ML20212B315

Safety Evaluation Re B&Wog EOP Technical Bases Document (Tbd).All Outstanding Issues Closed Re Generic Review.One Unresolved Issue May Be Addressed on plant-specific Basis. Staff Has No Plans for Further Generic Review of TBD

ML20212B315
Person / Time
Issue date:	09/14/1999
From:	NRC (Affiliation Not Assigned)
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ML20212B295	List: ML20212B292 ML20212B315
References
TAC-M54946, NUDOCS 9909200051
Download: ML20212B315 (34)

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION OF THE BABCOCK & WILCOX OWNERS GROUP i

EMERGENCY OPERATING PROCEDURES TECHNICAL BASES DOCUMENT -

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PREVIOUS TAC NO. M54946 ,

i SEPTEMBER 1999 i

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ATTACHMENT 1 9909200051 990914 PDR TOPRP EPfVBW C PM ,

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TABLE OF CONTENTS Table of Acronyms . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 1 INTRODUCTION AND

SUMMARY

. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 2 BACKGROUND AND IDENTIFICATION OF ISSUES . . . . . . . . . . . . . . . . . . . . . . . . . 6 2.1 Introduction . . . . . . . . . . . . . . . . . ..................................... 6 2.2 Redefinkion of Allissues, Definition of New issues, and issue Disposition. . . . . 7 3 TECHNICAL BASIS DOCUMENT REVIEW . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 3.1 Document Terminology, Relationships, and Staff Findings . . . . . . . . . . . . . . . . . . 8 3.2 Long Term Procedures Development Objectives and Staff Findings . . . . . . . . . 10

3. 3 . New Findings . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 3.3.1 Provision of a Generic Emergency Procedures Guideline . . . . . . . . . . 10 3.3.2 Subcooling Margin Determination and Response . . . . . . . . . . . . . . . . . 11 3.3.3 Inadequate Core Cooling . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 3.3.4 Criticality and Recriticality . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12

3. 3. 5 Priorities . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . 12 3.4 Abnormal Transient Operating Guidelines issues - Discussion and Findings. . . 12 3.4.1 Entry into Emergency Operating Procedures . . . . . . . . . . . . . . . . . . 13 3.4.2 RCS inventory Measurement . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 3.4.3 Reactor Coolant Pump Operation . . . . . . . . . . . . . . . . . . . . .. . . . . . . 17 3.4.4 Loss of AC Power . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18 3.4.5 Containment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18 3.4.6 Anticipated Transients Without Scram . . . . . . . . . . . . . . . . . . . . . . . 19 3.4.7 Pressurized Thermal Shock . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19 3.4.8 Cooldown and Interfacing With Other Procedures. . . . . . . . . . . . . . . . 19 3.4.9 High-Pressure injection Cooling . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20 3.4.10 Steam Generator Operation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20 3.4.10.1 Steam Generator Control .. . . . . . . . . . . . . . . . . . . . . . . . . . 20 3.4.10.2 Steam Generator Level Indication . . . . . . . . . . . . . . . . . . . 23 3.4.10.3 Steam Generator Trickle Feed . . . . . . . . . . . . . . . . . . . . . . . 23 3.4.10.4 Steam Generator Tube Rupture . . . . . . . . . . .. . . . . . . . . . 24 3.4.11 Degraded Core . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 28 3.4.12 Reactor Coolant System inventory and Pressure . . . . . . . . . . . . . . 29 3.4.13 Analyses . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 29 3.4.14 Miscellaneous . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . 30

- 3.4.15 Non-Hot /Non-Pressurized System Conditions . . . . . . . . . . . . . . . . . . 31 3.4.16 Containment . . . . . . . . . . . . . . . . . ., ...... . . . . . . . . . . . . . . . 31 3.4.17 Loss-of-Coolant Accident Outside Containment . . .. ... ...... 31 3.4.18 Severe Accidents . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 31 3.4.19 Loss of AC Power . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32 4 CONCLUSIONS . . . . . . . . . . . . .... . .... .. .. . . .. . .32 5 R EF ER E NC E S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32 I

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I TABLE OF ACRONYMS l AC alternating current l

ADV atmospheric dump valve j

AOP abnormal operating procedure (

ATOG abnormal transient operating guidelines l ATWS Anticipated Transient Without Scram l B&W Babcock and Wilcox B&WOG Babcock and Wilcox owners group

CDF core damage frequency .

l CR control room DHR decay heat removal EOP emergency operating procedure EPG emergency procedures guideline GL generic letter {

i HPl high-p?rcure injection l ICC inadequ4:e core cooling LBLOCA large-br' ak loss-of-coolant accident LOCA loss-of-coolant ace; dent LOOP loss of offsite power l NC natural circulation l NSSS nuclear steam supply system NRC Nuclear Regulatory Commission i

PORV pressure- or power-operated relief valve PTS pressurized thermal shock RC reactor coolant RCP reactor coolant pump RCS reactor coolant system l RT reactor trip RV reactor vessel SBLOCA small-break loss-of-coolant accident SCM subcooling margin SER safety evaluation report SG steam generator SGTR steam generator tube rupture TBD technical basis document

! TMI Three Mile Island TRACC tube rupture alternate control criteria 3

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1 INTRODUCTION AND

SUMMARY

A version of the Oconee Unit 3 Abnormal Transient Operating Guidelines (ATOG) (Reference 1) was reviewed by the staff in 1983 in lieu of reviewing a generic emergency procedures guideline (EPG) (References 2 and 3). That review identified open issues that the Babcock and Wilcox (B&W) owners group (B&WOG) addressed via a series of technical basis documents (TBDs),

each a revision of the previous version, and each containing improvements that a majority of the licensees judged should be included in the generic reference document. TBD Revision 6 (References 4 and 5) was audited by the staff. As reported in this safety evaluation report (SER), it was found to be sufficiently developed that all ATOG open issues were closed for those licensees who use the TBD in development of their plant-specific EPGs and emergency operating procedures (EOPs) consistent with this SER's findings.

The B&WOG has continued to develop the TBD since the staff audited Revision 6. The current TBD, Revision 8 was issued November 1,1996 (Reference 6). Revision 9 is planned for early 2000 (Reference 7).

The TBD is composed of three volumes that encompass the accident mitigation strategy to be embodied within a B&W facility's EPGs and its EOPs. Volume 1 of the TBD is a vendor-recommended, generic EPG. Volume 2 is the bases for Volume 1 and Volume 3 contains additional bases material and altemate strategies. As discussed in the June 17,1999, meeting with B&WOG representatives (Reference 7), the staff understands that Davis Besse, Three Mile island (TMI), and Crystal River use Volume 1 as the reference EPG for plant-specific EPGs and EOPs, and that plant-specific documentation that describes deviations between the plant-specific EPGs and the TBD is based upon Volume 1. This is consistent with the NRC's position during TBD review and is acceptable for preparation of plant-specific EPGs and of EOPs.

The staff also understands that Arkansas Nuclear One (ANO) Unit 1 and Oconee will convert to using TBD Volume 1 as the reference EPG following publication of TBD Revision 9. In the interim, the staff will accept use of all three volumes of the TBD at ANO Unit 1, and at Oconee as the reference for establishing plant-specific deviations.2 Following its audit of TBD Revision 6, as reported herein, the staff concluded:

1. Volume 1 of Revision 6 is acceptable as a vendor-recommended, generic EPG and is acceptable as a reference EPG for purposes of addressing deviations in plant-specific documentation. Volume 2 of Revision 6 is an acceptable bases for Volume 1. Volume 3 of Revision 6 is acceptable as a source of additional bases material and alternate strategies that may be used, where applicable, to develop the plant-specific EPGs and EOPs.

'The meaning of generic EPGs, plant-specific EPGs, EOPs, and other procedures is discussed in Section 3.1, below.

Ucensees may close TMI Action item I.C.1 by committing to comply with the TBD consistent with the findings in this SER. Any future inspections will then be plant-specific and will use the TBD and this SER as references.

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2. Revision 6 is sufficiently developed that no further generic review is necessary.8 Further revisions may be substituted for the approved Revision 6 without staff review or  ;

approval.

3. The TBD maintenance program should continue. Typical topics for the continuing program include improvements identified in this SER and improvements found to be needed by the B&WOG.

4. All ATOG open issues are closed for those licensees who use the TBD in the approved I manner to prepare plant-specific EPGs and EOPs.

5. The TBD issue of use of levelinstrumentation is closed on a generic basis. It may be pursued on a plant-specific basis.

6. At the end of the Revision 6 audit, aspects of the following TBD issues remained to be addressed (Reference 8):

a. Subcooling margin (SCM). Improve guidance for determination of SCM.

b. Criticality and recriticality. Provide guidance for inadvertent criticality that occurs ,

after entry into the EPG guidance or during normal cooldown before reaching decay heat removal (DHR) operation.

c. Priorities. Provide clear prioritization guidance.

d. Reactor coolant system (RCS) inventory measurement. Address RCS inventory measurement.

e. Steam generator (SG) trickle feed. Provide guidance for cooling by SG trickle feed.

f. SG tube rupture (SGTR). Address the SGTR issues identified in this SER.

g. Loss-of-coolant accidents (LOCAs) outside containment. Address LOCAs outside containment and provide suitable guidance.

