IR 05000296/1993201
| ML18036B140 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 01/08/1993 |
| From: | Imbro E, Malur S, Norkin D Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML18036B139 | List: |
| References | |
| 50-296-93-201, NUDOCS 9301220235 | |
| Download: ML18036B140 (74) | |
Text
U.S.
NUCLEAR REGULATORY COHHISSION OFFICE OF NUCLEAR REACTOR REGULATION NRC Inspection Report:
50-296/93-201 Docket:
50-259/50-260/50-296 License No.:
DPR-33 DPR-52 DPR-68 Licensee:
Tennessee Valley Authority (TVA)
Facility Name:
Browns Ferry Nuclear Plant Inspection Conducted:
From November 16 through December 4, 1992 Inspection Team:
S.
K. Halur, Team Leader, NRR Hai-Boh Wang, NRR Donald C. Prevatte, Parameter, Inc.
Omar S. Hazzoni, Parameter, Inc.
James H.
L 1vo, Parameter, Inc.
Prepared by:
S.
K.
H r,
Team Leader Team Inspection Section A
Special Inspection Branch Division of Reactor Inspection and Licensee Performance Office of Nuclear Reactor Regulation Dat Reviewed by:
DonaTd P. Norkin, Section Chief Team Inspection Section A
Special Inspection Branch Division of Reactor Inspection and Licensee Performance Office of Nuclear Reactor Regulation Approved by:
Eugene V. Imbro, Chief Special Inspection Branch Division of Reactor Inspection and Licensee Performance Office of Nuclear Reactor Regulation 930i220235 930ii5 PDR ADOCK 05000259
I
EXECUTIVE SUMMARY A U.S. Nuclear Regulatory Commission (NRC) Plant Design Change Inspection at Browns Ferry Nuclear Plant, Unit 3, was conducted by the Special Inspection Branch of the Office of Nuclear Reactor Regulation (NRR) from November
through December 4, 1992.
The purpose of this performance-based inspection was to verify that the process for the design of modifications to Unit 3 appropriately implemented NRC regulations and the Tennessee Valley Authority's (TVA's) commitments, and that the design controls for the design and field changes were adequate.
This was accomplished by inspecting selected design change notice (DCN) packages issued for Unit 3 during 1991 and 1992, as well as DCNs for Unit 2 which were also applicable to Unit 3.
The team reviewed DCNs and field changes to DCNs (FDCNs),
as well as
. calculations, drawings,, procedures, and other documents pertinent to the DCNs.
The team's sample review of the documentation for design changes indicated that the design change process was adequately implemented'and controlled.
In general, the design change process complied with the regulatory requirements and licensing commitments.
The design documents were generally thorough and exhibited good attention to detail and were readily accessible and retrievable.
TVA's engineering personnel were competent and were able to respond effectively to the team's questions.
They were knowledgeable of some of the industry initiatives, such as the reactor level ins'trumentation and electromagnetic interference qualification issues.
While the team's observations indicated that the overall design change process was properly implemented and controlled, the team had several specific concerns.
The team was concerned about the lack of evaluation of the radiological dose to the operator while performing local manual draining of the hardened wetwell vent line during emergency venting of the containment and the lack of evaluation of the ground-level release due to the open drain valyer..
At a meeting with the NRC staff on December 16, 1992, TVA agreed to change the design to eliminate manual operation of the drain valves during venting and evaluate the consequences of ground-level release of radioactive materials from open drain valves during various plant transients.
The team's other concerns were related to: (I) inadequate consideration of operating conditions in the flow and condensation calculation for the wetwell vent which resulted in non-conservative water accumulation rates and failure to consider water hammer effects in the vent pipe and 'structural adequacy of the stack deck, (2) inadequate evaluation of the control air system excess flow check valve setpoint that could result in loss of control air to a unit during transient events, (3) unsubstantiated assumptions in the electrical calculations, such as assuming 480 V for short circuit calculations instead of the prefault bus voltage of 508 V which would result in unacceptable short circuit margins, and (4) inadequate verification and justification of field changes.
TVA agreed to revise the calculations in response to the team's comment EXECUTIVE SUMMARY.
