ML20247N483
| ML20247N483 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 05/31/1989 |
| From: | Michael Ray TENNESSEE VALLEY AUTHORITY |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 8906050369 | |
| Download: ML20247N483 (35) | |
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-i TENNESSEE VALLEY AUTHORITY CHATTANOOGA, TENNESSEE 37401 SN 157B Lookout Place MAY 3iB89 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Hashington, D.C.
20555 Gentlemen:
In'the Matter of
)
Docket No. 50-260
-Tennessee Valley Authority
)
BROWNS FERRY NUCLEAR PLANT-(BFN) - RESPONSE TO NRC INSPECTION REPORT NO. 50-260/88-200 ON EMERGENCY OPERATING INSTRUCTIONS (E01s)
-This letter transmits TVA's reponse on the BFN E0Is as requested in the NRC letter from J. G. Partlow to S. A. White dated September 28, 1988.
, provides TVA's response to the NRC concerns identified in section 3 of the referenced report, which included the development and-implementation of the E0Is, evaluation of containment venting procedures,.
evaluation of personnel access to the reactor building during emergencies, and review of deviations of the E0Is'from the documents that were used to develop them. - Enclosure 2 contains a list of the commitnients made in this submittal.
If you have any questions, please telephone Patrick Carter at BFN, (205) 729-3570. '
Very truly yours, TENNESSEE VALLEY AUTHORITY f'
W(f Mana r, Nuclear Licensing and Regulatory Affair:
Enclosures cc: See page 2
- p k(.k b 60 gi An Equal opportunity Employer i
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MAY 311989
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ul-U.S. Nuclear Re1ulatory Commission cc (Enclosures):
Ms. S.
C.' Black, Assistant Director.
for Projects TVA Projects 1 Division U.S. Nuclear Regulatory Commission One White Flint, North 11555 Rockville Pike-Rockville, Maryland 20852 B. A. Wilson, Assistant Director for Inspection Programs TVA Projects Division U.S. Nuclear Regulatory Commission
. Region II h
101-Marietta Street, NW, Suite 2900 Atlanta, Georgia.30323 Browns Ferry Resident Inspector
-Browns Ferry Nuclear Plant Route 12, Box 637 Athens, Alabama 35609-2000 i
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k ENCLOSURE 1 RESPONSE TO NRC INSPECTION REPORT NO. 50-260/88-200 LETTER FROM J. G. PARTLOW TO S. A. WHITE' DATED SEPTEMBER 28, 1988 Note:
For the purposes of this enclosure, " rest' art" refers'to unit 2, l
cycle 6 restart unless specifically stated otherwise.
Section 3,1,1 - Comparison of BfN Plant-Specific Technical Guidelines (PSTGs) withBfNEmergencyOperatingInstructions(E0Is).
1.
NRC's Concern
" Step RC/L-1 in the E0Is included the diesel generators.as a system to be checked and activated.
The'BFN staff had specifically eliminated reference to the diesel. generators in this section of the PSTGs.on the basts that the injection systems at BFN did not depend on AC power, (e.g.,
high pressure coolant injection (HPCI) ar.d reactor core isole+ %n cooling.
(RCIC)_ systems)."
TVA's Response The_ reference to the diesel generators has been removed from step RC/L-l'in E01-l '.
- 2. 'NRC's Concern
" Step RC/L-1 in the E0Is did not include the caution contained in the PSTGs to ' place the applicable residual heat removal service water (RHRSW) pumps in service as soon as possible' after initiating low pressure coolant injection.
This action was more important at BFN than at some other BWRs because there was no bypass around the residual heat removal heat exchangers."
TVA's Response This caution has been addressed in step RC/L-1 in E01-1.
3.
NRC's Concern
" Step RC/L-2.1 in the E0Is reads, ' Maintain level above -150 inches (+20 inches on LI-3-52 and 62A).'
The same step in the PSTGs read 0 inches rather than +20 inches."
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1 TVA's Response The purpose of this step vias to establish an alternate control band for the reactor water level if the water level could not be maintained in the normal control band.
The bottom of this alternate control band is the top of the active fuel (0 inches on L1-3-52 and 62A). Therefore, the PSTGs call out a level of 0 inches as the bottom of that alternate control band.
However, the E0Is call out a level of -150 inches (+20 inches on the fuel zone instruments) because this level is more easily readable and adds a small margin of conservatism into that alternate control band.
This is consistent with the philosophy used throughout the E0Is.
f 4.
NRC's Concern "In addition to this numerical difference between the PSTGs and the E0Is, this step, which was repeated in several places throughout the E0Is, should be revised to include the numbers for both level instruments that were referred to (i.e., LI-3-58A and B for the -150 inch reading; see item 7 below)."
TVA's Response The E0Is have been reviewed and instrument numbers added to applicable sections where -150 inches is referenced.
5.
NRC's Concern "The table of alternate injection subsystems that accompanies step RC/L-3 of the E01s indicated a discharge head of 150 to 0 psig for the RCIC and HPCI systems when being supplied with steam from the auxiliary boiler.
The PSTGs indicated 350 to 0 psig."
TVA's Response The E0Is were revised to correct this discrepancy.
Both documents, PSTGs and E01s, now specify a range from 350 to O psig.
6.
NRC's Concern "The PSTGs indicated a value of 931.3 psig in step RC/P-1 as the pressure limit after safety relief valves (SRVs) were manually opened.
While a value of 931.3 psig would be impossible to read on the installed instruments, a value of 930 psig would be readable.
The E01s contained the generic value of 950 psig."
TVA's Response The value 950 psig has been changed to 930 psig in E01-1.
The value of 930 psig is the minimum pressure at which turbine bypass valves are fully open.
Pressure reduction below this would close bypass valves and increase flow thru the SRVs to the suppression pool.
This would be unnecessary heat additicn to containment.
If the turbine bypass valves were unavailable, 930 psig would provide an adequate operating margin to the SRV setpoint.
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.)
1 7.
NRC's Concern "Several steps in the E01s referred to the action to be taken if the control air supply to SRVs was lost.
An example was step RC/P-2.la which read, 'If the continuous SRV air supply is or becomes unavailable, place the control switch for each SRV in the CLOSE or AUTO position.'
The PSTG step used the words 'DW Control Air and Control Air.'
In addition to revising the E01s to reflect the current PSTGs, the licensee should add an additional phrase indicating that the containment atmosphere dilution system nitrogen can now be cross-connected to drywell control air."
TVA's Response The PSTGs and the E01s now agrec on the terminology to be used in these steps. They both reference the use of all three sources of pneumatics for operation of the SRVs: Control Air, Drywell Control Air, and Containment
. Atmosphere Dilution.
8.
NRC's Concern "The E0Is contained several references to the Yarway-type level instruments in the control room (2-LI-3-58A and 2-LI-3-58B).
The following excerpts from the E0Is all referred to these instruments:
- Page 1 of 2 in Contingency C2, ' Emergency Rx Depressurization'
- Heated reference leg instruments [ red labels] are
- not reliable during rapid Rx depressurization below *
- 500 psig.
For these conditions, use cold reference *
- leg instruments [ green labels] to monitor Rx water
- level.
- The table on page 2 of 6 in step RC/L referred to these instruments as Emergency Systems Range
(-155 to +60 in.)
- In other parts of the E01s (e.g., steps RC/L-2.3, Cl-7.3, and C5-2a) the use of the words '...Rx water level above -150 in.' assumed that the operator understood that these Yarway instruments were the ones to observe.
- The actual labels for these instruments in the unit 2 control room were black on gray and read as follows:
2-LI-3-58A Reactor Water Level A ACDT RANGE l
L
t The labels on these instruments in the simu ator were black on yellow and read:
2-LI-3-58A Reactor Water LEVEL A The BFN staff should coordinate the control room labeling project, maintenance of simulator fidelity activities, and nomenclature in the procedures to arrive at a consistent nomenclature and set of labels for these instruments."