The B&WOG indicated it would evaluate these items in its continuing TBD improvement program (Reference 9). Evaluation completion has been reported as follows:

a. SCM. Reference 10, October 2,1995

b. Criticality and Recriticality. Reference 11, August 28,1996

c. Priorities. Reference 10, October 2,1995

d. RCS Inventory Measurement. Reference 12, September 11,1997 Dhe staff indicated its intention to close the generic review in Reference 8.

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e. SG trickle feed. Reference 9, June 13,1995

f. SGTR. Reference 11, August 28,1996

g. LOCAs outside containment. Reference 12, September 11,1997 Therefore, these issues are resolved based upon the B&WOG correspondence listed above.

Although the staff plans no further TBD review, it would hke to maintain eight copies of the up-to-

- date TBD for reference purposes and would appreciate receiving eight copies of each revision as it becomes available. The staff requests that one copy be sent to the U. S. Nuclear Regulatory Commission, Attention: Document Control Desk, Washington, D. C. 20555. It also requests that seven hard copies be sent to the Chief, Reactor Systems Branch, Office of Nuclear Reactor Regulation, United States Nuclear Regulatory Commission, Washington, D.C.

20555. The Reactor Systems Branch's responsible person will retain one copy, will sent one to the NRC Technical Train;ng Center, will send one to the NRC Operations Center, and will send one to ead of the NRC regional administrators.

2 BACKGROUND AND IDENTIFICATION OF ISSUES 2.1 Introduction Early responses to the March 1979 accident at TMl are contained in the Nuclear Regulatory Commission's (NRC's)"TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations," NUREG-0578 (Reference 13). (See also Reference 14.) Many of the NUREG-0578 recommendations were incorporated into the NRC action plan, NUREG-0660 (Reference 15) and its clarifications, NUREG-0737 (Reference 16). Item I.C.1 of the action plan required development of EPGs.

As identified above, a version of the Oconee Unit 3 ATOG (Reference 1) was reviewed by the i staff in lieu of reviewing a generic EPG. This review is described in a September 19,1983, SER '

(Reference 2) and a December 14,1983 supplement to the SER (Reference 3).' in the-September 19 SER there were four conditions for acceptance of ATOG as the technical bases for the EPGs and EOPs. Two of the four conditions dealt with the "short-term" resolution of anticipated transients without scram (ATWS) and interrupted natural circulation. Following a review of a July 2,1983, B&WOG transmittal revising ATOG (Reference 17), the "short-term" aspects of these two issues were resolved as documented in the SER supplement of December 14,1983, and ATOG was implemented by those licensees having B&W-supplied nuclear steam supply systems (NSSSs). j The other two ATOG acceptance conditions dealt with the establishment of a comprehensive plan to deal with other technical issues discussed in the original SER and the agreement to

- implement the comprehensive plan. The B&WOG submitted such a plan on December 9,1983 ,

(Reference 18). The comprehensive plan led to a series of improvements that were l

'This was a generic review in that the review and accompanying SERs are not site-specific and apply to all l B&W facilities. A site-specific review would address numerous items not covered by the generic review.

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documented in a series of TBDs, each of which was an update of the previous issue. Revision 6 (References 4 and 5) was submitted for review in 1991 and is the subject of this SER.

Revision 7 (References 19 and 20) was issued for the staffs information in 1993 and, although not reviewed by the staff, is referenced during this review. Revision 8 was similarly provided to the staff in 1996 (Reference 6). Revision 9 is expected in early 2000 (Reference 7).

r 2.2 Redefinition of All lasues. Definibon of New issues. and issue Disna=4 ion The TBD was developed as the B&WOG addressed the Oconee ATOG open issues remaining from the September 19,1983 SER (Reference 2) in which the staff understood that all licensees with B&W NSSSs committed to use the TBD. The TBD consequently became a replacement for ATOG and, as a re. d, the issues became obsolete. Therefore, the B&WOG and the staff redefined the issues and consolidated them into categories that became the basis for addressing the ATOG open issues (Reference 21).

The redefined issue categories are:-

1. Entry into EPGs and EOPs

2. RCS inventory measurement

3. Reactor coolant pump (RCP) operation

4. Loss of AC power

5. Containment

6. ATWS

7. Pressurized thermal shock (PTS)

8. Cooldown and interfacing with other procedures

9. High pressure injection (HPI) cooling

10. SG control

11. Degraded core j

12. RCS inventory / pressure

13. Analyses

14. Miscellaneous These are addressed in Sections 3.4.1 through 3.4.14 of this SER.

Reference 14 also defined the following future needs:

15. Non-hot /non-pressurized NSSS conditions

16. Containment

17. LOCA outside containment

18. Severe accidents

19. Loss of AC power The apparent duplication of issues 15 through 19 with issues 1 through 14 is due to content '

differences.

The staff stated that closure of issues 15 through 19 was beyond the effort necessary to close the open issues identified in the 1983 SER. Consequently, issues 15 through 19 became 7

l - additional issues that were associated with the TBD review, not ATOG issues remaining from the 1983 SER. Issues 15 through 19 are addressed in Sections 3.4.15 through 3.4.19.

l l The staff aud.'ied Revision 6 of the TBD and identified the following additional issues:

i 1. Provision of a generic EPG

2. SCM determination and response

3. Inadequate core cooling (ICC)

'4. Recriticality

5. Priorities These are discussed in Sections 3.3.1 through 3.3.5.

r 3 TECHNICAL BASIS DOCUMENT REVIEW 3.1 Document Terminoloav. Relationshios. and Staff Findinos The following procedures are discussed in this SER:

• Emergency Operating Procedures (EOPs) - The control room (CR) procedures that operators use to respond to an emergency condition involving the NSSS, the containment, or systems required for support of the NSSS or containment. In general, EOPs are entered in case of a reactor trip (RT), if conditions occur that should result in RT, or as an operator option for addressing a plant condition where the operator believes the guidance is appropriate. The B&WOG guidance is for EOPs to also be entered for SGTR and if critical safety functions are jeopardized during shutdown operation when conditions have not been achieved for DHR system operation.

o Abnormal Operating Procedures (AOPs)- The procedures that operators use to respond to off-normal or abnormal conditions where entry into the EOPs is not needed or required. AOPs may be used in conjunction with EOPs when appropriate.

e Normal Operating Procedures - The procedures that operators use during routine operation. These procedures may be used in conjunction with EOPs and AOPs when appropriate.

• EOP Guidelines or Emergency Procedures Guidelines (EPGs)- The guidance and identification of procedural steps that provide the foundation for EOPs. There are two types:

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Generic-The EPGs applicable to a class of plants.5 Generic EPGs are typically less detailed than plant-specific EPGs, may not conts'.n steps that apply to a particular unit, and will not contain plant-specific information such as valve numbers that differ from unit to unit.

Plant-specific - The EPGs applicable to a specific NSSS and associated equipment at a licensee's facility.

When appropriate, part of the EOP coverage may be provided in non-EOP procedures provided:

e Interfacing information is contained in the EPGs and EOPs, and e There is a rational basis for locating guidance in non-EOP procedures.

Guidance regarding such coverage is provided by the plant-specific EPGs. The quality of such non-EOP procedures coverage must be comparable to that provided for EOPs, although this I requirement may not apply to the supporting documentation for non-EOP procedures.  !

In TBD Revision 7 Volume 3, the B&WOG extended the guidance to mitigation of loss of the DHR system. There was no requirement to develop this guidance as part of the EPG/EOP program, nor was it required to be incorporated into the TBD. The staff considers such additions and extension of coverage as a logical improvement, but it raises a question regarding the significant body of regulatory requirements that apply to the EOP development process that begins with the generic EPG and includes formal guidance, documentation, validation, and verification - material the staff typically references during inspections. A question raised by the l B&WOG's incorporation of DHR system loss into the TBD is whether comparable plant-specific EPG and EOP sections must be written and, if so, whether they are subject to the same development and inspection process as applies to use of the TBD during development of EPGs  !

and EOPs.

The staff recognizes that the B&WOG could have published such guidance separately from the TBD, in which case there would be no question regarding staff involvement via the EPG and EOP review and inspection process. Consequently, the staff will not require application of the formal EPG to EOP process on the sole basis that an owners group has chosen to include such coverage in a portion of its generic EPGs. Further, procedures developed from such EPG coverage do not have to be addressed as though they were EOPs developed in response to staff requirements unless the staff develops future requirements that change these conclusions.

However, this SER does not supercede any site-specific requirements applicable to procedure development, review or approval imposed via the quality assurance program, or any other regulatory statue.

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8 Generic EPGs and bases documentation were written to utilize the knowledge of an affected parties to develop a document that descnbes the concept and major steps to be contamed in EOPs in a class of nuclear power plants -in this case the B&W designed NSSS and associated equapment. This significantly reduced development effort since licensee's could base their plant-specific programs on documentation that was developed, reviewed, and had NRC approval.

