these lines are normally isolated from the RCS by two closed motor operated valves in series. In an interfacing systems LOCA scenario, it is postulated that either one valve is inadvertently le"ft open and that the other one fails, or that both valves fail while the RCS is pressurized, causing failure of the piping outside containment.

The SP/SO design includes the following features not normally provided in current plants specifically aimed at reducing the probability of this 3 scenario.

(i) The RHR isolation valves have been included in the system provided to allow leak testing of the valves in the lines connected to the l RCS during plant startup (Figure 105-A). Thus, the probability of one of these valves not being fully closed has essentially been eliminated.

(ii) The design pressure of all RHR piping downstream of the RHR isolation valves (including RHR pump casings) has been increased j such that no gross failure would occur even when exposed to full RCS operating pressure.

(iii) The RHR piping downstream of the RHR isolation valves is normally J in open connection with the EWST such that any leakage through j these valves is normally directed back into containment (Figure 105-B).

I O

WAPWR-RC 5.5-78 AMENDMENT 3 588Be:1d AUGUST 1989

n s-fC (iv) Failure of the RHR pump seal could still be postulated; however, the four separated rooms containing the four subsystems of the ISS have been designed for minimum volume such that when the water levels in the EWST and in the pump room are equal, there remains 3

(\ sufficient inventory in the EWST to ensure continued core cooling with the unaffected ISS subsystems (Figure 105-C).

l These features combine to significantly reduce the probability of a core

(~ melt in case of leakage from or failure of the RHR isolation valves, which has been shown to be the most probable interfacing system LOCA sequence.

The next most probable interfacing system LOCA scenario associated with the ISS results from the high head safety injection (HHSI) paths (Figure 105-D). Failure of the three check valves in series is expected to occur with a frequency of 2.0E-9 per year as derived in RESAR-SP/90 PDA Module 16 "Probabilistic Safety Study". Small or even moderate 3 leakage through these valves would have no effect, since they would be vented back to the EWST. Gross failure of the valves may result in overpressurization of the HHSI pump suction piping, although it should be noted that the injection lines contain flow limiting orifices (for HHSI pump runout protection) which would tend to limit suction piping pressures; more detailed analysis in this respect will be performed at the FDA stage 9.cn specific piping layout information will be available.

Finally, it should be noted that the HHSI injection lines contain remotely operated valves that could be used to terminate the LOCA, and that the layout is such that even with a LOCA outside containment, the EWST would not drain completely (see item (iv) above). .It can therefore be concluded that even in case of an interfacing system LOCA via the HHSI injection lines, .there is reasonable probability that a core melt would not occur.

Other interfacing systems LOCA scenarios in the ISS involve normally closed motor-operated valves and are even less likely to occur than those (

described above.

WAPWR-RC 5.5-79 AMENDMENT 3 5888e:1d AUGUST 1989

L l

" l l

l Outside the ISS, there are other low pressure systems which are connected I

to the RCS. A brief description of these connections is provided below.

i l

o RCS drain and vent connections, and the ISS emergency letdown lines l' are provided with redundant closed isolation valves. Since these lines do not penetrate the containment, the failure of both isolation vulvas would not only be a low probability event, but would not result in an "intersystem LOCA."

l o Sampling system lines both penetrate the containment and may be open during normal operation. These piping connections are provided with a 3/8-inch flow restrictor such that failure would not require safety system actuation. These lines contain redundant and automatic containment isolation valves.

o The CVCS letdown line is normally open during plant operation. This line contains redundant valves (nonna11y open) which will 3 automatically close on low pressurizer level, and redundant and automatic containment isolation valves. g o The CVCS alternate letdown line contains redundant, normally closed isolation valves, redundant and automatic containment isolation valves, an operator controlled throttle valve and a temperature sensor.

o The CVCS charging line contains multiple check valves and redundant isolation valves.

All piping connections up to and including their isolation valves are designed for full RCS pressure and temperature conditions.

9 O

WAPWR-RC 5.5-80 AMENDMENT 3 BB % :Id AUGUST 1989 l

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I o 5.5-81 AMENDMENT 3 WAPWR-RC 5888e:Id AUGUST 1989 l

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WAPWR-RC 5.5-82 AMENDMENT 3 5888e:1d AUGUST 1989

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i 106. Issue 106: Piping and Use of Highly Combustible Gases in Vital Areas Discussion Combustible gases such as H, 2 propane, acetylene, and other fuel gases are used during normal operation of nuclear power plants . es well as in l

-plant laboratories. Most combustible gases are used in limited l quantities and for relatively short periods of time at a nuclear plant. '

H' 2

the most prevalent combustible gas used in nuclear power plants, is i used as a coolant for electric generators in both BWRs and PWRs and is also used in PWRs in association with.the reactor water chemistry as well L as in the waste gas disposal functions. H 2 is used in the volume I control tank (VCT) which is usually located in the auxiliary systems building of PWRs. It is stored as high pressure gas in . storage vessels and is supplied as process to the various systems in the auxiliary systems building through standard piping, usually 3/4-inch in diameter.

As such, the piping is field-run and its location is plant-specific.

Leaks or breaks in the H2 piping and supply system could result in the accumulation of a combustible or explosive mixture of air and H 2within O the auxiliary systems building. Inasmuch as the auxiliary systems 3 building is a safety-related structure which houses most of the compo-nonts of the safety-related systems of the plant, the accumulation of combustible or explosive mixtures of gas represents a threat to the safety of the plant by virtue of the potential disablement of safety-related equipment in the event that the combustible gases are inadver-tently ignited. H detectors 2

can signal the presence and accumulation of gas, but these are not qualified as safety grade equipment and do not have an emergency power source. Thus, they are not regarded as suffi-O cient prot'ection against the development of H leakage 2

and subsequent

, uncontrolled combustion or explosion.

SRP Section 9.5-1, " Fire Protection," currently addresses the safe use of O combustible gases on site so that this matter is a concern primarily for operating reactors licensed prior to the issuance of SRP Section 9.5-1.

O '

WAPWR-RC 5.5-85 AMENDMENT 3 5888e:1d AUGUST 1989

1 l

s

[

L This issue is related to Issue 136, " Storage and Use of Large Quantities of Cryogenic Combustibles on Site." Whereas Issue 106 is concerned with

! the normal process system use of relatively small amounts of combustible gases on site, Issue 136 deals with the considerably greater hazards of much greater amounts of combustible materials introduced by new needs at ,

the site (i.e., solid waste processing and BWR hydrogen water chemistry control and the unique hazards associated with the transport and storage of large quantities of combustibles on site in a cryogenic liquid state).

Large releases of combustible gas and the accumulation of combustible or explosivo mixtures in air, in the event of a piping system break or large leak, can be prevented by the installation of excess flow check valves located close to the source of the combustible gas. SRP Section 9.5-1,

" Fire Protection," recommends the use of excess flow check valves. Other measures are needed to reduce the frequency of, or cause of, combustible gas accumulation accidents from such events as valve malfunctions or 3 leaks, connection or fitting leaks, operations errors, material failures, etc. Plants licensed in accordance with the guidelines of SRP Section 9.5-1 are assumed to be not affected by this issue. Excess flow check valves are an effective "fix" for piping system breaks, but other fixes, gj such as installation or upgrading of H 2

detection systems, design changes, procedural changes, etc., will be required for other types of accidental releases, i

SP/90 Response In the SP/90 plant, hydrogen is supplied to the volume control tank of the chemical and volume control system (CVCS); this tank is located in the auxiliary building which surrounds the containment. However, the volume control tank is located in an area which does not contain safety related equipment because of the general SP/90 philosophy of separation of safety related and non-safety related equipment and systems. Thus, l r

the potential for damage to safety related equipment in case of hydrogen ]

leakage and subsequent explosion is small.  ;

O WAPWR-RC 5.5-86 AMENDMENT 3 5888e:1d AUGUST 1989

s The SP/90 plant addresses the concerns of Generic Safety Issue 106 O primarily by separation. In the FDA stage consideration will be given to additional features that could reduce the potential for hydrogen leakage, e.g., excess flow check valves, higher schedule piping, etc.

O 107. Issue 107: Generic Implication of Main Transformer Failure This issue has not been further defined or prioritized by the NRC.

108. Issue 108: B!!R Suppression Pool Temperature Limits This issue is not applicable to Westinghouse pressurized water reactor designs.

109. Issue 109: Reactor Vessel Closure Failure on ESF 3

This issue has not been further defined or prioritized by the NRC.

110. Issue 110: Equipment Protection Devices This issue has not been further defined or prioritized by the NRC.

111. Issue 111: Stress Corrosion Cracking of Pressure Boundary Ferritic Steels in Selected Environments This issue has been determined to be a Licensing Issue to be addressed by the NRC.

O 112. Issue 112: Westinghouse RPS Surveillance Frequencies and Out-of-Service Times This issue has been classified as a resolved Regulatory Impact issue and requires no action on the part of Westinghouse.

O WAPWR-RC 5.5-87 AMENDMENT 3 5888e:Id AUGUST 1989

113. Issue 113: Dynamic Qualification Testing of Large Bore Hydraulic Snubbers l Discussion j

This issue addresses concerns regarding environmental and dynamic ,

qualification testing, inservice testing and surveillance of Large Bore )I Hydraulic snubbers.