TVA's Response The caution concerning heated reference leg instruments has been deleted due to the modification which moved the reference legs outside primary containment.
. Instrument numbers have been added to RC/L page 2 of 6, RC/L-2.3, Cl-7.3 and C5-2A.
The discrepancy noted between the tags in the simulator and the unit 2 control room will be resolved by the relabeling effort being done as a resolution of a Detailed Control Room Design Review (DCRDR) finding. As part of this program, duplicate tags, labels, etc. are being produced to relabel the unit 2 control room and the simulator concurrently. This tagging / labeling effort will be completed in accordance with the schedule for resolution of DCRDR which has previously been communicated to NRC.
9.
NRC's Concern "In Contingency C2, ' Emergency Reactor Depressurization,' step C2-1.3 directed the operator to ' defeat isolation interlocks if necessary, using one or more of the systems on the facing page.'
On the facing page (page 2 of 2), the references had the correct information, but were listed in a sequence that differed from that on the system list provided earlier."
TVA' Response The system lists for appropriate steps were revised to reflect the same sequences in E01-1.
10.
NRC's Concern "The diagram on page 4 of E01-1, which showed the organization of all the E01s and supporting contingency procedures, did not include the procedures in E01-3, ' Secondary Containment Control and Radioactivity Release Control."
l TVA's Response l
This diagram in E01-1 has been revised to include the organization of E0I-3.
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Section 3,1,2 - Comparison of Revision 2 of BfN Writers Guide (WG) with NUREG-0899
- 11. NRC's Concern "The WG (Section 4.2.8 page 11) did not differentiate between the exclusive and the inclusive word '0R'.
The exclusive '0R' is equivalent to 'A or B' but not both."
TVA's Response The exclusive "0R" is not used in the E0Is. All cases of the use of the word "0R" are the inclusive case.
12.
NRC's Concern "Using the word 'THEN' at the end of an action to instruct operators to perform another action within the same step runs. actions together (e.g.,
'Open the valve THEN close the breaker').
Actions that were embedded in this manner could create several problems:
embedded actions may be overlooked, may be confused with logic statements, and may be more difficult to verify with the single checkoff."
TVA's Response The HG has been revised to specify that action statements not be run together.
The.E01s were reviewed for consistency with this guidance.
Inconsistencies were found in Contingency C5, page 6, and Contingency C4, page 2.
These were revised in E01-1.
Section 3,1,2,1 - Guidelines for Logic Diagrams and flowcharts
- 13. NRC's Concern
" Item 6 of Section 4.3.8, page 11, of the WG stated, 'If multiple uses of logic terms are required to describe a condition or action, a logic diagram should be used rather than a conditional statement.'
An example of where this was not followed was on page 3 of step RC/P-2.1."
TVA's Response The HG has been revised to more clearly define the conditions under which a logic diagram will be used as opposed to logic terms.
The E01s were reviewed for consistency with this guidance.
No changes were made to the E0Is.
- 14. NRC's Concern "The WG did not discuss capitalization in logic diagrams.
Text should be written in both capital and small letters for two reasons:
(1) if all words are capitalized, then capitalization cannot be used for emphasis, and (2) text written in all capital letters was more difficult to read than one written in both capital and small letters.
An example was step RC/L-2.3."
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1 TVA's Response The WG has been revised to include a discussion on the use of capital and lower case letters in the logic diagrams.
The E0Is were reviewed for consistency with this guidance.
RC/L-2.3 is not a logic diagram and contains both capital and lower case letters.
No changes were made to the E0Is.
- 15. NRC's Concern "The WG did not state that notes and cautions in logic diagrams would be placed in the flow path directly before the steps to which they apply."
TVA's Response The WG has been revised to state that notes and cautions in logic diagrams will be placed in the flow path directly before the applicable steps. The E01s were reviewed for consistency with this guidance.
No changes to the E0Is were necessary.
- 16. NRC's Concern "The WG did not specify the size of type that will be used in logic diagrams."
TVA's Response The specific type size used in the E01s is not currently specified in the WG nor is it a practice at BFN to specify minimum type size in any site instructions.
It is the responsibility of the procedure writer to ensure the readability of the procedure.
In addition, as part of the verification process, procedure readability is checked.
Section 3.1.2.2 - Referencing
- 17. NRC's Concerr.
j "Section 4.4.1.3, page 12, of the WG discussed referencing, but did not distinguish between a reference that directed operators to return to the original procedure and one that did not."
TVA's Response TVA does not consider there to be a need to distinguish between the different l
types of reference statements.
The referencing is done correctly and there is no confusion as to where the operators should proceed for the next actions to be taken.
In some cases it is not possible to determine where the operators will be directed upon execution of the procedure they are currently following. A change in the operating parameters may dictate a change in the sequence of execution of the procedures, and this change may or may not be apparent at the time the operators are referenced to the new procedure or section. The only information that should be included with the reference phrase is to specify whether or not the current procedure should be exited or should be executed in parallel with the referenced procedure.
These statements are included in the existing E0Is.
This aspect of E01 usage is developed during simulator and classroom training.
l L------- -------- -------------------- -
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18.
NRC's Concern "The WG did not discuss methods, such as the use of tabs for easily identifying the sections and subsections of the E0Is."
TVA's Response Tabs are currently used in the E01 books to identify the different sections af the E0Is, although this guidance is not specifically addressed in the WG.
Tabbing is done to facilitate operator friendly procedures.
The current method is adequate.
Section 3.1,2,3 - Vocabulary and Syntax 19.
NRC's Concern "Section 4.3.3, page 9, of the WG stated that short, simple words should be used in the E0Is.
For this reason, the verb ' commence' in Appendix A should be replaced with 'begin'."
TVA's Response The statement in the WG has been revised.
The word "short" was replaced with
" easily understandable", thereby making it unnecessary to change words with which the operators are familiar.
20.
NRC's Concern "Only action verbs listed in Appendix A of the WG should be used.
These verbs should be used with consistency in the E01s.
For example, the words ' initiate' and ' start' should not be used interchangeably."
TVA's Response These verbs are not used interchangeably.
The list of action verbs included at the back of the WG lists both of these verbs.
Each of these words has a distinct and different definition.
Section3.1.2.4-formattingofActionSteps 21.
NRC's Concern "Section 4.4.1, pages 12 and 13, of the WG did not define and provide the format of the following types of action steps:
consequential steps, equally acceptable steps, recurrent steps, time-dependent steps, and diagnostic steps."
TVA's Response Consequential steps are written as override steps in the E01s.
These are formatted as "IF while executing..., THEN..." steps.
The WG has been revised to incorporate this format.
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.The following types of steps are not utilized in the E01s:
Equally. acceptable' steps l
Recurrent steps l
Time dependent steps-
. Instead of ~ diagnostic steps, the E01s use logic steps ("IF... THEN..
".. and-
"WHEN....THEN...") as methods of determining the. appropriate steps to be taken. The format for..these types of steps is discussed in section 4.3.8 of
. the WG.
~
'22.
NRC's Concern "Section 4.2, page 6, of the WG discussed major steps and substeps, but the WG did not discuss _the relationship between major steps and substeps. Major steps should be used as headings that summarize the actions in the. associated substeps. Only substeps should contain specific operator actions."
TVA's Response The WG has been revised to discuss using major steps as headings summarizing the specific actions in the associated substeps and that only the substeps
- should contain specific operator actions. - The E0Is were reviewed for-consistency with this guidance.
No changes to the E0Is.were necessary.
- 23. NRC's Concern
" Item 6 of Section 4.2.3, page 6, of the WG stated that bo'th override steps and ' steps containing an unexpected action' shall be enclosed in boxes.