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l 3.2 1.ona-Term Procedures Development Obiectives and Staff Findinas The initial SER (Reference 2) stated that the long-term objective in developing EOPs was to provide operator guidance that covered any NSSS condition that jeopardized or potentially jeopardized a barrier to release of radioactive material.' Therefore, EOPs or their equivalent l

would be entered at any time such a condition may exist, and coverage using such procedures would continue until (1) alternate procedures or guidance processes apply or (2) a controlled, stable condition is achieved that operators can maintain until support personnel can logically and carefully plan future strategy. With the exception of such situations as an ongoing severe

accident, if one of these exit conditions was satisfied, and a condition recurred or a new one occurred that potentially jeopardized tha fuel cladding, the RCS pressure boundary, or the l l containment, then the EOPs or their equivalent would be reentered to cope with the condition. )

3.3 New Findinas This section addresses findings that resulted from the audit of TBD Revision 6. These include l the following:

1. Provision of a generic EPG ,

2. SCM determination and response '

3. ICC

4. Recriticality

5. Priorities 3.3.1 Provision of a Generic Emeroency Procedures Guideline The staff finds the following use of the TBD by the licensees to be acceptable:

• Compare Volume 1 to plant-specific EPGs, o Justify inconsistencies between Volume 1 and plant-specific EPGs, and e Document this process in intemal documentation.

l This finding is subject to the following qualifications and clarifications that are intended to be consistent with the discussions described in Reference 22:

e if, functionally, a plant-specific EPG and EOP are accomplishing the same thing as specified in Volume 1, or differences are identified and justified, then the plant-specific l EPG and EOP are acceptable with respect to the difference.

• . An exact correspondence in the step numbers between Volume 1 and plant-specific EPGs is not necessary, nor is it necessary to document such differences when they l

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• EOPs and EPGs were originally developed to address potential probierns during power operation. The need to further address shutdown operation arose later, and many licensees have written AOPs to provide shutdown coverage.

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occur due to such actions as inserting plant-specific steps provided the modification of the step (s) does not invalidate the mitigation strategy.

• Use of Volume 3 bases material in plant-specific EPGs that is referenced in Volumes 1 or 2 as additionalinformation or clarification does not constitute a deviation, although the source should be identified for traceability.

• Each difference between plant-specific EPGs and Volume 1 that occurs because of a substitution of Volume 3 options in place of Volume 1 options is identified in plant-specific deviation documentation and the reason for the substitution is provided. It is not necessary to justify such substitutions if the Volume 1 and/or Volume 3 mitigation strategy are maintained because Volume 3 provides alternate acceptable strategies and supplementalinformation. Note, however, that a deviation will occur and must be appropriately addressed if the order of steps provided in Volume 1 is changed or the resulting mitigation strategy differs from that provided in either Volume 1 or Volume 3.

This discussion also applies to steps and information in the plant-specific EPG bases when they are compared to Volume 2.

The staff understands that ANO and Oconee do not compare plant-specific EPGs to TBD Volume 1, but instead compare to the TBD. This is acceptable until approximately July 2001, after which plant-specific EPGs must be compared as described in Section 3.3.1, above, to credit this SER as part of the licensing basis.7 3.3.2 Subcooma Marain Determination and Resoonse Reliable determinabon of SCM is necessary for correct functioning of the TBD guidance and l EPGs based upon that guidance. This was discussed with B&WOG representatives during the j February 2,1993 meeting (Reference 22). The B&WOG informed the staff that the evaluation had been completed in its letter of October 2,1995 (Reference 10). As previously mentioned,  ;

the staff determined that further generic review was not necessary. Consequently, it does not plan a generic review of the B&WOG evaluation.

3.3.3 Inadeouste Core Coolina i

The B&WOG submitted its approach to severe accident management to the staff in June 1993.

On this basis, it anticipated a transfer to severe eccident management would be made if clad i temperature exceeded 1800 'F (1256 'K). The TBD Revision 6 guidance for certain ICC )

mitigation actions that were to be performed if clad temperature reached 1800 *F (1256 'K) '

would instead be initiated at 1600 *F (1144 'K). As stated in TBD Volume 3, page I.B-1: 'The TBD guidance will be revised as necessary to maintain the proper interface with guidance 1 provided elsewhere for severe accident management." The change from 1800 'F (1256*K) 7 The July 2001 date is based on an estimated TBD Revision 9 publication date in the Spring of 2000 followed by sufficient time to reconfigure the plant specific EPGs and EOPs.

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satisfies staff concems regarding adequacy of the guidance with respect to the conditions for initiating that guidance. Therefore, this issue is closed.

The staff commented during the ATOG review that cooldown considerations also applied to a damaged core. The staff considers this topic to W appropriately addressed as part of severe accident management. Other aspects are addressed in Section 3.4.11 of this document. All

ATOG open issues on this topic are clona:

3.3.4 Criticality and Recriticality There is a potential for criticality during cooidown under normal conditions or following entry into the EOPs where criticality was not a concem prior to entry or during the initial actions. There is also a possibility of recriticality during cooldown. These possibilities are not addressed ir TBD -

Revision 6. The B&WOG and staff agreed that inadvertent criticality or inadvertent approach to enhcality should be addressed Ls part of the TBD. As a result, the B&WOG scheduled completion of an evaluation for December 15,1996 (Reference 9 and 10), and the B&WOG informed the staff that evaluation was complete on August 28,1996 (Reference 11). The staff does not plan to further review this B&WOG generic work.

3.3.5 Priorities A key to usage of TBD-based EOPs is operator response to priorities or safety functions. This could be better emphasized in the TBD where, for example, it is difficult to find whether symptoms have a tugher priority than the tube rupture attemate control criteria (TRACC) guidance for steam generator tube rupture (SGTR). Inadequate core cooling (iCC) is covered via Step 7.0 of the loss of SCM guidance as opposed to assigning ICC a priority and treating it as a symptom of loss of a safety function. In the case of large-break loss of coolant accidents (LBLOCA), a transfer occurs prior to SCM Step 7.0 to transfer to the LBLOCA cooldown section where one entry criterion is that the RCS is saturated - a potential conflict of instructions.

Criticality (as identified in Section 3.3.4, above) does not appear to be addressed except upon initial entry into the TBD, yet inadvertent criticality should have a high priority.

Similarly, response to a change in conditions may not be clear. For example, operators may be in a procedure path intended to provide cooling under a no-high-pressure injection (HPI) condition, and an HPl system may become available. The recommended operator guidance is not clear.

The B&WOG scheduled evaluation of this issue to be completed by December 15,1995 (Reference 9) and informed the staff of its completion in a letter dated October 2,1995 (Reference 10). Since the staff determined that further generic review was not necessary, the staff does not plan to assess the B&WOG evaluation.

3.4 Abnormal Transient Operatina Guidelines issues - Discussion and Finding The following categories were identified in Section 2 of this SER as topics remaining to be closed as a result of previous staff reviews (Reference 21):

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w; 1.' Entry into EPGs (EOPs)

2. RCS inventory measurement

3. RCP operation

4. Loss of AC power

5. Containment

6. ' A1WS

7. . PTS l 8. Cooldown and interfacing with other procedures I

9. HPl cooling

10. SG control

11. Degraded core

12. RCS inventory / pressure -

13. Analyses

14. Miscellaneous.

15. Non-hot /non-pressurized NSSS conditions

16. Containment i

17. LOCA outside containment l 18. Severe accidents j

19. Loss of AC power  !

l These are addressed in Sections 3.4.1 through 3.4.19, below. (References to EPG step  ;

l numbers are to those steps in TBD Revision 6, Volumes 1 and 2).

l l 3.4.1 Entrv Into Emeroency Operatina Procedures Open issue. Entry should cover the following items: l

1. RT or the existence of conditions requinng RT l 2. Forced shutdown for SGTR )

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3. Via interfacing with non-EOP procedures l ,

! . Staff Findina. Items 1 and 2 are clearly identified as entry criteria in the TBD, thus satis *ying the ;

staff requirement. Item 3, a concem dealing with emergency conditions when using non-EOP procedures to mitigate an abnormality, is addressed by the TBD entry criterion at Step 1.0 which states: "jf symptom occurs while reactor shutdown above DHR operation,.TtJEbl perform the appropriate section of this guideline." This removes the need for entry interfacing for transferring from non-EOP procedures provided it is appropriately addressed by such means as  ;

operator training. Transfer from the EOP to other procedures is addressed by general guidance  !

such as instruction to perform a normal cooldown or occasionally to specific plant procedures.

In either case, trained operators will recognize the appropriate procedures that are to be used.

The staff concerns have been addressed and this issue is closed.

Other aspects of coverage are addressed in other sections of this SER, such as Sections 3.3.4 (criticality and recriticality) 3.4.4 and 3.4.19 (loss of AC power),3.4.5 and 3.4.17 (containment),

3.4.6 (ATWS),3.4.15 (non-hot non-pressurized NSSS conditions), 3.4.17 (LOCA outside containment), and 3.4.18 (severe accidents).

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3.4.2 RCS Inventory Measurement The staff had previously concluded that EPGs should include reactor vessel (RV) level indications. This was not accomplished in ATOG, and thus was an open issue.

The staff has determined that level indications are discussed, but not generally recommended, in TBD Volume 3. They are not addressed in Volume 1. Void monitoring via the RCPs is not mentioned in the TBD. This is consistent with a B&WOG position that level indication is not needed and, if used, that it can be detrimental to mitigation of emergency conditions.

Comments regarding level indication were provided by B&WOG representatives at the February 2,1993 meeting (Reference 22). The staffs understanding of those comments, their intent, and the staff appraisal are summarized as follows:

• .B&WOG Levelinstrumentation is not necessary to apply EOPs that are developed from I the TBD guidance. i l

StiLffaoorassal. TBD Volume 1 does not include levelindication and, consequently, EOPs developed from only that guidance will not reference level instrumentation.

However, the B&WOG comment does not address the issue. The issue is whether level indication should be addressed in EPGs and EOPs.