1.0 INTRODUCTION.
SUMMARY OF DEFICIENCIES BASIS:
DEFICIENCY 93-201-01 IIL Fii E
i th Rdil i
1C q
d <<h Operation of the Vent Drain Valves (Section 3.2)
Generic Letter 89-16 states that as a mitigation measure, a reliable wetwell vent provides assurance of pressure relief with significant scrubbing of fission products.
In the summary discussions on the installation and operation of the hardened vent for BWRs with Hark I containments in Federal'egister, Vol. 55, No.
122, June 25, 1990, it is stated that the incremental occupational radiation dose for the proposed operation of the hard pipe vent path was insignificant (unmeasurable)
because the vent path would be operated from the control room..
In the background discussions in the same document it is stated that the vent.
was intended to provide a scrubbed pathway for containment pressure relief for
"
rare situations involving core damage.
BWR Owner's Group Emergency Procedure Guidelines, Revision 4, requires venting of the primary containment irrespective of the offsite radioactivity release rate when the suppression chamber pressure exceeds the primary containment pressure limit.
Section 3.2 of the ANSI N45-2. 11-1974, which is specified in TVA's plant modification and design change control procedure SSP-9.3, states that the design input shall include environmental conditions such as nuclear radiation anticipated during construction and operation.
DESCRIPTION OF CONDITION:
DCNs W17337 and W17491 (References
and 2) provide the hardened wetwell vent common header and vent line from the Unit 2 torus to the common header, respectively.
This line is'eing installed in response to Generic Letter 89-16 (Reference 3) to reduce the vulnerability of BWR Hark I containments to severe accident challenges.
One of the design features of these DCNs is that there are three normally closed manual drain valves which must be manipulated in order to effect venting.
Two of the drain valves are on the vent line itself, one inside the reactor building and one in a drain pit in the yard.
The third is located in the drain line for the deck area at elevation 665'-6" inside the stack.
The DCNs did not evaluate the radiological consequences due to the operation of the drain valves.
TVA stated that this modification was intended to address the TW event only (Reference 4), i.e., the transient event with failure to remove long term A-1
~ ~
decay heat, and that for this event there was no significant radioactive source term.
However, based on References 3, 4, and 5 (design bases for these DCNs), the team believes that transient events which entail fuel failure are to be considered in the design, and that the operation of the vent, including any attendant operation such as draining, is expected to be conducted remotely from the control room to minimize exposure to the operator.
Reference 3 states that, "First, it is recognized that all affected plants have in place emergency procedures directing the operator to vent under certain circumstances (primarily to avoid exceeding the primary containment pressure limit) from the wetwell airspace.
Therefore, incorporation of a designated capability consistent with the objectives of the emergency procedure guidelines is seen as a logical and prudent plant improvement."
Emergency procedures do not differentiate, between types of accidents; they respond to symptoms.
The BFNP emergency procedures (Reference 6),
and the BWR.
Owners'roup (BWROG) Emergency Procedure Guidelines (Reference 8) require venting irrespective of offsite radioactivity release rate.
Reference 5 states in the summary section that,
"The incremental occupational dose for the proposed operation of the hard pipe vent path is insignificant (unmeasurable)
because the vent path would be operated from the control room,"
indicating the intent that the operation be from the control room to minimize operator exposure.
At a meeting with the NRC staff on December 16, 1992, TVA stated that the manual drain valves in the reactor building and in the yard would be replaced with float 'operated automatic type drain valves, and that TVA will investigate the feasibility of keeping the stack deck drain valve normally open, provided that the consequences of ground-level radioactive material release during various plant transients were acceptable.
The NRC staff agreed with this approach.
REFERENCES:
1'.
DCN W17337 A, "Install the Common Header Portion of the Hardened Wetwell Vent".,
2.
DCN'W17491 A, "Unit 2 Torus to the Common Header Portion of the Hardened Wetwell Vent" 3.
Generic Letter 89-16, September 1,
1989, "Installation of a Hardened Wetwell Vent" 4.
NUREG/CR-5225, November, 1988,
"An Overview of BWR Hark-I Containment Venting Risk Implications" 5.
Federal Register, Vol 55, No 122, Monday'; June 25, 1990, Notices,
"Installation and Operation of Hardened Vent from Suppression Pool
- Airspaces of Boiling Water Reactors (BWRs) with Hark I Containments" A-2
l
6.