This issue addresses the staff's concern that there are no NRC 1

requirements for dynamic qualification testing or dynamic surveillance testing of large bore, hydraulic snubbers (> 50 kips load rating). The resolution of Issue A-13, " Snubber Operability Assurance," is the development of a Regulatory Guide pertaining to " Qualification and Accep-tance Test for Snubbers Used in Systems Important to Safety." However,

! the Regulatory Guide may only be applied on a forward-fit basis and the need for dynamic testing requirements for large bore hydraulic snubbers 3 (LBHS) in operating plants would remain unresolved.

The issue was raised because of the concern for the integrity of the i steam generator lower support structures when subject to a seismic event. However, the issue is applicable to all LWRs with components, structures, and supports that rely on LBHS for seismic restraint and other dynamic loads such as high energy line breaks and water hammers.

Under this issue, the actions which are to be taken by the staff include an extensive gathering of information regarding operating experience, environmental and dynamic qualification, test and surveillance, and state of the art testing capabilities of large bore hydraulic snubbers. Once having gathered this information, the staff intends to evaluate the g

feasibility of performing periodic inservice surveillance / tests on installed LBHSs, assess the overall impact of snubber reduction initia-tives and the GDC-4 broad scope rule implementation, identify potential improvements in LBHS reliability, and develop reconsnendations for environmental / dynamic qualification and inservice testing / surveillance.

O WAPWR-RC 5.5-88 AMENDbiENT 3 5888e:1d AUGUST 1989  !

_ - ___=__ _ -_ - _ _ _ _ _ _ ___ ___ _______ _ _________ _ _____________ _ _ _ _ _ _ _ _

x 4

SP/90Responho -

Westinghouse equipment specifications help assure LBHS _ reliability by defining a- complete set of- fabrication and functional testing pv requirements. .The Westinghouse equipment specifications

. snubber design requirements, environmental requirements, operability define. the l requirements, functional requirements, and establish a process that-ensures Westinghouse engineering controls and reviews all . vendor e procedures and process 6s used in the man'ufacturing cycle. Each step of the fabrication and shop ' assembly of the snubber, including the review and approval of all procedures, fabrication ' sequences and test results, is closely monitored by Westinghouse engineering and quality assurance personnel to insure ~ all requirements and testing criteria are adhered to. Additionally, each snubber must pass numerous performance qualification tests (i.e., control valve lockup, piston bleed rates, load rating, spring rates, piston drag, etc.) prior to its acceptance and release for shipment. 3

.The Westinghouse equipment specifications also require the vendor to provide functionality analyses, design reports, fluid / seal compatibility analyses, environmental qualification data for " soft parts", installation procedures, spare parts lists, and recommended maintenance practices.

Westinghouse's attention to detail and complete definition of~ snubber design, function, operability, and maintenance requirements help assure a high reliability in LBHSs.

I The steam generator LBHSs are assumed to perform their intended function for the 40 ' year operating life of the plant. The snubbers are required to provide minimum resistance against normal reactor coolant loop thermal expansion (drag less than or equal to 5000 lbs. per snubber) and to

" lockup" at velocities greater than or equal to 6 inches per minute. In the locked condition the snubbers are required to bleed at 0.15 to 0.25 inches per minute at full fruited rated load. In addition, the snubbers are to remain fully functional when exposed to a radiation level of 25 RAD /hr, 70% humidity, and 120 degrees Fahrenheit long term operation and 300 degrees Fahrenheit short term operation.

WAPWR-RC 5.5-89 AMENDMENT 3 5888e:1d AUGUST 1989

N Westinghouse has been defining the requirements for and procuring steam gi generator LBHSs for over twenty years. The technical and material W )

advances in LBHS technology that have taken place have been evaluated and adopted into the Westinghouse requirements for LBHSs. This knowledge will be implemented in the.SP/90. 3 The majority of LBHS procured by Westinghouse have been designed and manufactured by Paul-Munroe Hydraulics. The Paul-Munroe steam generator LBHS has implemented design features which help ensure the reliability of the snubber. Features such as a self cleaning bleed orifice and a corrosion resistant chrome lined carbon steel cylinder contribute to the reliability of the snubber. Additional information on the reliability of Paul-Munroe steam generator LBHSs was provided in our March 1989 response to NRC Question 730.3.

Currently dynamic testing of production scale steam generator snubbers is 3

not performed. As such, Westinghouse requires a rigorous series of static functional tests (drag, lockup, bleed) to be performed on each snubber. These tests, although not dynamic, verify the operability of the snubbers at their rated load capacity. Until recently, the means of g

dynamically testing these snubbers was not available. However, Paul-Munroe has performed a limited number of dynamic tests of steam generator snubbers. A drop weight test with an equivalent loading exceeding the rated load has been performed. Small scale dynamic testing has been performed including tests for a cyclic force of 90 kips applied at up to 16 Hz. and a cyclic force of 450 kips applied at up to 11 Hz.

These tests represent the approximate limits of current testing facilities. The results of these tests demonstrated that the Paul-Munroe snubbers performed as predicted during dynamic loading conditions.

Due to limits on the capacity of the available testing facilities, dynamic tests of steam generator snubbers for the full rated load and a full range of frequencies (1-33 Hz.) is not possible. Westinghouse,  ;

O WAPWR-RC 5.5-90 AMENDMENT 3 58BBe:1d /.UGUST 1989 1

~ ,

however, does support the future development of such facilities that will V permit. this type of testing and qualification of the steam generator snubbers. ,

Currently, inservice inspection requirements of steam ganarator snubbers

]

{V is defined by individual plant technical specifications. Many of these technical specifications reflect the latest requirements of ASME's OM4 i inservice inspection and testing criteria. Westinghouse believes that a j p certain amount of inservice surveillance (leakage and fluid level checks)

V is required to provide added assurance that the snubber will continue to

~

function as intended during plant operation.

114. Issue 114: Seismic - Induced Relay Chatter This issue is currently covered in Unresolved Safety Issue A-46, " Seismic Qualification of Equipment in Operating Plants."

3 115. Issue 115: Enhancement of the Reliability of Westinghouse Solid State Protection System C)

Discussion The ATWS rule 10CFR50.62 requires for Westinghouse plants the implementation of a diverse ATWS mitigation system, Auxiliary (or ATWS]

Mitigating Systems Actuation Circuitry (AMSAC). The functions prescribed for AMSAC are turbine trip and the initiation of auxiliary feedwater, independent of the reactor trip system.

O As a consequence of the Salem ATWS event, Generic Letter 83-28

{ established the requirement for the automatic actuation of the shunt trip attachment of reactor trip breakers for W plants. Although this modifi-cation provides a significant increase in the reliability of the reactor O trip breakers and hence the reactor trip system, it had not been previ-ously pursued as an action which would significantly reduce the potential l

of an ATWS event during the extensive dialogue and study of the ATWS O

WAPWR-RC 5.5-91 AMENDMENT 3 5888e:1d AUGUST 1989

s issue. Further, it is believed.that other similar actions to increase g ,

the reliability of the existing reactor trip system for W plants have T l also not received such consideration.

i With respect to W plants with the solid state protection system (SSPS) design, recent failures of the undervoltage (UV) drive have raised concerns with regard to the susceptibility of the design to common mode and random failures of redundant components.

The failures of the UV driver suggest a higher probability of SSPS failure than that calculated during the ATWS rulemaking proceeding. The higher probability "of SSPS failure in turn would lead to a higher probability of ATWS and, as such, would represent a higher risk to the offsite population surrounding the affected plants. The affected plants '

are those W plants with SSPS.

3 Incorporation of additional diversity for the UV driver function is a possible solution to provide diversity for the function and reduce the probability of an ATWS event.

g SP/90 Response This issue is not applicable to SP/90 design because of the incorporation of the Integrated Protection System (IPS).

116. Issue 116: Accident Management This issue has not been further defined or prioritized by the NRC.

117. Issue 117: Allowable Outage Times for Diverse Simultaneous Equipment Outages This issue has not been f urther defined or prioritized by the NRC.

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WAPWR-RC 5.5-92 AMENDMENT 3 5888e:1d AUGUST 1989 l

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. 118. Issue 118: Tendon Anchorage Failure q%.J This issue has not been further defined or prioritized by the NRC.

- 119. Issue 119: Piping Review Committee Recommendations

't i

This issue primarily involves changes to Regulatory Rules and revisions i to Regulatory Guides and the Standard Review Plan.

The NRC has concluded this as a Regulatory Impact issue and currently requires no specific. action on the part of licensees. ,

120. Issue 120: On-Line Testability of Protection Systems This issue has not been.further' defined or prioritized by the NRC.

121. Issue 121: Hydrogen Control for Large, Dry PWR Containment 3 Discussion Issue 121 concerns potential rulemaking with regard to hydrogen control for LWRs with large, dry containments considering the greater inherent capability of these containment to accommodate large quantities of hydrogen.

Ongoing NRC experimental and analytical programs are intended to provide data supplementing the experiments being carried out at the Nevada Test Station (NTS), the experience on hydrogen burn during the TNI-2 accident, O and earlier hydrogen burn experiments in order to support a final recom-mandation on whether safety shutdown equipment is likely to survive a hydrogen burn. Experiments are being planned to determine the signifi-cance of factors not included in the NTS experiments (e.g., precondi-O tioning of equipment to simulate aging, enclosing equipment in conduits or protsetive heat shields, energized equipment). Experimental and l I

O 4APWR-RC 5.5-93 AMENDMENT 3 8888e:1d AUGUST 1989

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I s

analytical studies based in part.on NTS generated data will then be used 1 to determine the local environmental stresses on equipment such as the convective and radiant heat flux to be expected in a hydrogen burn.