If.these two types of steps were formatted identically, operators could mistake a step containing an unexpected action for an override step.
Furthermore, because override steps must be remembered after they have initially been read, they should be formatted in a unique manner."
TVA's Response The WG has been revised and no E01 corrections were required.
Section3,1,2,5-FiguresandTables
- 24. NRC's Concern "The WG did not identify how figures would be labeled."
TVA's Response Section 4.2.5.2 of the WG specifies how figures are to be labeled.
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- 25. NRC's Concern "The HG did not state that all figures would be reproduced from originals and would be of quality equal to that of the originals.
A review of the E01s showed that the reproduction of figures was poor and, in most cases, the figures were small, making them difficult to use especially under reduced lighting conditions."
TVA's Response The WG now states:
Figures will be of a one-half page s1ze (or as close as possible to one-half page size where restrictions prevent one-half page size figures from being used) and of quality equal to that of the originals.
The E0Is now reflect this guidance.
Section 3.1,2,6 - Iden.tification and Location 26.
NRC's Concern "Section 4.3.5, page 10, of the WG indicted that information on equipment location should be provided, but did not address the format to be used when presenting this information."
TVA's Response The WG states: At the beginning of each appendix shall be placed a list of any tools or other hardware needed to execute the steps in that appendix and the storage location of those required tools or other hardware.
In addition, the WG section 4.3.5 has been revised to address equipment location format.
The E0Is reflect this guidance.
27.
NRC's Concern "Section 4.3.5, page 10, of the WG discussed component identification codes but did not discuss the format to be used when including these codes in procedures."
TVA's Response The WG has been revised to include this information.
The E0Is were reviewed for consistency with this information and required changes were made in E01-1.
Section3,1.2,7-OrganizationofE0ls
- 28. NRC's Concern "Section 4.1.1, page 3, of the WG discussed the information to be included on the E01 cover sheet, but did not include a requirement for the revision number and date."
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TVA's Response All site procedures have the revision number and date on the cover sheet.
In addition, current BFN guidance calls for the revision number to be stamped on the top of every page of the procedure and this requirement is reflected in the WG Section 4.2.1.
- 29. NRC's Concern "Section 4.1.3, page 3, of the WG discussed entry conditions, but did not specify the format to be used when presenting entry conditions.
The WG should specify this format and provide an example."
- TVA's Response Information regarding the format of the entry conditions for the E01s has been incorporated into the WG revision, and an example provided.
The E0Is were reviewed for consistency with this guidance.
No changes to the E01s are necessary.
Section3.1.3-ComparisonofE0IswithRevision2ofBFNWritersGuide
- 30. NRC's Concern
" Logic terms were used incorrectly throughout the E0Is.
The terms 'IF' and 'WHEN' were frequently used without using 'THEN'.
The terms 'AND' and '0R' often were not emphasized correctly.
Terms such as 'before' and
'except' were emphasized and used as logic terms."
TVA's Response The E0Is were reviewed to identify inconsistencies in the use of logic terms and emphasized terms. Changes found during this review have been made to the E0Is.
31.
NRC's Concern
" Item 6 of Section 4.3.8 of the WG stated that in those cases where multiple uses of logic terms were required, a logic diagram should be used rather than a conditional statement.
Logic diagrams were used only once in the E01s.
Use of logic diagrams would improve the usability of the E01s."
TVA' Response The WG has been revised to redefine " multiple uses of logic terms" to clarify the conditions under which a logic diagram could be used as opposed to logic terms.
The E0Is are usable as is and do not require revision.
- 32. NRC's Concern I
l "The figures and graphs throughout the E01s were difficult to use because 1
of their small size and the poor quality of reproduction."
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.TVA's Response j
'l
'See TVA's Response to NRC Concern No. 25.
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- 33. NRC's Concern
" Terms in the E0Is such as 'are', 'stop', and 'can' were not emphasized con s i s te ntly.. "
HA's Response The EDIs were reviewed for emphasis on the identified terms.
Certain words were emphasized in places to alert the operator to specific conditions, however in other places the same words do not require emphasis.
- 34. NRC's Concern "The cautions in the E0Is were not numbered.
For-example, the full-page caution on page 2 of Step RC/L h.ad no number or title."
TVA's Response It is not necessary to number or title the cautions in the E0Is.
There are enough cautions in the E01s that numbering or titling them would add extraneous information to the procedures and could confuse the operators.
The PSTGs number the general cautions but only because the general cautions are numbered in the Boiling Water Reactor Owners' Group (BWROG) Emergency Procedure Guidelines (EPGs) and TVA chose not to delete the numbering scheme utilized in the generic EPGs.
In the E0Is a distinction between the general and specific cautions is not made and, therefore, these cautions are not numbered.
35.
NRC's Concern
" Figures in the E0Is were sometimes not titled correctly.
For example, Figure D defined conditions under which emergency reactor depressurization was required.
The figure was entered with two variables, pressure and water level, but was entitled ' Pressure Suppression Pressure'. Another example was Figure F, ' Suppression Pool / Temperature'."
TVA's Response The figurts are titled in accordance with the BWROG EPG titles.
Their usage in the plant specific E0Is including the entry conditions is consistent with the industry.
Section3,1,4-ReviewofE01 Calculations
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36.
NRC's-Concern "The Boiling Hater Reactors Owners Group emergency procedure guidelinesL
]
(BWROG EPGs) included'a number of plant-specific limits. setpoints, and action-levels that required calculation of plant-unique values.
Appendix C of the EPGs provided detailed for developing input data and performing Lthese' calculations.
L The team reviewed a sample;of the input data development and final l
calculation and verification documentation fortthe final Appendix C L
calculations' listed below. The team verified the correlation of input l
data from EPG Table CI-T4, performed checking calculations, and confirmed to'the extent possible that the calculations had been' performed in accordance'with the_EPG Appendix C procedures.
The following calculations were reviewed:
low pressure coolant inject 1'on and core spray net positive suction head drywell spray. initiation pressure limit pressure suppression pressure-primary containment design pressure primary containment pressure limit
'drywell spray flow rate suppression pool cooling spray initiation pressure.
The team noted that these calculations included input assumptions, bases, and-the identification of the performer.
The team alsoLfound that all-but four of the calculation packages had received a documented independent-review.
These four calculation packages had been identified by the licensee as requiring further revision and/or review for various reasons and were expected to be completed tiefore plant restart.
The calculations had been performed, reviewed, and approved by the Nuclear Production organization, not the Division'of Nuclear Engineering. Although the team considered the performance and review of the calculations technically acceptable, the licensee had no procedure available for the control of the calculation activities.
This is discussed further in Section 3.1.6."
'TVA's Response Plant Managers Instruction (PMI) 12.6, Implementation and Maintenance of E01s, addresses control of calculations.
Form PMI-88 documents the calculation and review process.
Appendix C,-which includes the four calculations mentioned above, is presently being incorporated into a program manual which will fall under plant QA review.
This program manual will be in place prior to unit 2 restart.
S& tion 3.1.5-OngoingEvaluationofE01s
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~37.
NRC's Concern "Section 6.2.3 of;NUREG-0899, ' Guidelines for the Preparation'of Emergency Operating. Procedures,' recommends that. licensees establish a-program forLthe ongoing evaluation of the E01s.
The licensee established an ongoing evaluation program by implementing PMI-12.6, ' Implementation-and Maintenance of E01s.'
Section 4.8 of this-instruction described the
- dynamic, ongoing E01 evaluation process.
This section also delineated a method for anyone using or interfacing with the E01s.to make known their concerns or comments by filling out an E0I evaluation sheet, which was provided as' Attachment 2'to this procedure.
- However, interviews with licensee operations personnel' indicated that they were'not familiar with the program for sugg6 sting improvements to the E01s.
None of those interviewed had filled out an E01 evaluation sheet."
