ML20137W939
| ML20137W939 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 04/15/1997 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20137W930 | List: |
| References | |
| 50-289-96-201, NUDOCS 9704210100 | |
| Download: ML20137W939 (55) | |
See also: IR 05000289/1996201
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U.S. NUCLEAR REGULATORY COMMISSION
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OFFICE OF NUCLEAR REACTOR REGULATION
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Docket No.:
50-289
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License No.
Report No.:
50-289/96-201
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Licensee:
GPU Nuclear Corporation
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Facility:
Three Mile Island - Unit 1
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Location:
Middletown, PA.
Date:
December 2 through 20, 1996 and Ja.nuary 6 through 10,
1997
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Inspectors:
S.K. Malur, Team Leader, Special Inspection Branch
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C.J. Baron, Contractor *
R.B. Bradbury, Contractor *
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G. Garabedian, Contractor *
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T. Landry, Contractor *
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D. Schuler, Contractor *
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(* Contractors from Stone Webster Engineering Corporation)
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Approved by:.
Robert M. Gallo, Chief
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Special Inspection Branch
Division of Inspection and Support Programs.
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Office of Nuclear Reactor Regulation
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TABLE OF CONTENTS
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Executive Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
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El Conduct of Engineering
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El.1
Insoection Scone and Methodoloav
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El.2 Makeuo and Purification System
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El.2.1 System Description and Safety Functions . . . . . . . . . .
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El.2.2 Mechanical Design Review
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El.2.3 Electrical Design Review
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El.2.4
Instrumentation and Control Desi
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System Interfaces . . . . . . . gn Review . . . . . . . . .
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El.2.5
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El.2.6 System Wal kdown . . . . . . . . . . . . . . . . . . . .
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El.2.7 FSAR and SDBD Review
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El.3 Decay Heat Removal Sys tem . . . . . . . . . . . . . . . . . . . . .
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El.3.1 System Description and Safety Function
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El.3.2 Mechanical Design Review
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E1.3.3 Electrical Design Review
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El.3.4 Instrumentation and Control Design Review . . . . . . . . .
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E1.3.5 System Interfaces . . . . . . . . . . . . . . . . . . . . .
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E1.3.6 Sys t em Wal kdown . . . . . . . . . . . . . . . . . . . . . .
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E1.3.7 FSAR and SDBD Review
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'El.4 Control of Calcul ations . . . . . . . . . . . . . . . . . . . . . .
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Exit Meeting
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TAPPENDIX A Open Items
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' APPENDIX B Exit Meeting Attendees
B-1
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APPENDIX C List of Documer.ts Reviewed
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2 APPENDIX D List of Acronyms
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EXECUTIVE SUMMARY
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A design inspection at Three Mile Island - Unit 1 (TMI-1) was performed by the
Special Inspection Branch of the Office of Nuclear Reactor Regulation (NRR)
during the period November 12, 1996 through January 10, 1997 including on-site
inspections during December 2-20, 1996 and January 6-10, 1997.
The inspection
team consisted of a team leader from NRR and five engineers from Stone &
Webster Engineering Corporation.
The team selected for inspection, the engineered safeguards functions of the
makeup and purification (MU&P) system and decay heat removal system (DHRS)
because of the importance of these two systems in mitigating design basis
accidents at TMI-1. The purpose of the inspection was to evaluate the
capability of the systems to perform safety functions required by their design
bases, the adherence to the design and licensing bases, and the consistency of
the as-built configuration with the final safety analysis report (FSAR).
The
bngineering design section of inspection procedure IP 93801 was followed for
this inspection.
The team selected and reviewed the relevant portions of the FSAR, the design
basis documents, drawings, calculations, modification packages, surveillance
procedures, and other associated plant documents.
The team noted that the
design basis documents and modification packages for the systems were detailed.
and were well written.
Safety evaluations for plant niodifications reviewed by
the. team were thorough and reached appropriate conclusions.
The as-built
configuration of the systems was consistent with the design drawings, except
for minor discrepancies.
The system design documents reviewed by the team
adequately supported the design, except for the items discussed in the
following paragraphs.
The team identified that the environmental and dynamic effects of a letdown
E line break in the auxiliary building as discussed in Appendix 14A of the FSAR
apparently had not been adequately evaluated and documented.
The team
-referred this issue to the NRR staff for review regarding the extent to which
THI-1 was required to consider the effects of a letdown break in the auxiliary
building. The staff review concluded that the TMI-l licensing basis for pipe
breaks includes the postulation of full diameter breaks in the letdown line
between the containment penetration and the breakdown orifice as described in
Appendix 14A to the FSAR.
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In the calculations for available~ net positive suction head (NPSH) for makeup
' pumps and the required makeup tank pressure and level to preclude gas
entrainment in the system, the team identified concerns regarding missing
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inputs for a section of pipe and fittings and non conservative pump flows,
BWST level and temperature. The licensee evaluated the team's concerns,
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concluded that the makeup pumps would operate for a short period of time with
degraded NPSH, and initiated changes to procedures to limit makeup pump flow
during accident conditions.