In addition to the above, the staff intends to explore the possibility of forming local detonable , concentrations in large, dry PWRs and the probable consequences regarding containment and equipment survivability.

This has received attention from the standpoint of fundamental detonation l phenomena; however, the research effort is now being extended to develop {

the capability for predicting conditions in realistic configurations.

I' SP/90 Response The SP/90 Core and Containment Analyses as described in Section 5.5 of RESAR-SP/90 PDA Module 16, "Probabilistic Safety Study" indicate that there exists no scenario in which flammable hydrogen concentration is 3 reached. Even so, hydrogen ignitors are included in the SP/90 plant as described in Subsection 6.2.5 " Combustible Gas Control in Containment" of RESAR SP/90 PDA Module 10 " Containment System."

g The SP/90 is designed to maintain local hydrogen concentrations inside the containinent at less than 4 volume percent during a DBA and less than 10 volume percent during degraded core and/or a core melt accident. The hydrogen mixing is facilitated by redundant containment fan coolers designed for potential seismic and environmental conditions.

A manually actuated hydrogen igniter system is designed to limit uniformly distributed hydrogen concentration in the containment to 10%

during and following a degraded core and/or core melt accident.

Redundant igniter assemblies capable of maintaining a 1700*F surface temperature will be installed in enclosed areas throughout the containment.

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. . . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ )

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Large-Scale Hydrogen Burn Equipment Experiments (EPRI NP-4354) J s demonstrate the capability of equipment. qualified to operate during and I after DBA conditions to survive and perform properly during and after the high temperature spikes produced by hydrogen burn up to 13% volume j percent. j O Manually actuated redundant electric hydrogen recombiners are also provided with a capacity such that hydrogen concentration will not exceed l -

4 volume percent. Controls for the recombiner system are located outside-i the. containment and the heater portion of the system has been designed and tested for potential seismic and environmental conditions.

Information to the operator is provided by a redundant hydrogen monitoring system designed for potential seismic and environmental q cenditions and having a range of 0 to 20 volume percent.

122. Issue 122.2: Davis-Besse Loss of All Feedwater Event, Initiating Bleed 3 and Feed.

O Discussion The loss of all feedwater event at Davis-Besse on June 9, 1985 resulted in the formation of an NRC project team to investigate the event. The team's findings were published in NUREG-1154. As a result of NRC staff

]

review, several items were identified as candidates for short-term action j and were prioritized separately.

p Issue 122.2 deals with the adequacy of emergency procedures, operator

\

training, and available plant monitoring systems for determining the need to initiate feed-and-bleed cooling following loss of the steam generator heat sink. It is based upon Findings 10, 17 and 18 in Sections 6.1.1 and 6.1.2 of NUREG-1154. Essentially, the operators were reluctant to take j O the rather drastic step of initiating feed-and-bleed cooling, probably because they believed restoration of the AFW system was imminent. The fact that feed-and-bleed cooling releases primary coolant to the contain-l ment (implying an extensive shutdown for the purpose of decontamination)

O i

WAPWR-RC 5.5-95 AMENDMENT 3 l B888e:1d AUGUST 1989

O may also have influenced their actions. Finally, the normal control room instrumentation was inadequate to clearly inform the operators that feed-and-bleed was called for. The Safety Parameter Display System which would have displayed the necessary information was not operable. l The reactor vendors have provided their customers with feed-and-bleed procedures. Feed-and-bleed capability is not currently specifically required by the NRC although the techniques, benefits, and costs are being evaluated as part of USI A-45. Basically, feed-and-bleed cooling is a method of last resort which can avert core damage if main and auxiliary feedwater is lost and other methods of decay heat removal are

~

unavailable. For plants licensed without a PORV, the lack of feed-and-bleed capability was a significant issue and the need for a highly reliable AFW system was emphasized.

PRAs give considerable credit for feed-and-bleed cooling. A failure rate 3 of one or two percent is a typical assumption. However, the Davis-Besse event chronology leaves an impression that this failure probability may be overly optimistic.

In addition, it should be noted that, depending on specific plant design, there may be a fairly short time period in which feed-and-bleed cooling will be successful. If the plant operators delay too long before initiating feed-and-bleed cooling, their error may not be retrievable by later action.

This issue applies to all plants which can use feed-and-bleed techniques.

The solution is a matter of emphasis on safety vs. operation, training in O!

existing procedures, and possibly an upgrading of instrumentation at certain sites. In addition, the procedures themselves could be upgraded to make the criteria for initiation of feed-and-bleed cooling more direct and unambiguous, leaving less room for operator reluctance. (For example, in the case of Davis-Besse, basing the initiation of feed-and-bleed on hot leg temperature rather than on steam generator parameters ,

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WAPWR-RC 5.5-96 AMENDMENT 3 58BBe:Id AUGUST 1989 2

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.has been suggested.) Here, we will concentrate on ensuring that existing procedures are followed. The general technical aspects of feed-and-bleed decay heat removal will be addressed under USI A-45.

!$P/90 Response In case. of the RESAR SP/90 design feed-and-bleed operation has been j considered a viable means of removing decay heat from the start. l

'A Specific. features include:

U o SafetygradepressurizerPORVs(PowerOperatedReliefValves),

o A semi-closed circuit, whereby the PORV discharge is routed to the pressurizer relief tank, and from there to the in-containment Emergency Water Storage Tank (EWST), from which the HHSI_(high head safety injection) pumps take suction to return the inventory 3

to the reactor coolant system (RCS). Because of this feature, operator reluctance to initiate feed-and-bleed operation is expected to be significantly reduced, o RHR (residual heat removal) heat exchangers in the discharge of

, the HHSI pumps allow rejection of decay heat to the ultimate heat sink via the component cooling and service water systems.

The feed-and-bleed operation is described in more detail in Subsection 1.2.3.1.2.6 of RESAR-SP/90 PDA Module 1, ' Primary Side Safeguards

-System." Detailed Emergency Operating Procedure including feed-and-bleed operation will'be prepared at the FDA stage.

By providing designed-in-feed-and-bleed capability, the SP/90 plant is fully responsive to Generic Safety Issue 122.2.

Note that Issues 122.la through 122.le have been integrated into Issue 124, and Issue 122.3 has been assigned a low priority. Issue 122.3 will be evaluated at the FDA stage for its impact on the SP/90 design.

O WAPWR-RC 5.5-97 AMENDMENT 3 i

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123. Issue 123: Deficiencies in the Regulations Governing DBA and Single 4 Failure Criterion Suggested by the Davis-Besse Incident of June 9, 1985 This issue has not been further defined or prioritized by the NRC.

O 124. Issue 124: Auxiliary Feedwater Systems Reliability

/

I Discussion [

Operating experience As well as NRC staff and industry studies indicate that AFW systems continue' to fail at a high rate. These studies also indicate that plants with similar AFW system reliabilities (as calculated in accordance with the SRPguidance)donotnecessarilyexhibitsimilar AFW system availabilities. Based on these studies and on engineering judgment, the staff concluded that the PWR AFW system reliabilities 3 calculated in accordance with the SRP guidance may represent the relative reliability of AFW system hardware configurations for various plants, but do not represent the real availability of these crucial safety systems.

h In order to ascertain a high level of AFW system reliability and availability, the staff proposed a requirement that all operating plants demonstrate by PRA that their AFW systems are at least as reliable as I

-4 10 unavailability / demand after accounting for: (a) AFW system support systems, (b) common cause failures, or (c) operator errors. As input to the PRAs, each utility should use its plant-specific data if available. Such plant-specific data will reflect design faults, poor maintenance practices, and inadequate testing and surveillance and will indicate how well a particular plant is being operated, thereby g

identifying those plants that will need improvements.

l Because of the significance of the AFW system in reducing core-melt frequency, the staff has determined that all PWRs should meet the reliability criterion specified in SRP Section 10.4.9.

O WAPWR-RC 5,5-98 AMENDMENT 3 588Be:1d AUGUST 1989

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SP/90 Response The- SP/90 design includes both a' startup feedwater system and an emergency feedwater system which, in combination, perfor1n the functions

. (~'

of' auxiliary feedwater. systems currently ' employed in Westinghouse-

-' designed plants. The emergoney feedwater system, which serves as the safety system in loss of main feedwater transients has two set.or driven and two turbine driven pumps, takes suction from dedicated storage tanks and is actuated from diverse sources.

The startup feedwater system, though a control grade system, can provide backup feedwater delivery.

The unavailability of the EFWS over a 24-hour period is shown as 4 x 10

-6 in Table 3.7.2-2 of RESAR-SP/90 PDA Module 16, "Probabilistic Safety Study."

3 125. Issue 125.I.3: SPDS Availability Discussion This issue addresses the concern as to whether NRC requirements should be revised regarding SPDS availability.

Investigations subsequent to the TNI-2 accident have indicated a need for improving how information is provided to control room operators both during normal and abnormal conditions. TNI Action Plan Item I.D.2,

" Safety Parameter Display System (SPDS)," required that licensees install a system to continuously display information from which the plant safety status can be readily assessed. Generic Letter 82-33 (Supplement 1 to NUREG-0737) mandated that licensees install an SPDS. Licensee implementatiori of Item I.D.2 is reviewed and tracked as NPA F-09. The staff requirement imposed on the licensees does not contain specific reliability or availability requirements for the SPDS.