2-EOI-2, BFNP Emergency Operating Instruction,
"Primary Containment Control" 7.
ANSI 45.2. 11-1974, "guality Assurance Requirements for the Design of Nuclear Power Plants" 8.
BWR Owners'roup,
"Emergency Procedure Guidelines",
Revision 4, NEDO-31331, Class I, Harch 1987
A-3
l
~
~
~
J J
a
I t
DEFICIENCY 93-201-02 FINDING TITLE:
BASIS:
Discrepancies in Hardened Wetwell Vent Calculation (Section 3.2)
CFR Part 50, Appendix 8, Criterion III, requires, in part, that design control measures shall provide for verifying or checking the adequacy of desi gni Section 4.1 of ANSI N45.2.11 which is specified in TVA's plant modification and design change control procedure SSP-9.3, requires, in part, that design activities shall ensure that applicable design inputs are correctly translated into specifications, drawings, procedures, and instructions.
DESCRIPTION OF CONDITION:
DCNs
~ W17337 and W17491 (References 1 and 2) provide the hardened wetwell vent common header and vent line from the Unit 2 torus to the common header, respectively.
a The design analyses of the flow and moisture conditions inside the pipe and the stack 'during the operation of the hardened wetwell vent are contained in Calculation HD-f0999-920051 (Reference 4).
This calculation was not adequately verified for completeness, accuracy, and adequacy as evidenced by the follow'ing technical discrepancies:
I.
The estimated amount of moisture in the vent line did not address two potentially large sources:
wetwell spray and flashing of the wetwell wheh the containment pressure is reduced.
This additional water would affect the flow capacity of the vent, the accumulation rate and mass of water on the deck inside the stack, and the potential for water hammer in the line.
2., The analysis did not consider the moisture removal rate in the pipe at
'ower pressure conditions such as would be present subsequent to its
initial opening, and for the steady state flow rates corresponding to
.
lower than the design decay heat rate conditions.
3.
The stack deck drain line capacity computation used a drain flow rate that corresponded to the relatively low condensation rate at the end of eight hours rather than early in the event when the condensation rate is several times higher.
The calculation assumed a water depth of 3 feet on the stack deck without considering the ability of the deck to support such a
load.
The structural integrity of this deck under the most conservative conditions of water accumulation was not evaluated.
4.
The flow performance of the vent line is based on an assumption of "no flow area restrictions in the line."
However, the internal diameter (I.D.) of the three in-line butterfly valves is less than the I.D. of the A-4
f
5.
pipe, and they each have a three inch wide disk in the middle of the flow path.
This error impacts the flow capacity of the vent, and the condensation rate in the line.
The computation of the condensation in the vent during its initial heatup only considers heatup of the pipe; it does not consider the heatup of the surrounding media during this phase.
This produces a non-conservative result.
6.
The calculation concludes that, even with the vent line full of water, it is functional because there is sufficient pressure to blow the water out.
However, it fails to consider the structural integrity of the pipe for static water loads, water hammer loads, and other dynamic loads associated with blowing a slug of water out of the pipe.
It also fails to consider the rapid deposition of this large volume of water on the stack deck, with the resultant potential for structural failure of the deck.
TVA has agreed to revise this calculation to resolve the deficiencies mentioned above and other minor discrepancies identified during the inspection.
REFERENCES:
DCN W17337 A, "Install the Common Header Portion of the Hardened Wetwell Vent" 2.
DCN W17491 A, "Unit 2 Torus to the Common Header Portion of the Hardened Wetwell Vent" 3.
5.
6.
7.
Calculation HD-90999-920051, Revision 0, "Stack Evaluation During Wetwell Purge" Anchor-Darling Drawing 94-15973, Revision F,
"14", ANSI 150¹ Class Lugged Style Butterfly Valve with Limitorque Manual Operator" Anchor-Darling Drawing 94-15972, Revision E, "14", ANSI 150¹ Class, Fail Close, Wafer Style Butterfly Valve with Bettis Air Operator"
CFR Part 50, Appendix 8, "guality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants" ANSI N45.2. 11-1974, "guality Assurance Requirements for the Design of Nuclear Power Plants" A-5
e
DEFICIENCY 93-201-03 FINOI TILE:
1Ai Eyt E
F1 h kVi (Section 3.3)
BASIS:
Section 4. 1 of ANSI N45.2. 11, which is specified in TVA's plant modification and design change control procedure SSP-9.3, requires, in part, that design activities shall assure that applicable design inputs are correctly translated into specifications, drawings, procedures, and instructions.