• B&WOG. Operators do not need levelinformation to mitigate events.

1 Staff aooraisal:

• Level and void instrumentation is installed and provides backup data for indications referenced in the TBD For example, increasing void indication with RCPs running may indicate a malfunction of SCM indications if a significant SCM exists. Such verification and diverse indication capability are basic to both the EOPs and to the safety approach of using multiple, independent equipment for nuclear power plant operation.

• Level and void instrumentation provides useful insights into RCS behavior. For example:

- Knowledge of void location and of void and level behavior can help operators and support personnel in assessing plant response to operator actions, and can provide insights and verifications that potentially instill operator confidence under stressful conditions. This may be valuable if an approach to ICC conditions is occurring since it may allow an appraisal of time to initiation of ICC.

- Level information will be useful if there is no SCM and DHR operation is desired.

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• B&WOG. There is no action an operator would take in response to level indication that l is not already provided by the vendor-recommended EPG.

l Staff aooraisal. The staff found areas during its review where SCM-associated errors could occur that resulted in the improvements identified in Section 3.3.2, as well as in errors that have been corrected and are not addressed in this SER (such as inappropriate use of cold leg temperature and potential use of hot leg temperature when these would provide misleading information). Levelindication, and, if RCPs are running,

void indication based upon RCP characteristics, would signal that a loss of subcooling

. had occurred before superheat indications were received, thus providing an opportunity l

for more appropriate operator response. This is one example of an indication providing backup for other instrument-provided information that would be useful if there is a problem due to equipment failure, unanticipated guidance problems, or operator error.

Numerous potential event paths exist that lead to voiding. Early detection could be useful for such actions as venting to avoid a loss of natural circu!ation. Knowledge of void locations may also allow more effective steps to eliminate those voids. For example, opening a hot leg vent will not necessarily help mitigate an RV head void if head void elimination is desired.

I Knowledge that a void extended into or below the midloop level range would be useful if  ;

l DHR operation was desired.  !

• B&WOG Level instrumentation can be incorrect and can, therefore, lead to operator I error.

Staff aooraisal. The staff recognizes that some of the installed levelinstrumentation will provide erroneous indications under some conditions, such as when the level system is connected to RCS vent piping. Such connections can cause level indication to be '

meaningless when the vents are open. Lonti tubing runs are also of potential concem, particularly when containment craditions chinge during an accident. The staff additionally recognizes that thire are variations in the level range addressed by the installed level instrumentation. However, instrumentation that met specified criteria was required to be installed in all PWRs to satisfy TMl Action P!an Requirement ll.F.2 of NUREG-0737 (Reference 16), which also addressed incorporation of information into operator guidance. (NUREG-0737 also specified criteria to be applied to other instrumentation, such as SCM and temperature indications.) If level and void indication instrumentation errors occur over a known and sufficiently narrow range of applicability, then it may meet NUREG-0737 requirements and is appropriate for use in EOPs with suitable qualifications. Conversely, if the instrument installation is likely to provide incorrect information that can lead to operator error, it may not be responsive to the NUREG-0737 requirement.

• B&WOG. Requiring levelinstrumentation can unnecessarily add to operator burden l since it is not necessary for operator actens.

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Staff aooraisal. Use of independent and diverse sources of information is always necessary whether or not it is specifically required in the EOPs. Operators are correctly taught not to rely on one indication or source of information, but to always make use of attemale confirmatory data. General instructions that require operators to ignore level instrumentation, as expressed by some operators during inspections, contradict the requirement to use available information and, in contrast to the B&WOG position, may actually add to operator burden.

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l e B&WOG. The necessity for level indication reflected the need immediately following the l TMI accident. New post accident instrumentation, procedures, and training i improvements have changed that need.

Staff aooraisal. Levelindication was a part of the solution resulting from the TMI j accedent. The cited improvements are real and important, but the staff does not agree that they eliminate the need for level indication.

In light of the disagreement between the staff and B&WOG, the staff examined the need for and usefulness of RV level and void indication in some depth. Several aspects are discussed below.

Inadequate RV level sufficient to initiate an approach to core damage will result in core exit steam becoming superheated, with roughly 1000 *F (556 'K) of superheat necessary for initiation of damage. Superheat will usually be indicated by multiple in-core thermocouples that meet safety-related instrumentation criteria (although an exception is mentioned in the next paragraph). Procedures, training, and operating practice are likely to result in operators responding to a superheat condition, regardless of the reason for entering that condition.

Accurate RV level indication would cause no change in operator actions when in the superheat temperature range. The potential benefit would accrue if operator or instrument error was the cause of reaching that condition, and knowledge of level might result in avoiding the superheat condition.

RV level at or below the hot-leg elevation will result in a loss of SCM before core cooling is seriously jeopardized. SCM is a key to operator response, and, next to criticality indications and superheat, has the highest priority for operator attention. The staff believes operators are likely to discover a loss of SCM at critical locations within the RCS consistent with the discussion of Section 3.3.2, above, as long as the SCM indication works properly. Failure of SCM indication is likely to cause operators to calculate SCM from pressure and temperature data, with an increased possibility of error. Accurate RV level indication and void indication provides backup data for that condition. A similar statement may be made with respect to superheat indication and translation of voltage indications to temperature.

Historically, inadvertent injection of gas into the RCS has occurred. Level indication will provide information pertinent to location of the gas and insights into its influence, insights that SCM indication will not provide since SCM indication only addresses the potential for steam void formation, not gas void.

Level instrumentation may provide insight into plant status and mitigation effectiveness, including the following:

16

e Indication of void location as an input into decisions regarding void control and removal actions j 1

i' e Anticipation of the impact of void control actions, including such items as preparatory actions, where the impact will be a function of void size e Anticipation of RCS behavior, such as loss of and regaining of SG cooling during level loss following some small break LOCAs e Independent confirmation of other instrument readings and operator conclusions e input into decisions to initiate DHR operation and into guidance for not starting a DHR pump under inadequate level conditions The staff finds level indication to be of sufficient value that the data should be appropriately used  !

in the EOPs. This is most beneficially accomplished by providing guidance in the TBD, j Although the staff does not require incorporation of level guidance into the TBD, it does require 1 usage in the EOPs consistent with NUREG-0660, item II.F (Reference 15) and NUREG-0737 l ltem II.F.2 (Reference 16). This issue may be pursued on a licensee-specific basis.

The B&WOG scheduled evaluation of RCS inventory measurement to be completed by December 15,1997 (References 9 and 10). The staff was informed of completion on September 11,1997 (Reference 12).

3.4.3 Reactor Coolant Pumo Ooeration l

Ooen Issues. RCP operation should be evaluated for voided RCS conditions and appropriate EPG guidance should be provided. RCP operation for saturated RCS conditions should be  !

clarified. The effectiveness of RCP " bumps" should be demonstrated. The issues encompass  !

both initial accident mitigation and RCS cooldown. In addition, the B&WOG should consider guidelines for operator action to preserve RCP cooling capability in light of the potential for inadvertent isolation or to reinstate cooling following automatic or operator-initiated action.

Staff Findina. The B&WOG has studied voided conditions and NSSS response to mitigation actions. RCP operation is addressed for these conditions at appropriate locations throughout Volume 1, such as Steps lil.B.18, Ill.C.10, Ill.C.16, Ill.C.18, Ill.E.11 and Ill.F.9, Recovery Subsection SS-1, " Reactor Coolant Pump Restart," and Specific Rule i for loss of SCM which instructs the operator to immediately stop the RCPs. An extensive discussion is provided in TBD Volume 3. Staff concerns regarding RCP operation for voided RCS conditions and provision of appropriate guidance have been satisfactorily addressed. The issue is closed.

RCP seat cooling is addressed in a number of TBD sections, such as Verification Step 10, the bases for Step 10, in Volume 3, Sechon IV.H, and in references to plant-specific procedures.

The guidance is satisfactory and the issue is, therefore, closed.

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3.4.4 -Loss of AC Power j Ooen lasun EPGs should cover generic aspects of loss of all AC (station blackout) and recovery, EPGs should provide coping strategy for NSSS control during the period of loss and recovery, and EPGs should address loss of offsite power (LOOP).

Staff Findmg Immediate actions guidance is provided in Volume 1 and reference is made to plant-specific procedures. Additional information is provided in other volumes, such as Volume 3 Chapter N.H. The immediate actions guidance is appropriate. Transfer to appropriate plant-specific procedures for conditions such as LOOP and station blackout is acceptable provided such procedures are consistent and properly interface with the TBD guidance. Review of plant-specific procedures is outside the scope of this SER; however, the staff notes that procedures should provide adequate coverage until such time as accident management organizations are fully functional and can provide appropriate guidance and procedures. These items are adequately addressed and the issues, therefore, are closed.

.3.4.5 Containment Open Issues. The following are the containment issues:

1. EPGs covering containment should be provided. Inclusion of general guidance is acceptable for plant specific areas provided it is sufficient to ensure plant-specific coverage.

2. Coverage should be generally consistent with the areas covered in ATOG. In addition, the following areas should be addressed:

a. Radiationlevelforisolation

b. Operation of contaiement vents

c. Hydrogen concentration

d. Use of temperature indications Sjaff Findinos. TBD Revision 7 adds containment guidance in a number of areas, such as Steps Ill.B.4, Ill.C.3.5, Ill.D.5.5, and lli.E.10.2. It also modifies the instructions contained in Ill.F.5. This provides coverage that is generally consistent with the areas covered in the original ATOG.