O WAPWR-RC 5.5-99 AMENDMENT 3 5888e:1d AUGUST 1989

_..._._m_._.__m_ _ _ _ -__ _. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ ._.m . _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ . _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ __

w Events such as those that occurred at TMI-2, Davis-Besse, Oconee, Rancho g Seco, and others may have been less severe if an operable SPDS had been W available to the operators. For the Davis-Besse event, "

...The inoperability of the SPDS and lack of adequate indications of steam I generator conditions contributed to the control room operators not knowing that the steam generators were dry, which resulted in their failure to follow the appropriate procedures."

The primary purpose of an available SPDS would be to display a full range of these important plant parameters in order to aid the control room personnel in determining the safety status of the plant during abnormal and emergency conditions and in assessing where abnormal conditions warrant corrective operator action to avoid a degraded core event.

Operators need all available parameter information for their decision-making in avoiding a degraded core event and that a properly functioning SPDS would result in a lower frequency of control room 3 operator errors and a corresponding reduction in core-melt frequency.

The parameters should provide, as a minimum, information about the following: reactivity control; reactor core cooling and primary system heat removal; reactor coolant system integrity; radioactivity control; and containment conditions.

For the resolution of this issue, the NRC assumed that improvements in design and hardware changes, as well as improved maintenance and test procedures, will be required to assure the availability of a properly functioning SPDS at all operating plants.

As of June 1988, a draft generic letter (

Subject:

Task Action Plan I.D.2 - Safety Parameter Display System) was prepared to provide all licensees, applicants, and construction permit holders the benefit of the staff's experience in order to aid them in acceptably implementing SPDS.

The generic letter describes various methods used by some licensees / g applicants to implement SPDS requirements in a manner that is acceptable W to the staff. Also documented are design features found to be unaccepta-ble as well as the staff's reasons for reaching this conclusion.

O WAPWR-RC 5.5-100 AMENDMENT 3 E888e:1d AUGUST 1989

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(N Included are guidelines and specific examples for providing an SPDS with acceptable reliability -(availability). The NRC ' concluded that the j solution to this issue has been identified and the issue is expected to )

i be reso1'ved with the issuance of the generic letter. {

E/90 Response The alarm system for the SP/90 control room, as described. in Subsection 18.3.2 of RESAR-SP/90 PDA Module 15 " Control Room / Human Factors O Engineering," will include- the attributes of a safety parameter display system (SPDS), and will' meet the intent of the requirements of NUREG-0696, " Functional Criteria for Emergency Response Facilities," and NUREG-0835, " Human Factors Evaluation Criteria for Safety Parameter Display Systems."

Issue 125.11.7: Reevaluate Provision to Automatically Isolate Feedwater from Steam Generator During a Line Break 3' Discussion During the course of the investigation of the event, it was pointed out that the benefits of AFW isolation are probably 'more than outweighed by the negative aspects of the feature.

The automatic isolation of AFW from a steam generator is provided to mitigate the consequences of a steam or feedwater line break. The isolation logic, usually triggered by a low steam generator pressure l signal, closes'all main steam isolation valves and also isolates AFW from

! the depressurizing steam generator. (The AFW flow is diverted to an intact steam generator.) The purposes of the AFW isolation are three-fold:

O (1) The. break blowdown is minimized. Shutting of AFW will not prevent the initial secondary side inventory from blowing down. However, the isolation will prevent continued steaming out of the break as decay heat continues to produce thermal energy.

WAPWR-RC 5.5-101 AMENDMENT 3 5888e:1d AUGUST 1989

(2) Overcooling of the primary system is reduced. As the depressurizing .

steam generator blows down to atmospheric pressure, the primary system is cooled down, causing primary coolant shrinkage and (if the event occurs near the end of the fuel cycle) a return to criticality, which adds a modest amount of thermal energy to the transient.  !

Shutting off feedwater to the faulted steam generator will reduce this effect, although once again the initial blowdown will be the dominant factor.

The significance of these first two considerations is in containment pressure. The containment is designed to accommodate a primary system blowdown' follow'de by decay heat boiloff (the large break LOCA). A steam or feedwater line break within containment might cause the containment design pressure to be exceeded if the AFW isolation were not present.

3 (3) The AFW isolation is needsd to divert AFW flow to the intact steam generator (s). For the case of a two-loop plant with a two-train AFW system, this is needed to meet the single failure criterion in supplying feedwater to the intact steam generator. (The situation becomes more complex for other cases, e.g. a four-loop plant with a three-train AFW system.) Note that, unless the line break is in the AFW line, core cooling would still meet the single failure criterion even without the isolation, since the faulted steam generator would still be capable of heat transfer.

In summary, the automatic isolation is needed only to help mitigate a relatively rare event (steam or feedwater line break) and even then is only remotely connected with sequences leading to core-melt.

In contrast, this isolation has definite disadvantages. If both channels of the controlling system were to spontaneously actuate during normal g operation, all AFW would be lost and the MSIVs would close. Most newer W plants use turbine-driven main feedwater pumps. Thus, main feedwater would be lost also. If the plant operators fail to correctly diagnose O

WAPWR-RC 5.5-102 AMENDMENT 3 BBBBe:Id AUGUST 1989

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• l and correct the problem, only feed-and-bleed cooling would be available to prevent core-melt. Similarly, if spurious AFW isolation were to occur

during the course of another transient, once again only feed-and-bleed cooling would be available to prevent core-melt.

O The long-term success of AFW for main -feedwater transients, steam generator tube ruptures, and small LOCAs may also be compromised. During controlled cooldown, the thresholds for automatic AFW isolation are crossed. Procedures call for operators to lock out the isolation logic as the steam generator pressure approaches the isolation setpoint. Under the circumstances, the accompanying distractions make it possible that the operators will forget to override the AFW isolation logic in the permissive window. Thus, AFW reliability in these scenarios any be significantly degraded.

The safety significance of this issue arises from the fact that the negative aspects involve accident sequences which have more_ frequent 3

, initiators, and more significant consequences, than those of the positive V . aspects.

A very straightforward solution has been proposed: simply disconnect the

. AFW isolation- valve actuators from the automatic logic and depend on plant procedures, i.e., have the operators close the AFW isolation valves (by, remote manual operation from the control room) in the event of a line break. These procedures would require careful verification of the existence of a line break before isolating a steam generator from AFW.

SP/90 Response In the case of the SP/90 plant, this Generic Safety Issue would be applicable to the Emergency Feedwater System (EFWS), which is described in Subsection 10.4.9 of RESAR-SP/90 PDA Module 6/8, " Secondary Side 4 Safeguards System / Steam and Power Conversion System." f

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WAPWR-RC 5.5-103 AMENDMENT 3 B888e:1d AUGUST 1989

h The SP/90 EFWS includes an automatic isolation feature between pairs of steam generators (i.e. A/D and B/C), which is actuated on high differen-tial pressure; this signal is an indiction of a faulted steam generator.

This feature prevents losing both pumps in one subsystem as a result of spilling to the faulted steam generator. Note that mass addition to the containment and pump runout are not a primary concern because cavitating venturis are installed in each line feeding the four steam generators.

h If the automatic isolation were to be actuated spuriously, reliability of the EFWS for non-faulted conditions would not be affected, since no EFW pump would be isolated from its associated steam generator.

The reliability concern contained in Generic Safety Issue 124 is therefore not applicable to the SP/90 Emergency Feedwater System design.

126. Issue 126: Reliability of PWR Nain Steam Safety Valves 3

This issue has been classified as a Licensing Issue and requires no action on the part of Westinghouse. It is listed in NUREG-0933 (June 30, 1988) as having been resolved.

127. Issue 127: Maintenance and Testing of Manual Valves in Safety-Related Systems.

Issue 127 has been assigned a low priority and will be evaluated at the FDA stage for its impact on the SP/90 design.

128. Issue 128: Electrical Power Reliability Discussion:

e Issue 128 is an integration of other active high and medium priority generic issues that were directly related to onsite electrical systems:

Issue 48, "LCO for Class 1E Vital Instrument Buses in Operating Reactors"; Issue 49, " Interlocks and LCO's for Class IE Tie Breakers";

O WAPWR-RC 5.5-104 AMENDHENT 3 5888e:1d AUGUST 1989

o and Issue A-30, " Adequacy of Safety-Related DC Power Supplies." A number

~ of other issues and concerns concern the onsite electrical power systems, to a somewhat lesser degree. Some of these issues include: USI A-17,

" Systems Interactions," and USI A-47, " Safety Implications of Control Systems."

With respect to possible resolution for future plants, consideration was given to the IEEE Standards and NRC endorsements by regulatory guides.

Specifically, the following were considered: IEEE 603 and its referenced O' standards, and IEEE 308 and its referenced standards.

Any revisions to Regulatory Guides resulting from this issue may have an impact on future plants.

SP/90 Response The SP/90 plant DC and instrument AC power supplies are described in 3

- Section 8.3 "Onsite Power Systems" of RESAR-SP/90 PDA Module 9

" Instrumentation and Controls and Electric Power."