Generic Letter 88-14, requested that licensees verify that the design of the instrument air system is in accordance with the intended function.
DESCRIPTION OF CONDITION:
The BFNP control air system is a
common system serving all three units.
Because of a concern that a piping failure in one; unit could cause plant trips in all three units upon loss of the common system pressure, DCN WI7416A was prepared.
The DCN provided automatic isolation of the control air to the unit in which the failure would occur.
The isolation was effected by installation of an excess flow check valve in each of the individual unit supply lines upstream of the dryers.
In the event of pipe failure in one unit, the increased flow rate would actuate the excess flow check valve, and the system common air receivers would not be depressurized.
Only the unit experiencing the break would lose control air.
The normal control air system flow rate to each unit was estimated at 490 standard cubic feet per minute (scfm).
The excess flow check valve setpoint was established at 915 scfm, accounting for setpoint tolerance and additional flow during regeneration of the air dryers.
Although the excess flow check valves will provide the desired pipe break isolation, plant transients can also cause actuation of these valves and isolation of control air to the unit experiencing the transient.
For example, a unit trip will cause the MSIVs to close and other air operated equipment to actuate.
The demand of the HSIVs alone for this event is estimated at 1,040 scfm for the outboard valves (Reference 1).
The selection of the setpoint for the excess flow check valves did not consider and appropriately document the control air flow rates during various plant conditions (Reference 2).
Therefore, for any plant transient that involved HSIV closure or other large control air demand, the control air to that unit could automatically isolate.
This appears to have not been considered in the preparation of this DCN.
REFERENCES:
1.
Calculation HD-(0032-870334, Revision 2, "Control Air System - Pipe Sizing Flow" 2.
Calculation Md-N0032-920165, Revision 1, "Control Air System Flow Rate During Pipe Failure" A-6
t C'
~
L
3.
DCN M17416A, "Control Air Dryer Replacement" 4.
"Instrument Air Supply Problems Affecting Safety-Related Equipment" 5.
ANSI 45-2.11-1974,
"equality Assurance Requirements for the Design of Nuclear Power Plants" A-7
~ 4 J
k
DEFICIENCY 93-201-04 FINDING TITLE:
BASIS:
Document discrepancies and unsubstantiated assumptions in calculations (Section 4.3)
CFR Part 50, Appendix B, Criterion III, requires, in part, that design control measures shall provide for verifying or checking the adequacy of the design.
ANSI N 45.2. 11-1974, Section 3. 1, requires, in part, that applicable design inputs, such as design basis, regulatory requirements, and codes and standards, shall be identified, and documented, and their selection shall be reviewed and approved.
DESCRIPTION OF CONDITIO:
The following calculations contained unsubstantiated assumptions and other'.
discrepancies:
a) Calculation EEB/ED-(3999-920106, Revision 0, "Input Data", Sheet 3:
The reference to maximum voltage on all boards was not clear, since this maximum voltage was stated as two possible values:
270V and 280V.
TVA agreed to revise the'alculation to use the correct maximum voltage.
b) Calculation EEB/ED-(3999-920106, Revision 0, "Input Data", Sheet 4:
There was no basis for assuming a maximum of 10 contacts in a circuit.
The team found that the scheme shown in Figure 2 of Attachment 1. 1 of the same calculation displayed ll contacts connected in series.
TVA agreed to add detailed wiring/block diagrams in the next revision of the calculation so that the number of contacts in a particular path could be identified.
c) Calculation EEB/ED-(3999-920106,
"Input Data", Sheet 4:
There was no basis for assuming 0. 1 ohm ',resis'tance for each contact.
The team found that this value was not supported by the vendor documents.
TVA agreed to obtain the correct resistance value and revise the calculations.
d) Calculation ED-(3057-920329, Revision 0, Attachment A, Section 5.0: There was no verification of the assumed impedance of 8X for the new transformers.