.TVA's Response-
~PMI-12.6 has been included in the operator required reading program.
Operators are encouraged to use these forms in PMI-12.6 to make comm.ents on the E0Is..
Section3,l',6-QualityAssurance(QA)forPSIGs 38.
NRC's Concern.
"NUREG-0899, Section.4.4, ' Quality Assurance,' states that the PSTGs
-shculd be subject to examination under the BFN overall QA program to ensure that they are accurate and up to date.
The existing PSTGs and associated appendices and calculations were not controlled under the licensee's QA program, although the E0Is themselves were.
Site Director Standard Practice (SDSP)-2.11, ' Implementation and Change of Site Procedures and Instructions,' Revision 8,'provided.the primary controls for plant procedures and applied to the E01s.
In addition, the licensee had implemented Plant Managers Instruction (PMI) 12.6, ' Implementation and Maintenance of E01s,' Revision 2, to add specific features required to ensure that PSTG revisions, verification and validation requirements, and training are implemented for E01 revisions.
These procedures were generally acceptable for control of the E0Is.
However, the licensee had not implemented procedural controls for the preparation, review, and approval of the procedures generation package (PGP),=PSTGs and PSTG Appendices A through D.
_These documents had not been issued or administered as controlled documents in accordance with the TVA QA Program.
The TVA Operations Support Group E0P Coordinator l
advised the team that current plans included creation of an 'E01 Program Manual', which would' include the PGP, PSTGs, and appendices and would be subject to either the controls of SDSP 2.11 and PMI 12.6 or other equivalent controls.
The licensee further stated that a calculation control procedure would also be developed and applied to control the r
calculations of PSTG' Appendix C."
l
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TVA's Response An E01 Program Manual which will include:the PSTGs, PSTG appendices and deviations cross reference. document will be incorporated into the existing
~
plant document control system. As such, this manual will require QA' review,.
technica1' reviews as required,1affected section review, screening reviews for unreviewed safety question determinations, and other controls imposed under the existing. system.
This manual will be implemented prior to restart'with all the.above mentioned documents.
Section3,2-ContainmentVenting
- 39. NRC's Concern "E0I-2, Step PC/P-1 required initial venting of the containment in accordance E0I Appendix.13, ' Venting Primary Containment,' when drywell-pressure reached 2.45 psig.
The first vent path used'was 2-inch piping from the suppression chamber.followed by venting through 2-inch piping, from the drywell.
If venting through these paths was unsuccessful, step PC/P-5 required emergency venting through the.large-bore (20- to 24-inch) flow path in accor~ dance with E0I Appendix 17, ' Containment Venting Bypassing Interlocks,' if drywell pressure exceeded 55 psig.
This~ pressure'was chosen on the basis of maintaining operability of the safety relief valves. -The initial flow path selected was from the torus with a second flow path available from the drywell The licensee had reviewed the design of the vent pathway and. standby gas treatment.(SBGT) system piping and ductwork for the large-bore ~ flow paths and had concluded that expansion joints in the high-pressure and low-pressure transition in the duct immediately downstream of the outboard containment isolation valves would fail, resulting in venting of the containment to the reactor building.
The licensee viewed this as a desirable consequence in that it would prevent overpressurization damage to the SGTS system and keep it available to process the reactor building atmosphere for elevated release.
However, no specific analyses or tests had been performed to evaluate building pressurization, ground-level radioactivity release, and similar considerations."
l TVA's Response TVA will perform an evaluation on reactor building pressurization and ground
)
level' release prior to unit 2 restart.
40.
NRC's Concern i
" Additionally, the licensee had not evaluated the operability of the containment vent path isolation valves under postaccident differential pressure (damper 64-36, valves FCV-64-32, 64-29, 84-19, 84-20, and others as listed in E0I Appendices 13 and 17).
At the time of the inspection, the licensee was gathering the information necessary to determine whether the valve actuators, torque switch settings (as applicable), etc., could open and reclose the valves under expected containment venting and flow differential pressure conditions."
m
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TVAs Response-TVA considers this evaluation to be beyond the design basis of the plant.
However, TVA wil1 perform an evaluation of the containment vent path valve-operability during post accident differential pressure conditions. Once the evaluations are complete,.the necessary changes will be made to the PSTGs and
-the E0Is prior to unit 2 restart.
. 41.
NRC's Concern "Similarly, the E0Is did not consider local / manual venting under blackout conditions or loss of offsite power.
The licensee stated that no major changes in methodology or approach were planned until NRC approved revision'4 of the BWROG EPGs".
TVA's Response-Revision 4 of the.BWROG EPGs will be implemented during 1990. An evaluation of local / manual venting under blackout conditions or loss of offsite power will be performed in conjunction with implementation of EPG revision 4.
Sution3,3-PlanthikdownofE0Is 42.
NRC's Concern
" Provisions for remote in-plant communications were inadequate to. support implementation of the E0I appendices.
Most of the E01 appendices required manipulation of valves and equipment in the reactor building and auxiliary instrument rooms.
The operations personnel involved in the team's E01 walkdowns stated that these operations would most likely be directed by phene or radio from the control room.
For example,'E01 Appendix 10, ' Locally Venting. Control Rod Drive Withdrawal Lines',
required entry into the reactor building and into a contamination area and involved extensive valve manipulation on a possible combination of 185 control rods.
Discussions with plant operators indicated that the hand-held radio system was not effective for communications in all areas of the plant.
Although dial phones were located throughout the plant, their availability was not ensured during loss-of-power events.
Similarly, sound-powered phone jacks were located throughout the plant, but neither handsets nor headsets appeared to be readily available and the licensee was unable to tell the inspection team whether the systems would remain operational during loss-of-power events.
The operations superintendent stated thLt a new in-plant radio system with repeater and hand-held units was being acquired and installed; this should address the above concerns."
1 J
TVA's Response q
1 The BFN in-plant communications system consists of a combination of systems
{
which together provides adequate coverage for all areas of the plant requiring-i effective communications in order to carry out the actions specified in the E0Is.
In addition, a new in-plant radio system is being designed that calls for installation of additional cable, relocating one of the existing repeaters, and installation of a dedicated uninterruptable power supply.
TVA considers the existing communications systems adequate for performing the actions called for in the E0I appendices.
However, with the changes described above to the radio system, improved radio reception in many areas of the plant and added flexibility in execution of these actions is expected.
43.
NRC's Concern "The E0Is and appendices did not provide any information on valve location and frequently did not include the name of the valve.
The operators had difficulty in finding about 20-percent of the valves to be manually operated and required assistance from other operators or the control room.
Examples included the following:
valves85-614 (control rod drive vent valve for each hydraulic control unit) valves 74-681A and B (condensate transfer to RHR) valves74-622 and 74-624 (drain pump discharge to loop crosstie line), which were located above the torus.
To minimize unnecessary delays and congestion of.the communications system during an accident, the licensee should consider providing information on the valve locations in the procedures or in a place that is convenient for the operators.
Seven manual valves that had to be operated in step 1 of Appendix 7, Subsection g, were all located in areas of the core spray room that were very difficult to access.
Electrically and mechanically, this alternate injectiot lineup was probably the most difficult to perform of those listed in Appendix 7.
Either modifications should be made or this lineup should be noted as being the alternate injection option of last resort."
i TVA's Response This problem has been corrected. Manual valve locations have been inserted in E0I-l where required.
The alternate injection options, Appendix 7, Subsection g has been identified in the procedure as the least desirable option.
This change has been l
incorporated.
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- 44. NRC's Concern i
"Several:of the valves required to be locally operated by E0I Appendix 7,
' Alternate Injection Subsystems' were inaccessible from the reactor.
building floor or existing gratings.
Examples included'the following:
Appendix 7, Subsection a - valves 75-582A (condensate transfer to loop I core spray), 2-713, 2-712-(condensate transfer to RHR).