The Class 1E portion consists of four independent 125V batteries and four independent 120V vital AC buses; the latter supply power to the various cabinets of the integrated protection system as well as other Class 1E equipment (e.g. post-accident monitoring system, multipliers, etc.).

The non-Class 1E portion consists of one 250V battery and one 125V battery. The former supplies motors, while the latter is associated with redundant non-Class IE 120V instrument buses which supply power to the integrated control system.

The Class IE DC and Instrument Power supplies for the SP/90 plant are designed to be highly reliable. Separate non-Class 1E DC and Instrument Power supplies have been included to eliminate potential system inter-action issues. At this time, Generic Safety Issue 128 is rather O

I WAPWR-RC 5.5-105 AMENDMENT 3 1

E888e:1d AUGUST 1989

1 undefined. The SP/90 plant meets the requirements of IEEE Standards 308-1980 and 603-1980 which are being considered for endorsement by the j NRC; Westinghouse will review future revisions (if any) to related I i

Regulatory Guides and, if necessary, make appropriate changes to the {

SP/90 design. l 129. Issue 129: Valve Interlocks to Prevent Vessel Drainage During Shutdown i Cooling This issue has not been further defined or prioritized by the NRC.

130. Issue 130: Essentia5ServiceWaterPumpFailuresatMultiplantSites Discussig This issue relates to the reliability of the emergency service water 3 (ESW) system of multiplant sites having two ESW pumps per plant with crosstie capabilities. Studies will be performed to determine if the ESW system reliability at such plants is adequate. The potential benefits of adding an ESW pump at each plant or a swing ESW pump, modification of h

plant operating procedures and plant technical specifications will be evaluated. The effect on system reliability of modified technical specifications on the LCO's for the ESW pumps and alternate means of providing service water will also be considered.

$_P/90 P Response The SP/90 design provides four ESW pumps per unit. This issue is not applicable to SP/90.

131. Issue 131: Potential Seismic Interaction Involving the Movable In-Core Flux Napping System in Westinghouse Plants g

This issue has not been further defined or prioritized by the NRC.

O WAPWR-RC 5.5-106 AMENDMENT 3 588Be:1d AUGUST 1989

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132. Issue 132: RHR Pumps.Inside Containment' This issue has.not been further defined or prioritized by the NRC.

133. Issue 133: Update Policy Statement on Nuclear Plant Staff Working Hours This issue relates to staffing requirements for operating plants and is not an issue for the SP/90 PDA. The issue has been classified a Licensing Issue and requires no Westinghouse action.

134. Issue 134: Rule on Degree and Experience Requirements for Senior Operators Discussion This contemplated rulemaking action is due to a commission decision to enhance the levels of engineering and accident management expertise on 3 shift.- This is being done to further ensure the protection of the health

.and safety of the public by having personnel on shift with enhanced qualifications.

- The contemplated rule would require that applicants for licenses as a senior operator (50) of a nuclear power plant hold a baccalaureate degree in engineering or physical science from an accredited institution. Other baccalaureate degrees from an institution may be accepted on a case-by-case basis. The current requirement (for candidates with a baccalaureate degree)oftwoyearsofresponsiblenuclearpowerplant experience, would be amended t'o require at least one of the two years of operating experience'be with a similar commercial nuclear reactor operating at greater than twenty percent power.

The contemplated degree rule is related to the commission policy statement on engineering expertise on shift (50 FR 43621, 10/28/85).

This policy statement is satisfied by providing engineering and accident management expertise on shift through the separate shift technical advisor (STA) or by combining the STA/SO functions.

5.5-107 AMENDMENT 3 WAPWR-RC 5888e:1d AUGUST 1989

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SP/90 Response 1

This issue is associated with utility staffing requirements and is not 01' applicable to Westinghouse in relation to the SP/90 design.

135. Issue'135: Steam Generator and Steam Line Overfill Issues h l Discussion Steam generator overfill and its consequences have received staff /

industry attention because of the frequency and severity of such events.

A number of ongoing activities are concerned with steam generator overfill related issues, An integrated work plan was developed to bring these activities to a conclusion.

3 This work plan for Issue 135 includes consideration of activities related to1)steamgeneratortubeinspections, 2) radiological consequences of steam generator tube rupture (SRP 15.6.3), 3) open NRC staff actions regarding steam generators (Issue 67), and 4) the effects or water hammer, overfill, and water carryover on secondary systems and connecting systems.

The coordination of results of the different tasks will provide a basis for the staff to develop a' position on offrite dose, operator action time, and tube integrity. Water hammer mitigation studies will be carried out to give the staff a better understanding for developing positions on water hammer in main steam lines and operability of valves and other components.

SP/90 Response O

Issues related to steam generator tube integrity are addressed in USI A-3; of particular concern with Issue 135 is the impact of steam generator overfill.

O WAPWR-RC 5.5-108 AMENDMENT 3 i BBBBe:Id AUGUST 1989 1

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In December 1980, the Office for Analysis and Evaluation of Operational'

-Data (AE00) issued the ' report "AEOD Observations and Recommendations

]

concerning the problem of. Steam Generator Overfill and Combined Primary I and Secondary Side Blowdown" to discuss the principal safety considera-tions related to steam generator overfill. Several of the issues 1 discussed in the report include the consequences of hydraulic forces in the steamline, excessive dead weight loads in the steamlines, failure of steamline valves to ressat, loss of emergency feedwater pump turbine, n

V increased probability of steam generator tube rupture and acceleration of accumulated water in the steamline.

Each steam generator on the SP/90 is provided with 2 out of 4 high high steam generator water level logic. In addition, each generator also includes a line from the upper shell to the EWST containing two parallel isolation valves. These steam generator overfill valves are normally closed, fast acting solenoid operated, globe valves. The capacity of one valve is sufficient to prevent steam generator overfill assuming the 3 rupture of one steam generator tube and maximum Emergency Feedwater

. System (EFWS) flow. The valves automatically open on a high high steam generator water level signal, and close automatically when the level-has dropped sufficiently. Finally, each steam generator has an overfill

, block valve that is in series with the two overfill control valves.

These overfill block valves are normally open, motor operated gate valves which provide a redundant means of reclosing the overfill lines.

Hence, in addition to the steam generator feedwater isolation and control valves receiving an isolation signal on high high steam generator water level and a safety injection signal, a steam generator overfill proter tion system is designed to divert flow to the EWST if the high high level setpoint is reached.

Although the features described above will reduce the probability of the occurrence of a steam line overfill, Westinghouse will evaluate the structural adequacy of the main steam lines and associated supports under water filled conditions as a result of steam generator tube rupture.

O WAPWR-RC 5.5-109 AMENDMENT 3 8888e:1d AUGUST 1989

This evaluation will assume that the main steam line is water filled from g the steam generator out and the first isolation valves are at main steam W system design pressure and temperature. This load condition will be evaluated against Service Level D allowables.

136. Issue'136: Storage and Use of Large Quantities of Cyrogenic Combustibles On Site This issue involves the on-site storage and use of large quantities of liquified combustible gases, which pose the potential for severely damaging reactor safety related structures and/or equipment. This is not applicable to the 'SP/90 design. The issue was classified as a Licensing Issue and is considered to be resolved (NUREG-0933, June 30,1988).

137. Issue 137: Refueling Cavity Seal Failure 3 This issue has not been further defined or prioritized by the NRC.

138. Issue 138: Deinerting Upon Discovery of RCS Leakage g This issue has not been further defined or prioritized by the NRC.

I ~

139. Issue 139:' Thinning of Carbon Steel Piping in LWR's This issue deals with an increased need for utilities to periodically assess the integrity of piping systems to reduce me r sk of future injury to plant personnel or damage to e' paer.; caused by erosion / corrosion-induced wall thinning.

The issue has been classified as a Regulatory Impact issue and was considered to be resolved with the issuance of guidelines on erosion / corrosion in single phase piping as developed by NUMARC and found g acceptable by Staff. W O'

WAPWR-RC 5.5-110 AMENDMENT 3 E888e:1d AUGUST 1989 l

NOTE: The following generic issues (Issues 140 through 145) have not been further defined or prioritized by. the NRC and require no action on the part of Westinghouse. .They will be evaluated during the SP/90 final design stage.

140. Issue ~140: Fission Product Removal by Containment Sprays or Pools 141. Issue 141: Large Break LOCA with Consequential SGTR 3

142. Issue 142: Leakage Through Electrical Isolators 143. Issue 143: 'Availabil'ity of Chilled Water Systems 144. Issue 144: Scram Without a Turbine / Generator Trip 145. Issue 145: Improve Surveillance and Startup Testing Programs O

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O WAPWR-RC 5.5-111 AMENDMENT 3 5888e:1d AUGUST 1989

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- 6.5 IE BULLETINS AND LICENSING ' ISSUES 6.5.1 IE BULLETINS l

The NRC Office of Inspection and Enforcement monitor nuclear plant compliance during construction and operation by reviewing various mandatory reports, and by site inspections. Their enforcement actions must be based on established documented requirements. In addition to citations and punitive actions they issue Bulletins, k

The subject of a Bulletin must be generic or potentially generic for a set of plants. A Bulletin can mandate immediate corrective action; however, most Bulletins mandate that the utility respond with a description of their status and intention on the subject.

The following IE Bulletins include only those with subjects of potential impact. to the design or design interface with the Westinghouse Advanced Pres-I surized Water Reactor.