A small variation in the impedance could affect the short circuit rating of the equipment.
In response to the team's comment, TVA provided adequate supporting information for the assumed impedance value and agreed to revise the calculation to incorporate the justification for the assumption.
e) Calculation ED-(3057-920329, Revision A, Attachment A, Section 6.0:
There was no basis for assuming that a voltage of 480V was the most conservative value that should be considered.
In response to the team's question, TVA investigated the prefault bus voltage and found it to be 508V, rather than the 480V assumed in the calculations.
TVA further calculated that with the revised voltage value the short circuit margin was (-) 4%, which is not A-8
h
~
p
~
~
~
a e
acceptable.
TVA indicated that the problem would be resolved by relocating the power supply to a
150 HP chiller motor to another bus.
TVA agreed to revise the calculation.
f) Calculation ED-f3057-920035, Section 2.0:
TVA had no documentation to justify the elimination of the small transformer and the cable losses from the assumed loading conditions.
In response to the team's question, TVA prepared a detailed evaluation to support the assumption.
The team found this adequate.
TVA agreed to revise the calculation to incorporate the supporting document.
g) Calculation ED-(3057-920035, Section 4.0:
There was no documentation to support the statement that "service factors were conservatively considered equal to one."
In response to the team's question, TVA prepared a detailed evaluation of the actual motor loads and concluded that motor horsepower ratings will not be exceeded.
TVA agreed'to revise the calculation to incorporate the supporting document.
h) Calculation ED-03057-920329, Attachment A, Section 2.0 There was no documentation to support the assumption that a static load of 200 kVA was conservative.
In response to the team's question, TVA prepared a
calculation to verify the assumption.
TVA agreed to revise the calculation to incorporate the supporting document.
REFERENCES:
1.
Calculation EEB/ED-(3999-920106, Revision 0,
"CCVD For Unit 3 DC Circuits" 2.
Calculation ED-(3057-920329, Revision 0,
"480V Short Circuit Calculations" 3.
Calculation ED-(3057-920035, Revision 0,
"Emergency Diesel Generator Loading" 4.
.ANSI N45.2. 11-1974, "guality Assurance Requirements for the Design of
.Nuclear Power Plants" 5.
CFR Part 50, Appendix 8; "guality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants" V'
A-9
~ A
~
e
DEFICIENCY 93-201-05 FINDING TITLE:
Inadequate Justification for Basis for Approval and Inadequate Design Verification of FDCNs (Sections 3.4 and 4.7)
BASIS:
Section 8 of ANSI N45.2. 11, which is specified in TVA s plant modification and design change control procedure SSP-9.3, requires, in part, that the impact of the design changes to approved documents, including field changes, be carefully considered, and such changes be justified.
CFR Part 50, Appendix B, Criterion III, Design Control, requires, in part that design changes, including field changes, shall be subjected to design control measures commensurate with those applied to the original design.-
Criterion III also requires, in part, that the design control measures shall provide for verifying or checking the adequacy of design, such as by the performance. of design reviews, by the use of alternate or simplified calculational methods or by the performance of a suitable testing program.
DESCRIPTION OF CONDITION:
'0 a)
FDCN F19486A DCN W17724B, Section 2A, "Post Modification Tests," required that core megger'test be performed on the unit service station transformer.
This test requirement was deleted under FDCN F19486A.
The justification for the deletion of'the test was not documented.
Block 16 of the FDCN cover page required inclusion of basis for approval, but no basis for the approval was stated.
In response to the team's comments, TVA's contractor prepared a memorandum (Reference 5) that justified the field change on the basis that the transformer core was not affected by the modification, a megger test was done prior to the modification and the risk involved in disconnecting the core strap to perform the test and reinstalling the strap was not commensurate with the advantages of repeating the megger test.
The team agreed with the justification provided, but this change should have been properly justified and the design should have been verified before the FDCN was implemented.
b)
FDCN F19576A DCN W17667A was modified by FDCN F19576A, to defeat the emergency run back relay (27A) for LTC operation.
Two relays were added to the LTC control circuit to provide a means of defeating the emergency run back feature when the LTC controls are in manual mode.
No justification was provided in the FDCN for this change.