No prestaged ladders or reach rods, etc., were available.
Appendix 7, Subsection h:- valve 12-777 (steam supply.to Unit 2.
reactor core isolation cooling system) was located about 20 feet
.above floor level. Same as above.
Appendix 7, Subsection j'- valve 12-778 (steam supply to Unit 2 high pressure coolant injection system) was located about 15 feet above floor level.
Same as above.
The licensee stated that: access provisions had not.been evaluated as part of the procedure validation process but would be reviewed."
TVA's Response' Ladders have been staged for access to the valves. Access provisions have been evaluated and found to be adequate.
The validation procedures now address this concern.
- 45. NRC's Concern "The temperature instruments (II-80-34-6 and TI-80-34-7 and B).specified in the Caution.1 table (page 2 of 6, Step RC/L-1) for determining temperature in the vicinity of the various level instrument reference-legswerephysicallylocatedonthebackpanelsofthecontrlolroom.
Both in the control room and in the simulator, operators used the
~ temperature 1nstruments available on the front panels to make these determinations.
The licensee should consider methods of correcting this situation, such as moving the instruments to the front panels or performing a calculation that would show' conservative equivalency between the back and front-panel instruments."
TVA's Response A modification has been implemented which reroutes the reference legs outside primary containment for the nornal, emergency and post accident flooding instruments.
The caution no longer applies to these instruments and the E01s have been revised accordingly.
The caution now applies only to LI-3-55 (shutdown vessel flooding range).
- 46..NRC's-Concern l
"As part'of the detailed control room design review, the licensee has l
Installed phenolic placards on the glass faces for the above containment temperature indicators and recorders on the rear of main control board panel 9-4.
The placards listed the indicated or recorded data points but obscured most of the charts or dials, preventing the operators from observing recorded trend information without opening the instrument door.
The licensee stated that the placard installations would be reevaluated."
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TVA's Response The DCRDR program did not install-the phenolic placards on the glass faces-for the temperature indicators and recorders.
These were installed under the operator aid program.
However:the location oflthe placards was evaluated in the DCRDR program and, in_some cases, will be moved to facilitate monitoring of trends on those recorders.
This' portion of the DCRDR effort will be completed within the schedule previously communicated'to NRC.
- 47. NRC's Concern s
"E01-3 included a. table of secondary containment area temperature instruments and the_ maximum normal ~ operating values for each area (e.g.,
high pressure coolant injection (HPCI) room) representing an E01 entry condition.
The licensee stated that the maximum ~ normal operating values listed for each instrument corresponded to the alarm setpoints for the individual channels and had been verified and validated.
However, a sample of five of the alarm setpoints were checked via direct readout at panel'9-21, and the following four setpoints were found to exceed the E0I entry condition value in the above table:
HPCI room, El. 519 - 185 *F as found rather than the 175 *F entry condition northwest corner room, El 519 - 182 *F as found rather than the 175 *F entry condition southwest' corner room, El. 519 - 168 *F as found rather than the -
160 *F entry condition torus area, El. 519 - 180 *F as found rather than the 175 *F entry condition Since the indicators for these instruments only alarmed in the control room and were.only readable, one at a time, from a common monitoring panel room behind the main control board, the erroneously high setpoints could cause late recognit'ien of the fact that the entry condition had been' reached.
The licensee subsequently stated that the indicators were verified to be within the cumulative instrument channel maximum tolerances and were consistent with EPG and PSTG guidance, but that several of the setpoints had been recently changed and not yet incorporated into the E01s.
The licensee could not tell the team why the setpoints hed been changed without determining the effect on the E0Is."
TVA's Response The setpoints in question have all been reviewed subsequent to the E0I inspection and'all were found to be correct.
There were some minor
. differences between the entry conditions stated in the procedure and the alarm setpoint indicated at the local panel.
However, these differences were within the maximum tolerances allowed for these instrument loops.
The entry conditions listed in the MAX NORMAL OPERATING VALUE columns of the Secondary Containment Parameters Table are intended to coincide with the alarm setpoints for those instruments.
The specific value listtd in the table has no significance other tnan the fact that it does correspond-t.o the alarm setpoint.
Whether or not the setpoint has drifted a few degrees high or low is not critical to performing the procedure, as long as the setpoint is still within the allowable tolerances. The entry condition based on area temperatures is, therefore, not dependent on exceeding the exact value listed in the table, but rather on exceeding the value that causes the temperature switch to close and energize the annunciator.
The table in E0I-3 has been revised to indicate that the annunciator is the mechanism that is to be used for determining entry into the procedure.
This does not and will not imply that the panel should not be checked to verify the high temperature condition.
48.
NRC's Concern "During walkdown of E01 Appendix 12, ' Alternate Depressurization/ Pressure Control,' the HPCI turbine steam supply mimic on the vertical portion of panel 9-3 incorrectly showed the HPCI turbine stop valve (FCV-73-18) as being downstream of the HPCI turbine control valve (FCV-73-19).
The licensee stated that the mimic would be reviewed and corrected if required."
TVA's Response The mimic has been corrected to reflect actual plant configuration.
49.
NRC's Concern "The low-range drywell pressure instrument (PI-64-67B) and associated recorder (R-64-50) were scaled for absolute pressure in a range of 0-80 psia while the wide-range drywell pressure instrument (PI-64-160) was scaled for gauge pressure in a range of 0-300 psig.
These instruments were variously used for determining E0I entry conditions, E01 limits, initiation of containment venting, etc.
Some E01 steps and curves (e.g.,
entry conditions specified in Figure C, 'Drywell Spray Initiation,' of F01-2.1) only provided gauge values, others (e.g., E01-2, SP/L, PC/P-5, 1
Appendix 17) provided both gauge and absolute values.
1 Although the operators have been trained on the differences in instruments, this mixed convention provided the potential for misreading or misapplying the data required for the E01s.
1 The licensee stated that the use of instruments that give measurements in i
variant terms had been identified during the detailed control room design l
review and was the subject of Human Engineering Discrepancy (HED) 0201.
The HED was previously scheduled for completion before restart and would result in rescaling of the low-range instruments to units of psig.
During the inspection, the licensee informed the team that the HED would likely l
be reprioritized for completion during the next refueling outage instead I
of before restart because of problems with obtaining materials for the required modification.
The licensee should ensure, through operator training and procedural nomenclature, that this engineering unit difference does not cause confusion."
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l TVA's Response 3
i This modification involves replacing the psia pressure transmitters with psig l
units and converting the control room indicators to psig reading instruments.
- i When this is complete, the drywell pressure instruments will indicate in units of psig. A review of the existing E0Is and training material has been done to ensure that all operators understand the difference between the two types of instruments and how each is used in relation to the other.
Training stresses the use of the different pressure instruments and no changes were determined to be necessary at this time.
This effort will be completed within the DCRDR schedule.
l
- 50. NRC's Concerns "E0I-2.1.2 specified 160 *F drywell temperature as an entry condition for drywell/ temperature (DW/T).
The initial actions of Step DW/T-1 required only maximizing drywell cooling. Abnormal Operating Instruction (AOI) 2-A01-64-6, 'High Drywell Temperature', Revision 0, Step 4.1.4, required that the reactor be manually scrammed and E0I-2 entered if drywell temperature reached 160 *F.
The licensee was unable to provide the basis for the AOI requiring a reactor scram and stated that the E01 and AOI would be reviewed and revised as necessary to ensure consistency and a valid basis for the prescribed A0I actions."
TVA's Response An evaluation will be performed to determine the basis for the 160* drywell temperature scram prior to unit 2 restart.
- 51. NRC's Concern "The words ' Continuous SRV Air Supply' were interpreted by the operators as referring to annunciator 2-PA-3-70 on panel 2-XA-55-3E in carrying out the actions in step RC/P-2.1. An alarm on this annunciator would be sufficient to prompt action to place safety relief valves in 'Close' or
' Auto'.