1

1. IE Bulletin 79-01, 01A, 01B, Environmental Qualification of Class 1E Equipment Discussion Licensees are required to itientify and document all Class IE electrical equipment required to function under accident conditions and submit writ-ten evi,dence of its ability to function. In addition they must submit a master list of all installed safety-related electrical equipment and 7

' report on documentation that shows qualification for the adverse environ-ment. Also, -they must pursue programs to establish qualification er replacement of unqualified equipment.

t WAPWR Response This issue and its impact on the WAPWR design is fully discussed in 5ection 4.0, item 14.

O WAPWR-RC 6.5-1 NOVEL 13ER, 1983 2318M:1d

s

2. IE Bulletin 79-02, Revisions 1 and 2, Pipe-Support Base-Plate Designs Using Concrete-Expansion Anchor Bolts Discussion Structural failures of anchor bolts systems at the support plate interface with masonry walls resulted in an NRC mandate to: ,

1

1) Recalculate individual base plate and bolt design loads.

2) Determine the load capabilities of concrete anchors and the factor of safety.

3) Test anchor bolt capabilities documentation if documentation does not verify an adequate safety margin.

4) Report on the verification and discrepancies.

5) Provide schedules for redesign and corrections.

WAPWR Response O

The concerns discussed in IE Bulletin 79-02 have been addressed by Appendix B of the ACI 349 Code for Concrete Nuclear Structures.

3 Westinghouse has committed to this code in Subsections 3.8.3 and 3.8.4 of RESAR-SP/90 PDA Module 7, " Structural / Equipment Design."

3. IE Bulletin 79-05, 05A, 05B, OSC, Nuclear Incident at Three Mile Island Discussion The accident ~at Three Mile Island Nuclear Power Plant, Unit 2 (TMI-2) resulted in major core damage with minor radioactive releases to the environment. Improper valve positioning prevented early auxiliary-g feedwater system function. In addition, a power operated relief valve failed to close. The nature of the pressurizer instrumentation caused the operators .to believe that the pressurizer water level was high adding to the maloperation and continuation of the core damage sequences.

WAPWR-RC 6.5-2 AMENDNENT 3

'2318M:1d AUGUST 1989

O y

WAPWR Response There are no masonry walls identified at this time in the design of the SP/90 plant. If such walls are required at a later date, then designs 3

f)

V will be prepared complying with the " Interim Criteria for Safety Related

.Nasonry Walls" given in Appendix A to SRP 3.8.4.

11. IE Bulletin 80-18, Maintenance of Adequate Minimum Flow Thru Centrifugal

. Charging Pumps Following Secondary Side High Energy Line Rupture Discussion Under certain conditions the centrifugal charging pumps (CCPs) could be damaged due to lack of minimum flow before safety injection (SI) termina-tion criteria are met. The particular circumstances that could result in damage vary somewhat from plant to plant, but involve unavailability of the pressurizer power operated relief valves (PORVs), with operation of one or more CCPs repressurizing the reactor during SI- following a second-ary system high energy line break. Since the SI signal automatically iso-lates the CCP mini-flow return line, the flow through the CCPs is deter-mined by the individual pump characteristic head vs. flow curve the pres-

. surizer safety valve setpoint, and the flow resistances and pressure lo:-

ses in the piping and in the reactor core. Flow at or near shut-off '.ead {

may not be adequate to insure pump cooling, and resulting pump damage could violate design criteria before SI termination criteria are met.

J The NRC required utilities to calculate the design capability for minimum charging pump flow and implement as necessary the modifications to equip-ment or procedures to ensure flow.

O WAPWR-RC 5.5-7 AMENDMENT 3 2318M:Id AUGUST 1989 l

1 i

s HAPWRResponse l O

For the EAPWR design, the scenario described above does not apply since the charging pumps are not used as safety injection pumps. Also, the safety injection miniflow valves are not automatically closed on a safety I injection signal. Thus, there are individual miniflow paths provided for each safety injection pump which provide continuous miniflow for safety ]

injection operation. )

i

12. IE Bulletin 80-24, Prevention of Water Damage Due to Water Leakage Inside Containment i

Discussion f A flooded condition within containment resulted from a combination of

1) fan cooler service water leaks, 2) inoperable containment sump pumps,

3) lack of attention to sump level lights, 4) lack of sump level range

- and alarm instrumentation, 5) limited range and calibration error in atmosphere moisture measuring system, 6) hold-up tanks not uniquely dedicated to containment sump discharge, 7) local water level indicators in the fan cooler basin not calibrated, 8) no water level indicators in the reactor vessel cavity pit and, 9)the pit sump discharged to the containment floor.

The NRC required utilities to describe 1) all "open" cooling water ser-vices in containment, 2) history, and 3) isolation testing capabilities.

Also, they were to implement means for the detection of water accumula-tion and a positive means for flow indication from the containment sumps.

EAPWRResponse EAPWR design will address IE Bulletin 80-24 in regard to detection of water accumulation in the containment. The WAPWR design with the EWST g

addresses major leaks (see Section 4.0, item 21).

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WAPWR-RC 6.5-8 NOVEMBER, 1983 2318M:1d j j l il I

REQUEST FOR ADDITIONAL INFORMATION RESAR SP/90 )

!]

260.1 Describe measures which ensure that Westinghouse retains and exercises responsibility for the overall quality assurance progr{tm for the design and design-related activities for RESAR-SP/90. (IA1)1/

(q)

RESPONSE

H has added a transition sentence at the end of the first paragraph of Section 17.1.1 of HCAP 8370/7800 that describes quality assurance  !

responsibility. Additionally it is noted that by definition and use of HCAP 8370/7800, H retains and exercises responsibility for the I

quality assurance program for RESAR-SP/90 design and design-related activities. Reference NUREG-0800 (1A1) 260.2 Identify and describe the criteria for determining the size of the organization (s) responsible for design verification and other quality assurance activities of the RESAR-SP/90. (IAS)

)

RESPONSE

HCAP 8370/7800 as written provides that each Business Unit or subtier division General Manager (as applicable) is responsible for establishing and implementing a quality assurance program that meets the requirements of the Quality Assurance Plan. This plan also states that responsibility for establishing and assuring the f3

- effective implementation of commitments to the plan is assigned by

() the Vice President and/or General Manager of ESBU and NFBU. Staffing for all the needed disciplines including QA organizations and l inspection staffs is included as part of normal commitment implementation and does not require any special procedures.

Q Reference NUREG-0800 (IAS) 1/ Alpha-numeric indicators refer to acceptance criteria from Section 17.1 /

of the Standard Review Plan, NUREG 0800, July 1981.

! HAPHR-QA 260-1 AMENDMENT 3 0011D:1D AUGUST 1989 L _ __ _ __

-Q 260.3 Describe provisions for resolving disputes involving quality which result from a difference of opinion between quality assurance &

W personnel and personnel from other departments / divisions. (185)

RESPONSE

A clarification regarding dispute resolution has been added to the third paragraph of Section 17.1.1 of HCAP 8370/7800. These disputes, if any, are handled within the framework of the line organizations

~

involved using normal good business practices. No special procedures are used except for the provisions already existing in HCAP 8370/7800. Reference NUREG-0800 (1B5) 260.4 Describe measures which ensure that designated quality assurance individuals are involved in day-to-day SP/90 activities important to safety (that is, the quality assurance organization routinely attends and participates in work schedule and status meetings to assure they are kept abreast of work assignments and that there is adequate quality assurance coverage relative to procedural and inspection controls, acceptance criteria, and quality assurance staffing and qualification of personnel to carry out quality assurance assignments). (IB6)

RESPONSE

The subject of participation in work schedule and status meetings has been addressed in the last paragraph of Section 17.1.1 of HCAP 8370/7800. It should h noted however that the nature and frequency of participation is dependent upon the practice at a particular location and may not be on a daily basis. Reference NUEG-0800 (186) 260.5 Describe the scope of the Westinghouse quality assurance program.

(2A1)

BAPHR-QA 260-2 AMENDMENT 3 O

0011D:1D AUGUST 1989

RESPONSE

The scope of the H quality assurance program is as described in NCAP 8370/7800, Section 17.1.1 and 17.1.2. This was reviewed during the review meeting with staff on June 16, 1988, and found to O adequately address the commitments of NUREG-0800 (2A1).

V 260.6 Provide a commitment in the RESAR-SP/90 SAR that Westinghouse will meet the latest NRC-accepted version of HCAP-8370. Also, update the .

z commitment in Table 17.1 to the latest revision of Regulatory Guide s

• 1.28. (283)

RESPONSE

RESAR-SP/90 SAR will be revised to specifically address the latest ,

approved revision of HCAP 8370/7800 and will include commitment to Revision 3 of U. S. NRC Regulatory Guide 1.28. Reference NUREG-0800 (283).

O { l 260.7 Identify Westinghouse procedures reflecting that quality assurance ]

regulatory guides, general design criterion 1 of Appendix A to 10 CFR 1 Part 50,10 CFR Part 50 Subsection 50.55(a), and each criterion of l 10 CFR Part 50, Appendix B will be met by documented procedures. j RESPONSE: ,

It was agreed during the June 16, 1988 H/NRC meeting, after reviewing i HCAP . 8370/7800 as relates to NUREG-0800 (2B4), that the subject was adequately covered within the various sections of HCAP B370/7800.

?