The TBD discussion of containment hydrogen control appears oriented to the facility design basis condition, rather than the broader concem of more rapid hydrogen generation that might occur. For example, repressurization and venting of the containment to contro! hydrogen concentration is unlikely to be a viable option for a severe accident. Since the reviewed revision of the TBD was prepared before the staff and ir'dustry reached decisions regarding coverage of severe accdents, coverage of containment in the TBD should be consistent with such decisions.

. This issue is closed with respect to the ATOG and TBD reviews.

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1 3.4.6 Anticioated Transients Without Scram )

Open issues. The following open issues were identified:

1. Provide additional ATWS guadance prior to implementation.  !

2.~ Guidance should be updated to reflect the rulemaking on ATWS. 1

3. Address not isolating letdown during ATWS.

Staff Findinos Guidance was provided as described in Reference 3 and further clarifications are provided in TBD Section Ill.A, immediate Actions, Steps 1 and 2, and in TBD Volumes 2 and 3. The issues are, therefore, closed.

3.4.7 Pressurized Thermal Shock .

Open issue Guidance should cover exceeding pressure / temperature limits.

Staff Findina. Concerns pertinent to this topic are covered in such Revision 6 steps as 111.B.18, Ill.C.3, and Ill.D.1, and in VI-3 " Pressurized Thermal Shock (PTS) Rule." A bases section is provided in TBD Volume 3 and Revision 7 adds numerous clarifications. Acceptable coverage i of PTS is provided and, therefore, the issue is closed.

3.4.8 Cooldown and Interfacing With Other Procedures l Ooen lasues. The following issues were open at initiation of this audit:

1. . EPGs should cover RCS voiding during natural circulation and cyclic boiling-condensing conditions.

2. Operation of the DHR syd.em should be evaluated for saturated RCS conditions. Topics of potential concern include the RV level instrumentation system, DHR flow rate control, minimization of the likelihood of DHR loss during switchover, and plar.t-specific DHR j suction line configuration influence on operability.

3. Guidance should be provided for interfacing with non-EOP procedures and for inte:Tacing within sections of the EOPs.

4. The order of depressurization operations such as use of the pressure / power operated relief valves (PORVs) vs. auxiliary pressurizer spray should be clear. i Staff Findinas RCS voidmg during natural circulation conditions is covered in Steps Ill.B.3, Ill.B.8, the end of Step lli.B.15, numerous instructions regarding high point vent operation, in procedure NC,

" Natural Circulation Cooldown," in Recovery Subsection SS-2 " Establish Heat Transfer to Non-Operating SG," in other procedures, and in the bases discussions for those procedures.

Additionalinformation is provided in TBD Volume 3. Cyclic boiler-condenser conditions are 19 L

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l addressed in dealing with the syinptoms of voiding and in Volume 3 Step lll.B.2.9. Item 1 is adequately addressed.

' item 2 is outside the scope of the staff's requirements for revisions to the TBD, which is stated in Volume 3, page 1.B-1 as: "The scope of the TBD covers transient initiation from all plant modes other than decay heat removal system operation." (The staff notes that the other PWR vendor owners group's GPGs are limited to RT or conditions requiring RT.) The staff has accepted the  ;

scope of the TBD and consequently, item 2 has been adequately addressed. l The staff additionally notes that TBD coverage has been expanded in Revision 7 to address aspects of shutdown operation.

Interfacing with non-EOP procedures (Item 3) and within sections of the EOPs was a repeated problem in ATOG. The cross-referencing problem has been eliminated in the TBD. Interfacing with non-EOP procedures was addressed in SER Sections 3.4.1.2 and 3.4.4.2.

Numerous improvements in cooldown guidance have been provided and the question of cooldown techniques has been resolved. 1 I

All aspects of issues addressed in Section 3.4.8 of this SER are satisfactorily addressed except i resolution of level instrumentation concerns. This is addressed separately in Section 3.4.2 of this SER and allissues listed in Section 3.4.8 of this SER are, consequently, closed. l 3.4.9 Hiah-Pressure Iniection Coolina Open issuqa. The following are the HPl injection cooling issues: 1

1. The TBD guidance should be verified, revised where necessary, and included as HPl actions in the guideline.

2. The TBD should be upgraded, if necessary, to address inoperative PORVs located on the pressurizer and to address the success of HPI cooling.

Staff Findinas. HPi cooling operation is adequately addressed by numerous entries at appropriate locations in the TBD, such as for loss of SCM and during cooldown. Volume 3 and numerous bases discussions also address this topic. The issues are, therefore, closed.

3.4.10 Steam Generator Ooeration 3.4.10.1 Steam Generator Control Open issue. SG control is to be fully addressed.

Discussion and Findinas This requirement covers numerous items that originated from the ATOG evaluation (Reference 2). ' All are closed for tracking purposes as discussed in the following list. Most are 20 l

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closed on the basis of satisfactory completion of additional work. Several are redefined to correspond to the TBD and are addressed as TBD-applicable issues. These are closed with respect to the ATOG evaluation. ' Aspects pertinent to SGTR are closed as discussed in Section 3.4.10.4 of this SER.

The staff findings include the following (where the items are paraphrased from the original issue definitions to eliminate duphcatson and to gmup similar issues together):

1. Establish cooldown time Cooldown times have been established as illustrated in Section 3.4.10.4 of this SER. The issue is closed.

2. Show operator actions are effective for cooldown time and redioedive material release and better address SG isolation vs. steamino as mitiaation strateoies. The B&WOG has evaluated the effectiveness of operator actions, such as SG isolation versus steaming, and has calculated release behavior and dose. This issue is addressed in Section 3.4.10.4 of this SER.

3. Address multiole failures in either or both SGs. address multiple failures involvina more than iust SGTR. and address multiole feih__wes for RCP restart. TBD Section Ill.E covers combinations of multiple failures with SGTR via the symptom-based approach. A hierarchy of symptoms is basic to the TBD approach and govems transfer to higher priority symptom treatment regardless of when the need occurs or.where the operator is located within the guidance. This prioritization satisfactorily addresses dealing with mulhpie failures on the basis of the symptoms and actions to mitigate combinations of events on the basis of those symptoms. These issues are, therefore, closed.

Weaknesses were identivied in clarifying symptom priority and in addressing certain symptoms that appear after initial actions during an event. These are discussed in Sections 3.3.2,3.3.4 and 3.3.5 of this SER.

4. Include ICC conditions and totalloss of feedwater. These are satisfactorily addressed via the symptom treatment as described in item 3, above. These issues are, therefore, closed.

5. los.lude restrictions on HPi coolina. HPl cooling is addressed within the SGTR guidance (TBD Section Ill.E). SG level control with HPl in operation is addressed at Step 6.1.

Makeup to the borated water storage tank (BWST) is identified at Step 7.3. HPl usage in response to TRACC limits is covered beginning at Step g.0. Restrictions with respect to cooldown rate are identified at Step 7.0 and, with respect to PTS, at Step 8.0. PTS restrictions associated with HPl cooling are covered at Step 10.2 and at other locations, and have been further addressed in Revision 7. Further guidance is provided in cooldown sections, in HPl operation, and in PTS specific rules, with supplemental bases information provided in TBD Volumes 2 and 3. Similar coverage is provided within other sections as described in item 3, above. This issue is, therefore, closed.

6. Address vanous souloment failures and operator errors that occur durina mitiaation of the eye.01. Examples, such as altemate feedwater sources during treatment of loss of 21

heat transfer and during cooldown, were provided as separate issues during the ATOG review. The process identified in item 3, above, applies. For example, guidance regarding feedwater during response to a loss of heat transfer is provided in TBD Step 4 of Section Ill.C. The issue is, therefore, closed.

7. Address natural circulation and voidina. Natural circulation (NC) is addressed in Steps 6.0 and 8.0. Voiding is addressed in Step 12.0. Cooldown with NC is covered in either "SBLOCA/HPl Cooldown" for the RCS saturated or "NC Cooldown" for the RCS subcooled. These topics are additionally addressed in other sections, such as " Loss of SCM" and " Lack of Heat Transfer." Extensive bases material is provided in Volumes 2 and 3. The issue is, therefore, closed.

8. Address PORV and sorav failures durina dooressurization activities. The closed loop of Steps 9.3.c through 9.3.e that could result from PORV and spray failure can be left via attemate paths, such as using high point vents or realization of lack of SCM. Use of letdown, turbine bypass valves, and atmospheric dump valves (ADVs) is also identified.

PORV failure when attempting HPI cooling is addressed at Step 10.2.d. These conditions are addressed at other locations as well. For example, PORV failure when treating loss of SCM is addressed in Step 8.3 and a potential contributor to spray failure is addressed in Step 18.0. In addition to the above, use of the pressurizer and attemates are identified in cooldown guidance. The issue is, therefore, closed.

9. ' Provide better control of RCS-SG secondarv fluid transfer. orovide cuidance for an overfilled SG. and provide attemate control of SG level. Secondary fluid transfer is addressed in Steps 9.0 and 10.0 and control with respect to the TRACC limit is covered in Step 9.0. TBD Volume 3 discusses various SG level control approaches.

The TRACC limit is configured so that there is no need for the operator to know the tube leak rate, which is difficult to determine immediately after initiation of a SGTR. Instead, it is based on readily available information such as reactor coolant (RC) contamination data, measured onsite and offsite dose rates, and observed SG level. Viable attematives are provided if SG level control attempts are ineffective.