In response to the team's question, TVA stated that the reasons for the field change were that the automatic LTC movement could occur in the
~
g ~
c)
manual mode and this was undesirable during maintenance activities because the manual mode is utilized to ensure personnel protection.
The team agreed with this explanation.
However, the required justifications should have been included in the FDCN and the design should have been appropriately verified before the change was implemented.
FDCN 19573A DCN W18711B replaced and rerout'ed the reactor water cleanup piping which was subject to intergranular stress corrosion cracking (IGSCC).
FDCN 19573A was issued to repair IGSCC indications at one location that was not previously discovered.
The repair was done under a
GE Field Design Deviation Request (FDDR),
and the FDDR cover sheet indicated that the design verification was also performed by GE.
In order to incorporate the FDDR into the Browns Ferry documentation'ystem,.
an FDCN cover sheet was attached and signed by TVA engineers in various positions of responsibility, including the design verifier.
When the team asked how the TVA engineer who signed as the design verifier performed this task with the seemingly inadequate information that was in the package, the TVA engineers stated that this signature only meant that the package was being accepted from GE, not that the signee had actually performed a design verification.
The existing procedures (References 8, 9, and 10) do not allow signing the cover sheet of the FDCN if no verification was performed by TVA.
There were no specific exclusions of this requirement in the procedures for contractor generated documents produced in accordance with contractor's approved procedures.
REFERENCES:
2.
3..
4.
5.
6.
7.
8.
9.
io.
DCN W17724B,
"USST Modification" FDCN F19486A,
"Delete Test Requirement" DCN W17667A, "Install automatic Load Tap Changer (LTC) on CSST 'B'"
FDCN F19576A,
"Defeat Emergency Run Back Relay (27E)"
Bechtel Interoffice Memorandum, from Carl Hurty to Warren Banner, dated December 2,
1992 DCN W18711B,
"Replace Reactor Water Cleanup Piping" FDCN F19573A,
"Repair IGSCC Indications" Site Standard Practice, SSP-9.3, Revision 4, "Plant Modifications and Design Change Control" Site Standard Practice SSP-9.5, Revision 0,
"Design Engineering" Site Engineering Practice SEP-9.5.6, Revision 0,
"Design Verification" A-11
11.
"equality Assurance Requirements for the Design of Nuclear Power Plants" 12.
CFR Part 50, Appendix B, "equality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants" A-12
IIll A
'c
PPENDI LIST OF OBSERVATIONS During the plant design change inspection at the Browns Ferry Nuclear Plant, Unit 3, the U.S. Nuclear Regulatory Commission's inspection team made the following observations.
The sections to which these observations apply are given in parentheses.
Observation 93-201-01,
"Inadequate Requirements for Ensuring Valve Performance After Stem Packing Replacement" (Section 3.6)
Observation 93-201-02,
"No Evaluation of Lithium Batteries as Potential Fire Hazard" (Section 5.2)
Observation 93-201-03,
"Inadequate equal'ification of Digital Recorders for Electromagnetic Interference Testing Requirements" (Section 5.2)
4 +.w
E.T.Knuettel O.S.Hazzoni J.F.Williams R.N.Lilleston J.Valente J.E.Haddox R.R.Baron O.V.Zeringue B.A.Wilson D.C.Prevatte H.Wang S.K.Halur J.H.Leivo C.A.Patterson R.A.Husser G.D.Pierce P.J.Rush J.T.Hunday C.H.Crane J.W.Smithson D.B.Harrison R.H.Wright D.T.Langley P.Salas J.R.Rupert PPENDIX C
EXIT MEETING ATTENDEES
~OR Il Tl TVA Licensing NRC Consultant NRC Project Manager TVA Recovery Support PE TVA Recovery Manager TVA Engineering Manager TVA Site guality Manager TVA VP BFN Operations NRC RII,Chief DRP Branch
NRC Contractor NRC Inspector NRC Team Leader NRC Contractor NRC Senior Resident Inspector NRC Resident Inspector TVA Site Licensing Manager NRC NRR Intern NRC Resident Inspector TVA Maintenance Manager TVA Special Projects Manager TVA Recovery Manager TVA Electrical Engineer TVA Electrical Engineer TVA Compliance Manager TVA Engg.
5 Modifications Manager
C T
~
l