This step should be modified to allow use of control air and/or containment atmosphere dilution system nitrogen."
TVA's Response E01-1 and the associated PSTG now contain specific references to all three tystems that can be used to supply air to the safety relief valves.
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52.
NRC's Concern "The key number specified in step RC/Q-5.4 for gaining access to panel 9-16 (individual scram switches) in the auxiliary instrument room had been changed from number 192 (E0I) to number 85 (key locker).
Finding the proper key in the control room key locker required the operators to conduct a search of the key list for the name of the panel (which was not provided in the procedure) in order to identify the new key storage s
location.
This problem was the same for the other keys used ir. the walkdown of Appendix 9 (key numbers 167 and 168, panel 9-27).
+:
-- V Either the key locker storage locations should be changed to match the as-written E0Is or the E01s changed to match the key locker storage locations.
The E0Is should also include the noun name of the key.
A long-term method should be developed to ensure that key locations are not changed until a corresponding change to dependent procedures has been completed.
In addition to the key-numbering problem, the team noted that the operators had difficulty with the key sticking in the panel door lock-mechanisms and in turning the key.
The licensee should lubricate or repair these locks so that a key will not break under the stressful conditions of an emergency."
~
TVA's Response TVA has corrected the inconsistency in key numbers between the procedure and the key locker. The noun name of the lock location has been incorporated in E01-1.
A Maintenance Request was written to have the locks on the panels for scram test switch (panel 9-16) and rod sequence control system (panel 9-27) inspected and either lubricated, or replaced.
This work has been completed and the locks now operate freely.
53.
NRC's Concern "A general problem in the Unit 2 control room stemmed from the removal of nearly all operator aids from the control boards. Aids that the operators expressed a strong desire to have reinstalled included the following:
A general arrangement drawing of the suppression pool, showing the location of safety relief valve (SRV) discharges and temperature detectors, should be posted near SRV controls and torus temperature instruments.
A drawing showing the general arrangement and location of thermocouple in the drywell should be posted near temperature instrument TI-80-34.
The suppression pool heat capacity and temperature limit curve should be posted.
The checklist used for keeping track of the status of system interlock bypasses and entry conditions should be posted in the control room.
This checklist was posted in the simulator, but several of the bypasses that could be required during the use of the E01s were missing.
The licensee should consider adding the following bypasses to this checklist:
-- reactor core isolation cooling (RCIC) test mode isolation bypass
-- RCIC high reactor water level trip bypass
-- RHR injection valve timers bypass
-- containment venting bypass The licensee stated that a plan existed for replacing the operator aids and that the above specific recommendations would be evaluated."
~
4 h
TVA's Response The operator aid program has been changed to upgrade the requirements for maintaining operator aids.
PMI 12.12. " Conduct of Operations," has been revised to specify all of the programmatic controls that are in place to control operator aids. Many of the operator aids that were removed from the control room have been replaced, subject to the new requirements of PMI 12.12. The specific recommendations made by the operators were evaluated and operator aids placed in the control room, as necessary.
In addition, a review of the simulator and the control room was performed and appropriate changes made to ensure that appropriate operator aids posted in the simulator were also posted in the control room.
54.
NRC's Concern "The directions in the E01 appendices were internally inconsistent in format.
Examples of these inconsistencies were the following:
- Valve identification numbers (e.g., FCV-1-14) were used in
-j directions to close or verify closed valves. A different number (e.g., 2-HS-1-14a) was used to identify the corresponding valve control switch handle that must be operated in the control room.
The name of the valve was not used in the procedure.
- Only the pump name (e.g., CS Pump A) and not the component or control switch identification number was used in other sections of the procedure directing the verification of run status or the starting of pumps.
- Elsewhere, the appendices provided directions in short-text form (e.g., ' Verify HPCI Running').
Directions in an appendix and throughout the E0Is should be made more internally consistent (e.g., both the name and the associated control device number should be used when providing directions for a verification or action step involving the operation of a component)."
TVA's Response A revision to E01-1 has been made to the appendices to address nomenclature of valves called for in the E01s.
Major components are more readily identified by noun name than unique identification number.
Specifying handswitch numbers or controller numbers for pumps or turbines would not be as operator friendly.
No changes are warranted for the last two examples.
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g.a 55[
NRC's Concern 1
i
" Appendix 7, subsection g, required jumpers to be installed for the pressure suppression chamber head tank pumps in reactor motor-operated valve boards 2C and 28.
These jumpers had to be attached to terminals that were very' deep within the compartments.
Emergency installation of these jumpers would be both difficult and dangerous. Alligator-clip-type jumpers would not attach securely to the terminals, and loosening the lugs for the installation of spade-type jumpers would be dangerous ana difficult. The licensee shculd consider modifications to facilitate easier and safer installation of these jumpers.
Some jumpers required to be installed by the appendices on terminal strips had to be connected to terminals with insulated shanks or to terminals that did not protrude far enough from the terminal strip to allow the reliable connection of alligator-clip-type jumpers.
The use of spade-type ends on jumpers would provide a solid, reliable connection, but would involve unscrewing the terminal screws to attach the jumpers. Licensee evaluation of a method of jumpering that would be fast and reliable appeared to be warranted."
TVA's Response TVA agrees that there is some difficulty involved in attaching jumpers inside the panels.
However the level of difficulty and danger does not warrant it as a restart item.
TVA is researching ways to facilitate installations.
Necessary changes will be made before startup following the unit 2 cycle 6 refueling outage since this may require additional cable routing.
56.
NRC's Concern "Similarly, Agastat timing relays 10A-K45 A and B that were installed in panels'9-32 and 9-33 as specified in appendix 16 were difficult to reach and were located close to other energized wiring."
TVA's Response Senior Reactor-Operators determined that the level of difficulty and danger in zeroing the Agastat timing relays is not sufficient to warrant any plant modification to facilitate the performance of that action.
57.
NRC's Concern "Also, the terminal numbers that were referenced in Appendix 8 on wires to the relay terminals inside panels 9-15 and 9-17 (main steam isolation valve interlocks) were extremely small and hard to read.
The wires were marked with number tape.
Larger or more easily readable wire and terminal numbering should be used."
' N g
L TVA's Response Labeling of wires to relay terminals in the panels called out in the EDIs has been reviewed.
This review was performed by operations, training, and plant engineering personnel and determined that the labeling was adequate.
58.
NRC's Concern "The team noted that the spoolpiece used to connect auxiliary steam to the RCIC steamline was unsecured on the gratings in the area of valves71-565 and 12-777 and exposed to flange seating surface or other damage. -The licensee should protect such pieces of equipment to ensure they will be available and functional.
The team noted that
~
the blank piping flanges were misaligned by about two inches.
Discussions with operators confirmed that, although the installation was difficult, the spoolpiece could be installed by two people in less than two hours."
TVA's Response The misalignment of the blank flanges has been evaluated and determined that reworking the pipe in not warranted. A maintenance request was written to secure the spool piece and install blank flanges.
completed.
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This work has been 59.
NRC's Concern "The team noted that the spoolpiece used to connect auxiliary steam to the HPCI steamline was unsecured under debris on the mezzanine near valves73-587 and 12-778.
As in the case of the RCIC spoolpiece, this equipment could also be damaged or discarded.
The team noted that the blank flanges where the spoolpiece was to be installed were out of axial alignment by approximately three inches.
Discussions with operators and mechanical maintenance personnel confirmed that connecting this spoolpiece would require chainfalls to spring the flanges into alignment and would require four people working about four hours to complete the connection.
The licensee should consider realigning the auxiliary steamline with the connection on the HPCI steam piping."