These issues are, therefore, closed.

The SGTR mitigation methods used and their influence on RCS-to-SG fluid transfer are discussed in Section 3.4.10.4 of this SER.

10. Provide imoroved lono-term deoressurization. Depressurization is addressed by Steps 7.6,8.0,9.0,10.4,10.5, and 10.6 in the SGTR guidance. Depressurization following that guidance is addressed by "SBLOCA/HPl Cooldown" for inadequate SCM,

" Forced Flow Cooldown"if RCPs are running, and "NC Cooldown"if RCPs are not running and NC is to be used. This ATOG issue is satisfactorily addressed and the issue is, therefore, closed.

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One method of feeding SGs and accomplishing cooldown/depressurization, trickle feed, is not addressed in TBD Volume 1 although it is discussed in Volume 3. This is addressed in Section 3.4.10.3 of this SER.

11. Provide better control of environmental releases. The TBD and work accomplished in support of the TBD provide significant changes in environmental understanding and control when compared to the older ATOG. This ATOG issue is closed since it is superseded by the new approach and the staffs evaluation of that approach, as discussed in Section 3.4.10.4 of this SER. '

I 3.4.10.2 Steam Generator LevelIndication 1

Open issue. Licensees shall:

• Consider the potential error in indicated SG levels, including the effect of condensing pots if steam is not available for condensation in the pots.

• Assure that suitable redundant indications are provided.

Staff Findina. The vendor believes that this is specific to each plant and need not be addressed in the TBD. The staff concurs and the issue is closed on this basis.

3.4.10.3 Steam Generator Trickle Feed Discussion. A!! hough TBD Revision 6, Volume 1 addresses cooling due to feeding SGs for purposes of contiolling SG level, it provides little consideration of alternate cooling methods, such as trickle feed, when a SG is not available using other means. Instead, the fallback action of HPI cooling is used. This approach was based on the depth of understanding for various cooling approaches. The B&WOG considered HPl cooling to be well analyzed and understood.

Conversely, trickle feed was not well analyzed at the time of preparation of TBD Revision 6, nor had it been tested. The B&WOG also expressed a concern that trickle feed could damage SGs sufficiently that replacement would be necessary, which is the major reason for not including trickle feed in Volume 1, although the concern did not extend to the point where damage would cause a LOCA. Consequently, SG analyses were performed to address trickle feed. Even with completed analyses, some licensees may prefer to use HPl cooling.

Staff Findino. As stated above, according to the B&WOG, there appears to be little or no concern that trickle feed will cause a LOCA. Consequently, unless it is established that trickle feed is a significant safety concern, the staff found that lack of trickle feed guidance is an omission of a potenttalty viable cooling option and, thus, trickle feed should be added to Volume 1 (with bases information in appropriate volumes). The option of a preference of trickle feed over HPl cooling versus HPl cooling over trickle feed may be left to the licensees provided that both options are included in the TBD. The B&WOG has completed its evaluation of SG trickle feed (References 9 and 10). This work was completed after publication of TBD Version 6 and is in the same category as other work addressed by References 9 and 10 where the staff 23 I

G does not plan a generic review. The issue is closed for purposes of this SER. The staff notes, l however, that aspects pertinent to the response of damaged SG tubes were not addressed in its review.

l 3.4.10.4 Steam Generator Tube Ruoture l

Discussion All PWR vendors / owners believe it is important not to open SG safety valves during SGTR because loss of a control of the release will occur if a safety valve opens and subsequently fails  !

to close. Consequently, all PWR EPGs provide guidance to steam SGs with broken tubes until conditions are reached where there is reasonable assurance that SG safety valves will remain closed following termination of steaming. )

1 The B&W NSSS response to a SGTR differs from NSSSs equipped with "U"-tube SGs in two significant ways:

• A full power RT in the B&W design will cause SG safety valves to open.

e For practical purposes, a SG tube leak in the B&W-designed SG cannot be stopped I

without violating SCM requirements unless the SG secondary becomes water-solid.

isolating a SG to stop the leak may remove its cooling capability for the duration of the event and, if offsite power is not available, usually extends the time until the DHR system can be operated. There is concern that the increased time and unusual operating configuration may increase the likelihood of complications during cooldown.

Consequently, the vendor-recommended mitigation differs from other PWR owners group EPGs in severalimportant ways. The vendor-recommended mitigation provides the following:

• Operators are to reduce power prior to RT/ turbine trip so that SG safety valves will not open following RT. Steaming to the condenser will continue for about a half hour while power reduction is accomplished.

e Operators are to continue to steam both SGs until reaching DHR initiation conditions unless one of the following tube rupture attemate control criteria (TRACC) limits is reached:

1.5 rem (0.015 Sv) thyroid or 0.5 rem (0.005 Sv) whole body at the site boundary (some B&W licensees have a technical specification of 1.5 rem to the thyroid that was intended to meet the pre-1991 10 CFR 20 requirements).

The borated water storage tank low level setpoint is reached.

i The SG is filled.

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The SG containing broken tubes, or the greatest s'fected SG, is usually isolated for non-B&W NSSS designs; thus, theoretically, terminating the release from that SG. The release can continue with the vendor-recommended mitigation scheme.

The vendor's objectives are to minimize the possibility of complications that could extend the event or lead to core damage by-

• Decreasing system cooldown time.

• Minimizing vulnerabilities due tc additional failures.

e Reducing operator burden by keeping the plant in a normal configuration.

)

e Minimizing SG tube stress.

e . Keeping both SGs in service in case a SG becomes unavailable.

The vendor provides an alternative to steaming the broken SG if TRACC limits are reached, which is to isolate the affected SG. - As a further backup, once through cooling can be initiated using the HPl pumps and pressurizer PORVs, and both SGs are isolated. However, if one HPl pump is used, it will be about 10 to 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> before the RCS can be cooled while maintaining pressure below 1000 psi to prevent opening SG safety valves. Two HPl pumps might be able to provide adequate cooling about three hours after SGTR initiation. Consequently, if one were to commit to two-pump operation, and one was subsequently lost in less than 10 to 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br />, it may be impossible to regain use of a SG, and RCS pressure would consequently exceed the SG safety valve relief pressure.

The conhnued steaming and associated continued release of radioactive materialis potentially of concem since, intuitively, one would prefer to immediately terminate all release of radioactive material. Therefore, the staff examined the vendor-recommended approach in some detail (Reference 23).

The vendor's TRACC limit evaluation did not use many of the licensing-approach conservatisms specified in the Standard Review Plan (SRP). Doses were obtained for 116 weather conditions at the site judged to provide the highest site boundary dose. The instantaneous maxima at any site boundary location were summed throughout each of the weather conditions to obtain a total l dose for each condition.

If the steam lines were isolated and the atmospheric dump valves were used for steaming, the vendor found that the whole body dose was less than approximately 10 mR (0.1 mSv) and that  ;

the thyroid dose dominated release consequences. Releases were negligible if offsite power was available because steaming would normally be to the condenser with a 10' decontamination factor for iodine. Predicted doses for steaming directly to the environment are summarized in Table 1, which also relates the steaming time to the leak rate that would fill the SG, release 99 percent of the initial iodine, and that was used to establish TRACC steaming limits. For example, if the leak rate was 1263 gpm, the SG could be steamed for one hour, at which time the SG would become water-solid. If the leak rate was 6400 gpm, then 99 percent of 25 j

the iodine initially contained in the RCS would have been released from the RCS in one hour in comparison, a single double-ended tube break will pass 370 gpm at 2200 psig and 580 *F, the average temperature for most B&W plants. On average, this decreases to 145 gpm with an SCM of about 50 *F. If on natural circulation, the rate will be approximately 175 gpm at 2 hrs Table 1. Limiting Leak Rates and Thyroid Dose for Steaming to the Environment minunummilinumiminia Leak rate in gpm that Thyroid Dose,

... ....................... .... Rern ,

Steaming will will was used Mean Peak Comments duration, fill release to set hrs the 99% of TRACC SG' initial steaming iodine limit i 1263 6400 1313 0.04 0.38 1313 gpm corresponds to three HPI pumps at 1225 l psia 4 445 1600 533 0.09 0.97 12 hr 533 gpm steaming limit used because it fills SG 292 533 l 12 533 0.20 1.58 '

24 170 267 267 0.22 1.49 50 106 128 130 0.22 1.33 130 gpm selected as bounding to reduce number of calculations 120 <100 53 130 0.22 1.34 130 gpm selected as above and 130 gpm at 10 hrs. If cooling with RCPs operating, the pump NPSH becomes restrictive during cooldown and SCM will increase, increasing the leak rate.

The peak thyroid dose shows a maximum at 12 hrs. This behavior is attributed to variation introduced by the number of samples used in the weather sampling methodology.' Also of interest is that the fraction of iodine remaining in the primary system ranged from 0.403 after

'As discussed at the beginning of Sectat 3.4.10.4 of this SER, a SG leak in a B&W-designed SG cannot be stopped unless the SG secondary side is completely filled with water.

'The calculated dose is based on 116 samples. The vendor did not repeat the calcu!ations for an increased number of samples because it judged that sufficient precision had been obtained to support its conclusions. (Using more samples in the calculation base would theoretically reduce calculation

• noise.") The staff agrees with the vendor decision.