HAPHR-QA 260-3 AMENDHENT 3 0011D:10 AUGUST 1989

s 1

260.8 Provide a summary description of how responsibilities and control of quality-related activities will be transferred from Westinghouse to the licensee during the phasecut of design. (2C3)

RESPONSE

HCAP 8370/7800, specifically Sections 17.1.1, 17.1.3, and 17.1.6 addresses the subject of this question for NUREG-0800 (2C3). For added clarification however, the last paragraph of Section 17.1.17 has been revised to show design document and quality assurance records maintenance. Transfer is handled on a case-by-case basis based on specific contract requirements.

260.9 In the area of personnel training and qualification, clarify that;

a. Proficiency tests are given to those personnel performing activities affecting quality, and acceptance criteria are developed to determine if individuals are properly trained and qualified;

b. Certificates of qualification clearly delineate (a) the specific functions personnel are qualified to perform and (b) the criteria used to qualify personnel in each function;

c. Proficiency of personnel performing activities affecting quality is maintained by retraining, reexamining, and/or decertifying.

(2D)

)

RESPONSE

The subject of this question was found to be adequately addressed in Section 17.1.2 of HCAP 8370/7800 during the June 16, 1988 H/NRC meeting. Reference NUREG-0800 (20)  !

O 260-4 AMENDMENT 3 O

HAPHR-QA AUGUST 1989 0011D:1D

i n

260,10 Describe internal and external design interface controls, procedures, (T and lines of communication among participating design organizations d and across technical disciplines to assure that structures, systems, and components are compatible geometrically, functionally, and with processes and environment. (3D)

RESPONSE

The third paragraph of Section 17.1.3 has been revised to more clearly aodress the subject of this review question that relates to f NUREG-0800 (3D).

260.11 Describe measures which ensure a documented check to verify the dimensional accuracy and completeness of design drawings. (3E1) i

RESPONSE

The subject of this question was found to be adequately addressed in

[ Section 17.1.3 of HCAP 8370/7800 during the June 16, 1988 H/NRC meeting. Reference NUREG-0800, (3E1).

260,12 Describe measures which ensure that design drawings are reviewed by the quality assurance organization to assure that they are prepared, reviewed, and approved in accordance with company procedures and that they contain the necessary quality assurance requirements such as inspection and test requirements, acceptance requirements, and the extent of documenting inspection and test results. (3E2)

RESPONSE

The subject of this question was previously covered in NCAP 8370/7800

/ Section 17.1.1. Clarification was added to the fourth paragraph of HCAP 8370/7800, Section 17.1.3. Reference NUREG-0800 (3E2) 260-5 AMENDMENT 3 HAPHR-QA AUGUST 1989 0011D:1D

t i  :

s-I 260.-13 Describe measures which ensure that procedural control is established for design documents that reflect the commitments of the SAR; this &

control differentiates between documents that receive formal design verification by interdisciplinary or multi-organizational teams ' and W

those which can be reviewed by a single individual (a signature and 1 date is acceptable documentation for personnel certification).

Design documents subject to procedural control include, but are not limited to, specifications, calculations, computer programs, system descriptions. SAR when used as a design document, and drawings including flow diagrams, piping and instrument diagrams, control logic diagrams, electrical single line diagrams, structural systems for major facilities, site arrangements, and equipment locations.

Specialized reviews should be used when uniqueness or special design ,

considerations warrant. (3E4c) l RESPONSE; HCAP 8370/7800 Section 17.1.3 describes measures for procedural control for design and verification. By definition a design document  !

is a commitment to the SAR and it was agreed that revision of HCAP 8370/7800 is not necessary. It was further pointed out during the June 16, 1988 B/NRC meeting that Table 17-1 of HCAP 8370/7800 )

provides necessary clarifications to regulatory Guide 1.64 Rev. 2. $

Reference NUREG-0800 (3E4c) 260.14 When design verification is by test, describe measures which ensure that such testing is performed as early as possible prior to installation of plant equipment, or prior to the point when the j installation would become irreversible. (3ES) j 1

RESPONSE: <

HCAP B370/7800 Sections 17.1.3, 17.1.7, and 17.1.8 address the O j subject of design verification by test and procedures for tracking j contingencies such as incomplete tests. Section 17.1.3, seventh paragraph, has been revised to add clarification related to this commitment. Reference NREG-C800 (3E5) l l

BAPHR-QA 260-6 AMENDMENT 3 00110:1D AUGUST 1989

260.15 The first paragraph of Section 17.1.3 of HCAP 8370/7800 uses the term 7 .

" quality standards." Clarify if this term includes such things as

.. ! codes and industrial standards, test and inspection requirements, and special process instructions. (4A2)

.g- RESPONSE:

(

Clarification regarding this question has been provided by revising Section 17.1.3, first paragraph and Section 17.1.4, first paragraph.

Reference NREG-0800 (4A2) 260.16 The second paragraph of Section 17.1.6 of HCAP 8370/7800 indicates that Westinghouse may use master lists to identify the current versions of documents. Clarify how such a list is updated and distributed. (682)

RESPONSf:

f-- The second paragraph of Section 17.1.6 has been revised to indicate G)/ updating. Reference NUREG-0800 (6B2) 260.17 Clarify whether procedures for supplier verification specify the characteristics or processes to be witnessed, inspected or verified, and accepted; the method of surveillance and the extent of documentation required; and those responsible for implementing these procedures. (7A2) s RESPONSE:

i This question has been resolved by revision of the first paragraph of Section 17.1.7. The second sentence has been revised and an n additional sentence added immediately following that clarifies the co;mitment. Reference NUREG-0800 (7A2)

( BAPHR-QA 260-7 AMENDMENT 3 00110:10 AUGUST 1989 l

l

i

% i 260.18 Position 3 on Regulatory Guide 1.123 clarifles controls on  !

Westinghouse procurement of ~ standard hardware and catalog items.  !

Describe measures which ensure that, for such procurement, special  !

quality verification requirements shall be established and described to provide the necessary assurance of an acceptable item. (7B4) 1 l

EESPONSE:

Table 17-1 of HCAP 8370/7800, Regulatory Guide 1.123 Rev. 1, l

paragraph 3 (Clarification) has been revised to include testing of the " item" as required. Reference NUREG-0800 (7B4) 260.19 Describe measures which ensure that the Westinghouse quality assurance organization is involved in the qualification activities of special process procedures, equipment, and personnel to assure they are performed satisfactorily. (9B1).

RESPONSE

A sentence clarifying Section 17.1.9 as relates to this review question, has been added. Reference NUREG-0800 (9B1) 260.20 Describe measures which ensure that the acceptability of inspection results is determined by a responsible individual or group. (10C3)

RESPONSE

The last paragraph, last sentence of Section 17.1.10 has been revised to address this review question. Reference NUREG-0B00 (10C3)

O l

BAPHR-0A 260-8 AMENDMENT 3 0011D:1D AUGUST 1989

L 260.21 The second paragraph of Section 7.1.11 of HCAP 8370/7800 lists some test prerequisites. Describe measures which ensure that test

]v prerequisites have been met before doing the test. (11B1)

! RESPONSE:

O In response to' this question it is noted that " test prerequisites" are completed prior to test by definition. It was agreed, during the June 16, 1988 review meeting, that HCAP 8370/7800 does not require revision to resolve this review question. Reference NUREG-0800 (11B1) 260.22 Describe quality assurance and other organizations responsibilities for'the Westinghouse calibration program. (12.2)

RESPONSE

The first paragraph of Section 17.1.12 has been revised to clarify rm quality assurance responsibility. Additionally a sentence at the end of the second paragraph has been added to indicate the use of audits -

to verify the effectiveness of the calibration program. Reference NUREG-0800 (12.2).

260.23 Describe the review and concurrence of calibration procedures and identify the organization responsible for these activities. (12.3)

RESPONSE

O i

The response to review question 260.22 above and the commitments provided in Section 17.1.5 of HCAP 8370/7800 resolves this question.

m Procedures identify the organizational responsibilities for calibration activities. Reference NUREG-0800 (12.3) )

}

( BAPHR-QA 00110:1D 260-9 AMENDMENT 3 AUGUST 1989 l

_ ____-_-___-__-_a

s f

260.24 Clarify-whether measuring and test equipment is labeled or tagged to indicate due date of the next calibration. If not, describe the method of control. (12.5) l RESPONSE.

In Section 17.1.12, the last sentence of the first paragraph has been 04 .

{

revised to clarify the subject of labeling, tagging, and etc. It l should be noted that labels or tags are not always used to indict,te 1 when the next calibration is required. This is because it is not always practical depending on the nature and use of the tool or equipment being calibrated. There are always however, procedures j that are used for calibration control. Reference NUREG-0800 (12.5) 260.25 Identify the Westinghouse management (by position title, for example) authorized to:

a. allow calibration with less than four-to-one accuracy ratios.

(12.6)

b. allow calibrating standards to have the same accuracy as standards being calibrated. (12.7)

RESPONSE

In Section 17.1.12, the first paragraph of HCAP B370/7800 has been revised to clarify the use and approvals of transfer ratios.