,TVA's Response j
1 Maintenance requests were written to install blank flanges, secure the HPCI spool piece to the grating, and realign the steamline with mating surfaces.
This work has been completed.
,'('
60.
NRC's Concern
" Bay numbers were not shown on the outside of the instrument panels in the auxiliary instrument room. Operators attempting to enter a specific bay to perform an interlock bypass or other local action had to open the panel backs to determine the bay number.
Bay numbers were marked on the inside doors of panels with pencil. One relay (16A-K1C, AC) was labeled with a piece of. masking tape and a handwritten letter code.
The licensee should survey the cabinets for similar labeling problems and take corrective action."
TVA's Respcase Auxiliary instrument room labeling has been reviewed and deficiencies noted.
All necessary changes have been completed.
61.
NRC's Concern "The general state of housekeeping in areas observed during the walkdowns of E0Is (outside the control room) was poor (e.g., the team noted debris around HPCI spoolpieces, debris in auxiliary instrument room panels, etc.)."
~
TVA's Response Due to the extended outage many areas have collected debris from various modifications in progress. The specific work packages all have housekeeping sign-offs to verify cleanliness of the general area when work is complete.
No additional action is warranted for this comment.
62.
NRC's Concern "T'ne BFN staff had placed orange arrows with "E0I" in black letters in auxiliary instrument room panels to direct operators to the locations of relays, buses, and terminals where jumpers and bypasses were called for in the E01 appendices.
The "E0I arrows were beneficial to the operators during the walkdown of the appendices."
TVA's Response No response required for this concern Section3,4-ValidationandVerificationProgram 63.
NRC's Concern "Although the applicable validation procedure did not contain detailed guidance, discussions with operations support personnel showed that validation activities were being conducted to the same standards as those applied during the DCRDR task analysis and survey.
The team reviewed the results of these DCRDR-type reviews of portions of E01-1 and E01-2 and all of E0I-3 that were performed in April 1988.
Since the licensee was meeting the intent of the corrective action applicable to this TER problem, PMI 12.9 should be updated to formalize and require this rigorous approach to ensure that it is applied in future validation efforts."
TVAs Response PMI-12.9 has been revised to incorporate directions concerning the acceptability of instrumentation used in the E0Is.
64.
NRC's Concern "The types of problems noted in Sections 3.1 and 3.3 of this report concerning differences between the E0Is and the PSTGs can be partially attributed to the fact that the PSTGs continued to undergo modification after they were used to develop the currently approved revision of the E01s.
Because the PSTGs were not a controlled document, the team could not determine when these changes were made.
Because some of the differences between the E0Is and PSTGs cannot be attributed to these post-E0I development revisions, the licensee should increase the level of attention to future verification efforts.
The abcVe statement is based on a review by the team of a new draft of the E01s in which many of the E0I/PSTG differences noted in Sections 3.1 and 3.3 continue to exist."
TVA's Response The implementation of the E0I Program Manual, which includes PSTGs, into existing plant document control procedures will resolve discrepancies between the E0Is and PSTGs. Once implemented, changes will be subject to normal review and approval cycles for incorporation.
The E0I Program Manual will be in place prior to unit 2 restart.
3,5-Post-accidentReactorBuildingWabitabilityandReentry 65.
NRC's Concern The effects of accident radiation levels in the reactor building on the operators' ability to perform local operations had not been analyzed.
NUREG-0737, item II.B.2, and an NRC Confirmatory Order of July 10, 1981, required the evaluation of personnel access to the reactor building during emergencies.
In its response, prepared before the current symptom-based E0Is were issued, the licensee concluded that the radiation levels would preclude reactor building entry, but that the previous event-based E01s and plant design would support accident mitigation without reactor building reentry.
The symptom-based E01s require entry to compensate for eauipment failures.
Since the development and implementation of the current symptom-based E01s, the licensee had not reevaluated its former position and analysis.
i
t TV4's Response TVA's original evaluation of NUREG-0737, Item II.B.2, concluded that the facility was specifically designed to mitigate major design basis events, utilizing remotely operated equipment from shielded vital areas.
Additionally, access to reactor building locations was unnecessary and, in fact, would be precluded by the severe radiation fields associated with the i
NUREG-0737 source terms.
Subsequently, TVA in conjunction with the BWROG developed the symptom based E01s for BFN. These E0Is provide the operator with a variety of procedural options utilizing both safety-related and nonsafety-related equipment in mitigating a broad spectrum of occurrences, including design basis type events and beyond design basis occurrences. Contingency steps are specifically provided that in the event of majcr degradation of primary safety systems, the operators may utilize all available means to maintain vessel level and containment.
These contingencies include the alignment of auxiliary water injection sources, such as condensate transfer and drain pumps. Operation of these type systems typically requires local manipulations in the reactor building. For cases where significant core damage has occurred, it is likely that the accessing of such systems will be severely restricted because of the radiation hazard to personnel.
The actual radiation field that would be present is, of course, highly dependent on the specific event, extent of core and containment damage, and other factors such as which safety systems were in service. The E01 contingency steps, in essence, provide a means to utilize all available means to mitigate core and containment damage including local manual actions if possible.
The subject E0I provisions are not inconsistent with the NUREG-0737 evaluation or with industry practice.
Section 3,6 - E0I Exercise ilsing the Plant Specific Simulator Simulator Scenariolo,1 66.
NRC's Observation "The ASOS entered and executed E0I-1".
TVA's Response No response required for this observation.
67.
NRC's Concern "The reactor operator told the ASOS that he was going to control reactor pressure by using the safety relief valves before the ASOS I
reached that point in the E01s."
TVA's Response The philosophy of operation at BFN under abnormal or emergency conditions allows the unit operators to perform certain actions before being directed to do so by the ASOS.
In the first few minutes of an abnormal or emergency situation, the operators must take prompt actions per PMI 12.12 to attempt to stabilize the key reactor parameters.
No change to existing procedures is warranted.
t e
u 68 '
NRC's Concern
" Communications between the operators was generally not very good in that there was no formal command / acknowledgment system."
TVA's Response Subsequent to the E01 inspection, revision 3 to PMI 12.12. " Conduct of Operations," was approved. As part of this revision, additional guidance concerning shift communications was incorporated into the procedure.
The procedure now requires repeating back directives that are issued in order to ensure that the receiver of the message clearly understands the directive.
In addition, when the task is completed, the operator will report this back to the supervisor.
Additionally, BFN has provided a specific training lecture (4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />) on communications to all operators during their normal requalification training.
This was one in a series of lectures on teamwork and diagnostics given by the corporate training staff.
BFN has also recently installed a videotape system to allow operators to review their simulator performance with the specific purpose of improving communications and teamwork.
69.
NRC's Concern "One crew erroneously believed the plant was in an anticipated transient without scram (ATWS) condition for the first five minutes following a scram in which all the control rods were fully inserted."
TVA's Response The belief that an ATHS had occurred was brought about by the failure to receive indication that the one rod out interlock permissive had been satisfied (signifying all rods fully inserted).
The crew immediately began checking the full core display to determine if all rods were fully inserted and requested an On Demand (0D)-7 printout from the process computer to verify the rod positions.
Until such time as the crew was able to verify all rods fully inserted, they took the conservative approach and assumed that an ATHS did exist. While this may have required some additional actions to be taken, the basic actions taken to control the key reactor parameters were the same for both the ATHS case and the non-ATWS case.
These types of misdiagnoses were considered in the development of the EPGs.
Through the use of symptom-oriented procedures and step completion verification actions, errors of this type should not have a significant impact on the outcome of the transient.
This philosophy was addressed by NRC in the Safety Evaluation Report written for revision 4 of the EPGs which states:
"The guidelines address operator errors by checking the effects of directed operator actions and providing guidance for those cases where previous operator actions were unsuccessful."
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70.
NRC's Observation "The ASOS generally entered and executed the correct E0Is and contingency procedures."