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, ' steaming for one hour to 0.0002 after 120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br /> of steaming. Note that these releases are for i

(' steaming directly to the environment, a condition that would be maintained only as long as the l condenser was not available. Given a LOOP in conjunction with a SGTR (an unlikely event),

I continued condenser unavailability due to a LOOP for more than a few hours is unlikely.

The staff generated SGTR event trees to provide insight into plant response for two situations (Reference 17):

1

, .e Isolate. Isolate the SG with ruptured tubes as soon es is practical.

• Steam Steam both SGs until DHR is operating.

For SGTRs smaller than HPl capability, the staff predicted SG isolation would have a 30 percent i higher core damage frequency (CDF) than steaming both SGs, a prediction that must be treated with caution since many assumptions were used and prediction uncertainty is high.

Limited sensitivity calculations showed that in no case did steaming reduce CDF by more than a factor of two. Apparently, one or more branch probabilities or recovery factors must be significantly in error for steaming to provide an order of magnitude reduction in CDF.

The staff's SGTR study was limited to the most likely SGTR and did not include SGTRs where the break size was larger than HPl capacity, an occurrence with a frequency smaller than i one-tenth of the analyzed cases. In addition, it did not include LOOP in conjunction with SGTR, l where the frequency is 104/ reactor-year. l l

The vendor-calculated dose and the staff CDF predictions are summarized in the Table 2.

Although no dose is provided for the isolation case with LOOP, the staff notes that a significant  ;

fraction of the release is expected before an initialisolation is provided because of the need to 1 provide initial power reduction and cooldown.

Table 2. Dose Prediction and Core Dama e Fre uenc nummmmmmmmmmmmmmmmmma item CDF, number per Thyroid dose if no core ,

reactor year damage, rem l Steam isolate Steam isolate SGTR no 3X104 4X104 -0 -0 LOOP SGTR, LOOP <104 <104 <0.2 -

Staff Findinos A SGTR violates the RCS pressure boundary and seriously jeopardizes the containment by introducmg a potenhal path for RCS coolant loss directly to the environment. It is vital that fuel clad integrity is retained and it is important that the integrity of the remaining containment control 27 L

be maintained. Fuel damage with loss of remaining containment control would be a threat to public health and safety. If the fuel clad is intact, the TBD mitigation schemes would control total offsite releases to a very small fraction of 10 CFR 100 values and usually to less than 10 CFR 20 limits for annual release to uncontrolled areas. Releases will generally be negligible if functional containment control is also retained because, for practical purposes, releases will be limited to inert gases. The priorities are clear. A SGTR mitigation scheme that first and foremost protects the fuel clad will provide the best protection of the public health and safety.

The second priority is containment control which,' if retamed, can mitigate a release even with fuel damage.

l The staff's limited analyses show little difference in CDF for mitigation with an " isolate the j affected SG as soon as is reasonable" philosophy in contrast to " steam both SGs for the  !

duration of the event," a result that would appear to favor isolation. However, the staff notes that l isolation does not mean there is no release since immediate isolation is not oractical in the B&W  !

NSSS design. Further, the staff notes that human factors strongly influence's conclusions, and l the benefit of the vendor's recommendations to keep the plant in a familiar condition and to {

minimize cooldown time may not be adequately reflected in the numerical results.

Consequently, the staff cannot conclude that isolation is the most reasonable path consistent with protection of the fuel clad for all situations. Plant-specific considerations or unique aspects of an event may strongly favor steaming to both protect the core and to attain a cooldown as quickly as is practical.

The vendor's dose calculations indicate that little release will occur as long as steaming is to the condenser and core damage has not occurred. A small release is predicted if the condenser is not available. The limits included in the TRACC approach provide a reliable upper limit on a release that is based upon information that is readily available during the event.  ;

The staff finds that the protection of the public health and safety is best achieved if the choice of steaming /not steaming the affected SG is made by the individual licensees so that plant-specific ,

and event-specific considerations can be employed.

in regard to the content of TBD Revision 6 Volumes 1 and 2, the staff found that the following changes should be incorporated into a revision to the TBD:

• The equivalent of TBD Volume 3 Figures Ill.E-2 and lil.E-3 that provide steaming time as a function of pre-existing 1-131 and transient peak l-131, respectively, should be incorporated, and e The TRACC radiation criteria that result in doses that are no greater than discussed above should be incorporated (TBD Volume 1 does not provide values).

~

The B&WOG scheduled completion of the SGTR work by December 15,1996 (Reference 9 and 10), and reported completion on August 28,1996 (Reference 11).

Based on our review and analyses, the ATOG issues related to SGTR mitigation are closed.

28

o 3.4.11 Deoraded Core An unnecessary transfer to ICC guidance was possible in the ATOG due to short term superheat indication that would be expected during a large break LOCA. This has been

- corrected in the TBD. See, for example, TBD Volume 2 pages 111.B.17 and Ill.B.1.

Elimination of steam from the SGs and the RCS to enhance potential heat removal actions prior to reaching extreme ICC conditions was identified as a potential concem. This is addressed by such TBD ICC Steps as 2.0 and 4.0.

1 The TBD's ICC guidance extends to cooldown, with extensive coverage of some areas, thus satisfying another ATOG issue. Consistent with the staff's expectation for continued improvement, TBD Revision 7 adds further guidance in this area.

The general approach continues to be one of using whatever equipment is available that will

. assist in mitigation, thus satisfying any concems related to individual equipment failures.

Other aspects of this category are acceptably addressed as discussed in Section 3.3.3 of this SER. The issues are, therefore, closed.

3.4.12 Reactor Coolant System Inventory and Pressure Open issues The open issues are as follows:

1. EPG coverage of multiple failures should include cyclic boiler / condenser phenomena.

2. EPG coverage should include small break LOCA without HPl.

3. The ATOG Part ll discussion of low pressure injection covers such items as dealing with blocked sump recirculation flow. This topic should be updated and expanded to include switchover to recirculation flow and guidelines should be provided.

4. Analyze and provide attemate methods for control of RCS pressure, including inoperative PORVs and no pressurizer spray.

Staff Findinos. Boiler / condenser cooling is identified in, for example, TBD Volume 3, Section ll.D.2.3.2. The cyclic phenomena are discussed in Section Ill.C.3.5.D. LOCA without HPl is addressed in such steps as Ill.B.8. Switchover is addressed in such cooldown sections as HPl CD-4 step 9, with suitable reference to plant-specific procedures. Alternate methods are inherent to the TBD in that each symptom is addressed and potentially available equipment is

- considered to mitigate undesirable characteristics. For example, high point vent operation is included that will assist in depressurization. The staff finds that acceptable guidance is provided and these issues are, therefore, closed.

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p l

3.4.13 Analyses Open issues. . Many of the analysis issues were requests for work or information that did not specifically need to appear in ATOG or the ATOG bases. Often the staff only needed to establish that the maary understanding existed in others, the vendor incorporated results into the TBD bases or provided it to the staff separately so that the staff could complete its review. For example, Section 3.4.10, above, included such information. Many of the identified areas of concem are also addressed in sections of this SER such as Sections 3.4.3 and 3.4.12, above. A few are specifically addressed below.

1. Provide analyses as needed to indicate which system parameters can be used to cuide operator actions. Information in TBD Volume 3 Chapter 11.B addresses this issue. See also Reference 24 for an example of considerations.

2. Provide analyses as needed to demonstrate the feasibility of a recoverv technioue for some well defined accidents. TBD Volume 3 provides results. See also Reference 25 for examples.

l

3. Provide analvses as needed to ootimize or to select the best recovery oaths. Optimum or best recovery paths are not necessary for successful mitigation.

4. Verify the codes used to calculate NSSS behavior by comoarison to exoerimental systems data. Continue to provide adeauate thermal / hydraulic au.a.nititative r data

- aoolicable to B&W systems. Provide carticular attention to uniaue asoects of the desian such as the hot lea and SG confiourations. Data have been obtained and verifications have been performed. Results are incorporated into the TBD Volume 2 and Volume 3 bases.

5. Cover transients where known or susoected unusual behavior is of concem. Include l unoer head voidina. other RCS voidina. oscillatory flow behavior in the RCS. oscillatory i pressure behavior. and lor'a term cooldown. Examine reasonableness of calculated 1 behavior and assumptions. Results are incorporated into the TBD Volume 2 and Volume 3 bases.

Staff Findinas. The staff is satisfied with the analysis bases for the TBD-provided guidance and the ATOG issues are, therefore, closed.

3.4.14 Miscellaneous ,

This section addresses miscellaneous items that were not included in the other sections. Each j

is listed and addressed separately below:

1. Open issue. Usage of words that have an unclear meaning, such as "very rapidly" and

" slow," should be avoided. Such terms as "when" are not always clear with respect to whether other actions should continue.

1 I l 30

i L Staff Findina. ATOG is no longer in use and the TBD generally avoids nonspecific

[ usage. The issue is closed.

i

2. Ooen Issue. ATOG part ll discusses actions if sump recirculation flow is blocked and l cannot replace lost coolant. Such actions will be successful only if the RCS has no significant leaks below the DHR system suction elevation. Expand upon this, including the general topic of use of the LPI system to maintain RCS inventory, and provide operator guidelines.

l Staff Findina. Numerous parts of the TBD and TBD bases address switchover to sump recirculation flow, including consideration of an adequate level in the sump and making sure there is adequate NPSH. (See, for example, LOCA cooldown guidance and TBD Volume 3 Section IV.b.) The spec?ic action for a blocked sump is not addressed, but throttling to provide adequate NPSH is a potentially effective action for a partial sump l blockage. Reference is also'made to plant-specific procedures for recirculation operation. This is acceptable and the issue is closed.