In-place procedures at the various divisions and suppliers vary somewhat but do provide for appropriate and knowledgeable personnel to authorize smaller ratios and accuracies if needed because of state of the art considerations or end use applications. Reference NUREG-0800 (12.6 and 12.7)

BAPWR-QA 260-?O AMENDMENT 3 00110:1D AUGUST 1989 l

260.26 Describe measures which ensure that procedures for altering the

f. ') . sequence of required tests, inspections, and other operations require V the same controls as the original review and approval. (14.3)

RESPON3E:

( )

V A clarifying sentence has been added to Section 17.1.14 regarding altering the ~ sequence of required tests, inspections, and other operations important to safety. Reference NUREG-0800 (14.3) 260.27 Identify the organization (s) responsible for documenting and preventing the inadvertent use of nonconforming items. (14.4)

RESPONSE

A clarifying sentence has been added to Section 17.1.15 regarding organizations responsible for dispositioning nonconforming items.

, Reference NUREG-0500 (14.4) 260.28 Describe quality assurance and other organizational responsibilities

. for the definition and implementation of activities related to nonconformance control. This includes identifying those individuals or groups with authority for the disposition of nonconforming items.

(15.2)

RESPONSE

(

The clarification for 260.27 above and an additional sentence added at the end of the first paragraph of Section 17.1.15 of HCAP 8370/7800 provides resolution of this review question regarding quality assurar.ce and other organization responsibilities. As v indicated in NCAP 8370/7800 procedures describing responsibilities and activities related to nonconformance control have been 260-11 AMENDMENT 3 HAPWR-QA 0011D:1D AUGUST 1989

s I established at the various divisions and suppliers to address the broad variety of products and services. Quality Assurance routinely l j

includes hardware and software nonconformance control during audit and surveillance activities. Reference NUREG-0800 (15.2). q 260.29 Describe quality assurance and other organizational responsibilities for the definition and implementation of activities related to l quality assurance records. (17.2) i l

RESPONSE

As indicated in the June 16, 1988 H/NRC review neeting, HCAP 8370/7800 includes an adequate description of the H commitments regarding Criterion 17 for 10CFR50 App. B and NUREG-0800 (17.2).

Provided below is requested clarification of the information provided in HCAP 8370/7800 as already accepted by the U.S. NRC. The information provided relates to the quality assurance and other j organizational responsibilities for definition and implementation of activities related to quality assurance tecords.

H has in-place a quality assurance records plan generated by Quality Assurance that describes the overall records program and the interface with H Corporate Records Center (CRC). The H CRC has been found acceptable as a records retention facility by the NRC as indicated in Table 17-1 of HCAP 8370/7800. As part of the i

requirements of this plan each depertment generates a records flow schedule that describes the specific nature, retention method and times, location, and any special instructions for quality assurance records generated by the department. The flow schedules are maintained and kept current by responsible Department Management. In addition, departmental procedures describe the requirements for the type of records requiring protection, retention, and/or transmittal to the customer (applicant). The records requirements and methods of HAPHR-QA 00110:1D 260-12 AMENDMENT 3 AUGUST 1989 9l l l

l l

j

the' departments and CRC are audited by quality assurance in accordance with internal audit schedules and procedures.

Subtier suppliers are also required to have quality assurance records protection and retention procedures. These procedures are audited during normal audit and surveillance activities. Quality Assurance data packages that include records required by specifications and procurement documents, are reviewed at the suppliers facility as necessary. Depending on contract requirements quality assurance data j packages are sent directly to the customer and/or M at the time of

'- shipment of the item for input to the records retention program.

Suppliers are required to maintain all quality assurance records for periods specified by procurement documents, after which they must notify H for disposition instructions.

Records retention times and procedures are determined from the criteria of standards ANSI N45.2.9, ASME Code Sections III-XI, ANSI /AS'ME ' NOA-1, Regulatory Guides 1.28 and 1.88, and any other c specific requirement as indicated by the customer. The revisions of Q these standards to which H has committed are noted in Table 17-1 of HCAP 8370/7800. Reference NUREG-0800 (17.2) 260.30 Describe measures which ensure that audit data are analyzed by the quality assurance organization and the resulting reports indicating any quality problems and the effectiveness of the quality assurance program, including the need for reaudit of deficient areas, are reported to management for review and assessment. (1881)

O Q RESPONSE:

As described in NCAP 8370/7800 H has procedures ensuring that audit data is analyzed by the quality assurance organization (who normally perform all audits) and that resulting reports indicating any quality problems and the effectiveness of the quality assurance program, l

BAPHR-QA 260-13 AMENDMENT 3 0011D:1D AUGUST 1989 1

, i including the need for reaudit, are reported to management for review and assessment.

l t

Routine review and approval of internal and supplier audits is  !

accomplished by management. Assessment of the quality assurance program is generated for and reviewed by division management. The assessment includes trending of previous audit data and recommenda-tions to correct recurring problem areas. HCAP 8370'7800 and supporting procedures adequately define the method for meeting this commitment. Reference NUREG-0800 (IBB1) 260.31 HCAP 8370/7800 should be revised to reflect the current Westinghouse organization and related quality / quality assurance responsibilities (see Westinghouse letter NS-NRC-87-3220, PA-87-519, dated April 29, 1987 for latest information received by the NRC).

RESPONSE

HCAP 8370/7800 has been revised to indicate the current Westinghouse organization and is reflected in Figures 17-1, 17-2, and 17-3.

Notification of organizational changes was made to the NRC by H

, letter NS-NRC-88-3369 (PA-88-996) dated August 12, 1988. As required by Section 17.1.1 of HCAP 8370/7800 Westinghouse will provide notification should additional changes be made. This letter referenced above superceeded letter NS-NRC-87-3220 (PA-87-519) referenced in review question 260.31.

4 260.32 The second paragraph of SAR Section 17.1 refers to a NAPHR Integrated PDA submittal. He understand there has not been and will not be an I integrated PDA for RESAR SP/90; therefore this paragraph should be revised or deleted from the SAR.

O1 1

l BAPHR-0A 260-14 AMENDMENT 3 O

0011D:1D AUGUST 1989 I

l L____________________

q RESPONSE:

V This section will be modified to remove the reference to an integrated PDA submittal, and to reflect incorporating the latest revision of HCAP-8370/7800.

O<

260.33 SAR Table 3.2-1, "C'f unification of structures, Systems, and Components for the Eeactor System," has a column headed

• Quality (m Assurance." This column shows either "HQCS-1" or "HQCS-2" for the systems / components in the table . and it references a footnote which

(^ states: "HQCS-1 and HQCS-2 satisfy the requirements of 10 CFR 50 Appendix B." Clarification is needed since neither SAR Section 17 nor HCAP-8370 address the HQCS documents.

RESPONSE

i HQCS-1 and HQCS-2 are preplanned q'uality systems requirements specifications that establish requirements for subtier supply quality s system used in the design and manufacture of equipment specified in t procurement documents. In order to clarify the FSAR, Subsection 3.2.2 of RESAR-SP/90 PDA Modules 4, 5, 7 and 6/8 is being' revised to indicate that the QA Program described in Chapter 17.1 is applied to all SC-1, SC-2, SC-3 structures, systems and components. l f

O lO O HAPWR-QA 260-15 AMENDMENT 3 AUGUST 1989 0011D:10

3b i TABLE 17.1 Regulatory Guidance Applicable to Quality Assurance Program I i

1. Regulatory Guide 1.26, " Quality Group Classifications, and Standards for j Hater, Steam, and Radioactive Waste Containing Components of Nuclear j Power Plants" (2/76) i

2. Regulatory Guide 1.28, " Quality Assurance Program Requirements (Design and Construction)" (8/85) .

1

3. Regulatory Guide 1.29 " Seismic Design Classification" (9/78) j 4 Regulatory Guide 1.30, " Quality Assurance Requirements for the  !

Installation. Inspection, and Testing of Instrumentation and Electrical Equipment" (8/11/72) 5 Regulatory Guide 1.37, " Quality Assurance Requirements for Cleaning of Fluid Systems and Associated Components of Hater-Cooled Nuclear Power Plants" (3/16/73).

6. Regulatory Guide 1.38 - Revision 2 " Quality Assurani.e Requirements for Packaging, Shipping, Receiving, Storage, and Handling of Items for Hater-Cooled Power Plants" (5/77)

7. Regulatory Guide 1.39 - Revision 2. " Housekeeping Requirements for Water-Cooled Nuclear Power Plants" (9/77)

8. Regulatory Guide 1.58 - Revision 1 " Qualification of Nuclear Power Plant Inspection, Examination, and Testing Personnel" (9/80).

9. Regulatory Guide 1.64 - Revision 2 " Quality Assurance Requirements for the Design of Nuclear Power Plants" (6/76)

10. Regulatory Guide 1.74., " Quality Assurance Terms and Definitions" (2/74)

11. Regulatory Guide 1.88 - Revision 2. " Collection, Storage, and Maintenance of Nuclear Power Plant Quality Assurance Records" (10/76).

12. Regulatory Guide 1.123 - Revision 1. " Quality Assurance Requirements for Control of Procurement of Items and Services for Nuclear Power Plants" (7/77).

13. Regulatory Guide 1.144, Revision 1, " Auditing of Quality Assurance Programs for Nuclear Power Plants" (8/80).

14. Regulatory Guide 1.146, " Qualification of Quality Assurance Program Audit Personnel for Nuclear Power Plants" (8/80).

G HAPHR-QA 260-16 AMENDMENT 3 00110:1D AUGUST 1989

- - _ _ _ _ _ - _ - _ _ _ _ -