TVA's Response No response necessary for this observation.
71.
NRC's Observation "The ASOS incorrectly read a logic diagram in Contingency C1,
" Alternate level Control," and proceeded to Step C7 instead of Step Cl-7.
The ASOS corrected his error before proceeding with Step C7."
hVA'sResponse No response necessary for this observation.
72.
NRC's Concern a
"The ASOS did not consistently use the checkoff blanks associated with each step for peacekeeping (finding and keeping the correct place in the procedures) as required by management directive."
TVA's Response Subsequent to the EDI inspection, revision 3 was made to PMI 12.12, that provides more detail on the responsibilities of control room personnel during abnormal and emergency conditions.
Included is guidance for the ASOS in procedure peacekeeping for the AOIs and E01s.
Training presently stresses peacekeeping while executing the E01s and no further action is warranted.
73.
NRC's Observation "Neither crew was aware that a station blackout condition had occurred.
For example, the ASOS asked about the availability of ac-powered injection systems (core spray and residual heat removal) and the abnormal procedure for station blackout was not entered."
TVA's Response During the first few minutes of the scenario, the ASOS is preoccupied with stabilizing-key reactor parameters while analyzing overall plant condition.
Part of this. process involves feedback from operators on equipment availability.
The immediate operator actions for the station blackout were performed (attempt to restore diesel generator availability), however, reactor parameters required priority in other procedures.
No action is necessary for this observation.
74.
NRC's Observation "The ASOS made the decision to perform the reactor flooding contingency procedure on the basis of existing plant conditions instead of actually following the E0Is that directed that action."
2
.TVA's Response
.The ASOS was following the directions in the E01s to flood the reactor vessel based on loss of vessel level instrumentation.
Each operator is aware that flooding is required whenever level cannot be determined.
This scenario j
placed the reactor in steam cooling while attempts were made to restore an injection system.
The procedure subsequently required depressurization.
At the appropriate point the determination was made that level indication was not available.
This alerted the operators to flood the vessel which was appropriate.
No action is necessary for this observation.
75.
NRC's Observation "The reactor operator left three safety relief valves open while he was supposed to be controlling reactor pressure in a band specified in the E01s. This resulted in an unplanned reactor depressurization before the emergency depressurization step in the E0Is was reached."
TVA's Response The operator was given a band in which to control reactor pressure.
The oper'ator became involved in multiple actions directed by the ASOS and allowed pressure to drop below the prescribed band.
The reactor did not depressurize to the point where the normal cooldown rate was exceeded, however, some unnecessary heating of the torus occurred.
The operator discovered the error and made corrections.
No action is required far this observation.
SimulatorScenarioNo.3 76.
NRC's Observation "The ASOS generally entered and executed the correct E0Is and contingency procedures, which involved simultaneous execution in several areas."
TVA's Response No response necessary for this observation.
77.
NRC's Observation "During the first 15 minutes of the scenario, the ASOS could not keep up with all the multiple-action steps in the E01s that had to be performed concurrently."
i TVA's Response Due to the time frame involved with certain scenarios, the ASOS is responsible for determining procedure priorities based on reactor parameters. Once the most urgent parameters have been addressed, the remaining actions are performed. No changes are deemed necessary for this observation.
)
e f8.
NRC's Observation i
"One crew did not vent the suppression chamber as required by the E01s when the suppression chamber air temperature was less that 210*F.
Because the ASOS erroneously read drywell temperature instead of suppression pool temperature and found that it was more that 210*F, he assumed venting could not be initiated at that time."
TVA's Response The ASOS took the correct approach to venting.
Drywell temperature is the parameter to consider when suppression chamber venting is considered.
No action is necessary for this observation.
79.
NRC's Observation "One crew entered the drywell spray initiation curve using drywell temperature instead of suppression chamber air temperature as required by E01-2.
This delayed the attempt to spray the drywell for about 15 minutes.
TVA's Response Larger curv'es have been incorporated into the E01s.
See TVA's response to NRC Concern No. 25.
80.
NRC's Observation "The presence of a safety parameter display systen, would have greatly improved the ability of the operators to use the E0Is and manage the emergency.
The safety parameter display system is not scheduled for installation until the next operating cycle after restart.
TVA's Response TVA agrees with the observation and is presently planning installation of the safety parameter display system during refuel cycle 6 which has previously been communicated to NRC.
81.
NRC's Concern "Use of the shift technical advisor (STA) and shift operations supervisor (SOS) during major events needs to be improved.
In addition, the roles of STA and the SOS should be better defined to ensure that someone is aware of the overall integrated picture of plant status and the mitigating actions being taken."
TVA's Response Subsequent to the E01 inspection, revision 3 to PMI 12.12 was approved.
Part of this change was to more clearly define the roles of all operations personnel including STAS during abnormal and emergency conditions.
Training has also been modified since the inspection to include a SOS and STA during regular requalification simulator performances.
STAS currently attend i
classroom training with operations personnel during requalification.
r
- 7. -
s7.'
NRC's Concern "Because of time restraints, the AS05 did not consistently use the procedure checkoff blanks."
TVA's Response Subsequent to the E0I inspection, a revision was made to PMI 12.12, that provides more detail on the responsibilities of control room personnel during abnormal and emergency conditions.
Included in here is guidance for the ASOS in procedure peacekeeping for the AOIs and E01s.
Training presently stresses peacekeeping while executing the E0Is and no further action is warranted.
83.
NRC's Concern "One crew relied on the reactor operators to evaluate the heat capacity temperature limit and other curves in the E0Is, but the other crew relied on the ASOS to perform these evaluations. The roles and responsibilities of the operators in all crews should be consistent."
TVA's Response All lice ~nse personnel and STAS are trained in the use of curves and figures in the E01s. Due to time restraints each member is subject to checking limits, however, the ASOS is cognizant for the procedure.
No action is warranted for this comment.
SimulatorScenarioNo.4 84.
NRC's Observation "The ASOS entered and executed the correct E01s and contingency procedures.
However, during one performance of the scenario, the ASOS did not physically get out the E0Is and read them until approximately seven minutes after the entry condition was met.
TVA's Response The ASOS was aware of the entry condition to E0I-3 and was attempting to isolate the leak per the procedure.
No action is necessary for this observation.
+S
/r ENCLOSURE 2
.t COMMITMENTS CONTAINED IN RESPONSE TO NRC INSPECTION REPORT NO. 50-260/88-200 LETTER FROM J. G. PARTLOW TO S. A. WHITE j
DATED SEPTEMBER 18, 1988 List of Commitments NOTE: In parentheses at the end of each commitment is the response number in enclosure 1 where that commitment can be found.
1.
The E0I Program Manual including PSTGs, PSTG appendices, and deviations cross reference document will be incorporated into the existing plant document control system before unit 2 restart.
This manual will require QA review, technical reviews as required, affected section review, screening reviews for USQDs, and other controls imposed under the existing system. Once implemented, changes will be subject to normal review and approva! cycles for incorporation.
(36, 38, and 64) 2.
TVA will perform an evaluation on reactor building pressurization and ground level release prior to unit 2 restart. (39) 3.
TVA will perform an evaluation of the containment vent path valve operability during post accident differential pressure. Once the evaluations are complete, the necessary changes will be made to the PSTGs and the E0Is prior to unit 2 restart.
(40) 4.
Revision 4 of the BhROG EPGs will be implemented during 1990. An evaluation of local / manual venting under blackout conditions or loss of offsite power will be performed in conjunction with implementation of EPG revision 4.
(41) 5.
An eve'uation will be performed to determine the basis for the 160* F drywell temperature scram prior to unit 2 restart.
(50) 6.
TVA is researching ways to facilitate jumper installations.
Changes, if required, will be made before startup following the unit 2 cycle 6 refueling cycle.
(55)
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