3.4.10 Non-Hot /Non-Pressurized System Conditions Comment. Issues applicable to a "non-hot /non-pressurized" NSSS may md to be addressed further, i . \

l Staff Findina. As discussed above, this issue is outside the scope of the TBD. It was also l addressed by the Commission's direction for the conduct of shutdown operations (References )

26 through 28) and no action is necessary herein. Such issues are closed with respect to ATOG and TBD review.

l 3.4.16 Containment I 1

l Open issue Containment must be covered in the long term.

l Staff Findina. The TBD provides containment coverage as identified Section 3.4.5 of this SER.

Future containment coverage improvements can be provided by the B&WOG long term TBD improvement program and by interfacing with non-EOP plant procedures, as appropriate. This issue is closed 3.4.17 Loss-of-Coolant Accident Outside Containment l

Ooen issue. LOCA outside containment must be covered in the longer term to a depth l consistent with its potential effect on public safety.

Staff Findina. Following its audit of TBD Revision 6, the staff determined that this issue should  !

be examined and suitable guidance provided, as appropriate, following the examination. It further determined that the examination should include consideration of the potential benefits of addressing LOCA outside contamment and of the drawbacks associated with potential inappropriate actions if a diagnosis is required. This issue is closed for purposes of this SER, as discussed in Section 1, since the B&WOG planned to address LOCA outside containment 31 1

(

l '_

- (References 9 and 10),~and reported that they had done so on September 12,1997 (Reference 12).-

3.4.18 Severe Accidents l Comment. Coverage of core damage with gross fuel melt and fuel relocation may be necessary l when the NRC provides guidance.

l l- Staff Findina. This issue is closed as discussed in Section 3.3.3 of this SER.

3.4.19 Loss of AC Power l

Comment it may be necessary to address areas such as control of core damage and control of releases for the period beyond that determined to be in compliance wit'n 10 CFR 50.63.

Staff Findina. This issue is within the scope of severe accident guidance rather than EPG review and is closed on this basis.

4 CONCLUSIONS All outstanding issues are closed with respect to this generic review. One unresolved issue, use  ;

i- of level instrumentation, may be addressed on a plant-specific basis. Consequently, the staff  !

has no plans for further generic review of the TBD. j i

5 REFERENCES l l

1. " Abnormal Transient Operating Guidelines (ATOG)," Oconee Nuclear Station, Unit 3, Parts I and ll, Babcock & Wilcox, Nuclear Power Generation Division,74-1123297-00,

- March 1982. I

2. > Eisenhut, Darrell G., " Safety Evaluation of ' Abnormal Transient Operating Guidelines',

! (Generic Letter 83-31)," Letter from Director, NRC Division of Licensing to all operating l

reactor licensees, applicants for an operating license and holders of construction permits for Babcock & Wilcox pressurized water reactors, September 19,1983.

3. Crutchfield, Dennis M., " Abnormal Transient Operating Guidelines (TMl Action l

' Item I.C.1)," Letter from Chief, Operating Reactors Branch #5, NRC to Daniel D. Wl;itney, Chairman, Operator Support Subcommittee, B&W Owners Group, December 14,1983.

4. Deatherage, D., Letter from Chair, B&WOG Owners Group Operator Support Committee to Warren Lyon, NRC, OG-992, February 12,1992.

5. " Emergency Operating Procedures Technical Bases Document," Revision 6, (Volumes 1,2, and 3), The B&W Owners Group Operator Support Committee, 74-1152414-06, December 31,1991.

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6. " Emergency Operating Procedures Technical Bases Document," Revision 8, (Volumes 1,2, and 3), The B&W Owners Group Operator Support Committee,74-1152414-06, November 1,1996.

7. Birmingham, Joseph L., " Summary of June 17,1999, Meeting with the Babcock & Wilcox Owners Group to Discuss Emergency Operating Procedure (EOP) Inspections," NRC Memorandum to Cynthia Carpenter, Chief, Generic issues, Environmental, Financial and Rulemaking Branch from Project Manager, Generic Issues, Environmental, Financial and Rulemaking Branch, June 29,1999.

8. Holahan, Gary M., " Babcock and Wilcox Owners Group (B&WOG) Emergency Procedures Guidelines Review (The Technical Bases Document), TAC No. M54946,"

Letter to Ed Jacks, Chairman, B&WOG Operator Support Committee from Director, Division of Systems Safety and Analysis, NRC, February 17,1995.

9. Jacks, Ed, " Letter from E. Jacks (Entergy Operations, Inc.) to G. M. Holahan (NRC),

OG-1487, dated March 20,1995," Letter from Chairman, B&W Owners Group Operator Support Committee to Gary M. Holahan, Division of Systems Safety and Analysis (NRC),

OG-1522, June 13,1995.

10. Jacks, Ed, " Letter from E. Jacks (Entergy Operations, Inc.) to G. M. Holahan (NRC),

OG-1522, dated March 20,1995 (sic)," Letter from Chairman, B&W Owners Group Operator Support Committee to Gary M. Holahan, Division of Systems Safety and Analysis (NRC), OG-95-1546, October 2,1995.

11. Stallard, A. R., "ATOG Safety Evaluation Report," Letter from Chairman, B&W Owners Group Operator Support Committee to Gary M. Ho!ahan, Division of Systems Safety and l Analysis (NRC), OG-1607, August 28,1996.

12. Bremer, Ross F., "ATOG Safety Evaluation Report," Letter from Chairman, B&W Owners Group Operator Support Committee to Gary M. Holahan, Division of Systems Safety and Analysis (NRC), OG-1670, September 11,1997.

13. "TMI-2 Lessons Learned Task Force status Report and Short-Term Recommendations,"

NRC, NUREG-0578, July 1979.

l 14. Linn, Mark A., " Closure of ATOG SER lssues," Letter from Chair, B&W Owners Group l Operator Support Committee to Guy Vissing (NRC), September 5,1986.

! 15. "NRC Action Plan Developed as a Result of the TMI-2 Accident," NRC, NUREG-0660, Published May 1980, Revised August 1980.

, 16. " Clarification of TMI Action Plan Requirements," NRC, NUREG-0737, November 1980; NUREG-0737 Supplement No.1, January 1983.

33 8

r i I '

t l 17. Whitney, D. D., " Supplement to ONS-3 Final ATOG," Letter from Chairman, B&WOG Operator Support Committee to Darrell G. Eisenhut, Division of Licensing, NRC, July 2, 1983.

18. Whitney, D. D., "B&W Owners Group Plan and Schedule for Addressing the Safety l Evaluation of ' Abnormal Transient Operating Guidelines,'" Letter to Darrell G. Eisenhut, NRC, from Chairman, Operation Support Committee, B&W Owners Group, OSC 83-21, December 9,1983.

19. Atherholt, David W., Letter from Chairman, B&W Owners Group Operator Support Committee to Warren Lyon, NRC, OG-1308, November 12,1993.

20. " Emergency Operating Procedures Technical Bases Document," Revision 7 update of Revision 6, (Volumes 1,2, and 3), The B&W Owners Group Operator Support Committee,74-1152414-07, August 20,1993.

I

21. Lyon, Warren C., Resolution of issues identified in the 1983 Safety Evaluation of the B&W Abnormal Transient Operating Guidelines (ATOG), (TAC NO. M54946)," NRC memorandum to Robert C. Jones, Reactor Systems Branch, June 28,1995. i

22. Lyon, Warren C., " Meeting Report - Meeting at B&W (Lynchburg, VA) on December 1 - 3,1992...," NRC Memorandum for Robert C. Jones, Reactor Systems i Branch, February 2,1993. 4

23. Lyon, Warren C.," Examination of Steam Generator Tube Rupture (SGTR) in Babcock and Wilcox (B&W) Nuclear Steam Supply Systems (NSSSs), TAC M54946," NRC Memorandum to Chief, Reactor Systems Branch (NRC), March 6,1995.

l

24. Lyon, Warren C., " Steam Generator Tube Rupture (SGTR) Event Trees for Comparison of isolation of the Affected SG with Steaming the Affected SG for Babcock and Wilcox l

Nuclear Steam Supply Systems (TAC No, M54946)," NRC Memorandum to Chief, Reactor Systems Branch (NRC), April 24,1995.

25. " Technical Advisory Group Investigation of Once Through Steam Generator Thermal i Hydraulic Experimental Data Requirements," B&W BAW-2079, March 1989.

26. L. Joseph Callan," Issuance for Public Comment of Proposed Rulemaking Package for

! Shutdown and Fuel Storage Pool Operation," Rulemaking issue for the Commissioners I

from Exocutive Director for Operations, NRC, SECY-97-168, July 30,1997.

27. Hoyle, John C., " Staff Requirements - SECY-97-168 - Issuance for Public Comment of Proposed Rulemaking Package for Shutdown and Fuel Storage Pool Operation," NRC Staff Requirements Memorandum (SRM) to Executive Director for Operations from Secretary of the Commission, December 11,1997.

l 28. " Shutdown and Low-Power Operations for Nuclear Power Reactors," Federal Register, Volume 64, No. 23, page 5623, February 4,1999.

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