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The design of the control scheme for the makeup isolation valve was such that,
under certain system alignment conditions, if the valve was not supplied with
power from the appropriate DC power supply source the high pressure injection
syst'em performance would be affected in the event of a line break downstream
of the last check valve in the line.
This problem was also identified by the
licensee a few weeks prior to this inspection, but the DC power source
alignment in the plant was not verified.
In response to the team's question
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whether the system was correctly aligned in the plant, the licensee inspected
the system lineup and corrected the DC power alignment, and issued temporary
orders to require periodic verification of system alignment by control room
staff.
The team identified discrepancies in the FSAR, Technical Specifications,
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design basis documents, calculations, drawings and other documents and
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inconsistencies in these documents.
The licensee had not, in some cases,
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performed safety evaluations for these FSAR discrepancies to verify that no
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unreviewed safety questions were involved or had not updated the FSAR.
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Although many calculations reviewed by the team were satisfactory, the team
identified some design control weaknesses in the performance of calculations
and in the control of calculations.
Because of the team's concerns regarding
the use of several nonconservative inputs and assumptions in the analysis for
switchover of DHRS pump suction from the borated water storage tank (BWST) to
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the reactor building sump under post-accident conditions, the licensee
evaluated the condition and concluded that the system was inoperable, notified
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the NRC in accordance with 10CFR 50.72 on December 21, 1996, and revised
operating procedures to resolve the issue. The ' licensee also issued a
licensee event report (LER 96-002) on January 20; 1997.
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In a memorandum the licensee performed calculations that provided the bases
for concluding that the check valves in the DHRS pump suction line from the
BWST need not be tested in the closed position as part of the inservice test
program. The team identified that the analysis in the memorandum assumed
nonconservative post-accident reactor building pressures.
The licensee
evr.iuated the analysis and issued interim operability restrictions to preclude
-cecurrence of conditions after an accident that would challenge the leak
tightness of these check valves.
~ ~ The-team identified that memoranda, technical data reports, and plant
engineering evaluation requests were used to perform calculations.
The
calcula'tions did not follow the administrative and quality assurance
requirements in applicable engineering procedures and had not been
appropriately verified, approved, and controlled. Also, some calculations
appeared to perform the same or very similar analysis and those that were not
currently valid were not identified as superseded.
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Other design control issues identified by the team included inconsistencies in
the calculated instrument loop errors and in loop. error values used in
surveillance procedure instrument data sheets, and. lack of verification and
- approval of field sketches for BWST level transmitters that were used for.
~ ' static head correction .
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The licensee performed a 10 CFR 50.59 safety evaluation for a revision to FSAR
section 6.4.2 to include tne revised flow and NPSH values for the DHRS and
reactor building spray (BS) pumps. The calculation in support of this FSAR
revision showed that considering instrument inaccuracies and operating
procedure limits on system flow, the available pump NPSH did not meet the
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required value unless the containment overpressure was taken into account.
However, the NRC safety evaluation report (SER) on this topic assumed no
containment overpressure.
The licensee's safety evaluation did not identify
that the potential for an unreviewed safety question was involved.
The licensee performed safety system functional inspections of both the
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systems during 1992.
The team selected recommendations and discrepancies
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, documented ir the self-assessment reports for both systems and verified
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whether corrective actions were completed.
For the decay heat removal system,
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almost one-half of the items reviewed by the team had either not been acted
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upon or actions initiated were not timely.
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The team identified that molded case circuit breakers in two safety related
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motor control centers (MCCs) for ventilation systems'used during refueling had
not been periodically tested since their installation in 1986 because they
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were not included in the general maintenance database until 1993. They are
now scheduled to be tested in 1997 and 1998. ,Also, in 1993, the testing
frequency for a feeder circuit breaker in a safety-related MCC was deleted
from the database.
The licensee implemented appropriate measures to resolve the immediate
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concerns identified by the team.
For the other issues, the licensee initiated
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appropriate reviews and corrective actions, such as revision of design
documents, changes to procedures, and further evaluations of the identified
. issues.
Notwithstanding the weaknesses described above, the team concluded
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that there was evidence that the reviewed systems generally adhered to the
design and licensing bases and the as-built configuration was consistent with
the FSAR.
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El Conant of Engineering
El . T Inspection Scote and Methodolooy
The purpose of the inspection was to evaluate the capability of the selected
systems to perform safety functions required by their design bases, the
adherence to the design and licensing bases, and the consistency of the as-
built configuration with the final safety analysis report (FSAR). The systems
selected for inspection were the Makeup and Purification (MU&P) system and the
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Decay Heat Removal System (DHRS).
These systems were selected on the basis of
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their importance in mitigating design basis accidents at TMI-1.
The
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inspection focused on the engineered safeguards functions of these systems and
the interfaces with other systems.
The inspection was performed in accordance with the applicable portions of
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Inspection Procedure (IP) 93801, " Safety System Functional Inspection." The
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engineering design and configuration control section of the IP was the primary
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focus of the inspection.
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The open items resulting from this inspection are included in Appendix A.
Documents reviewed by the team are listed in Appendix C.
The acronyms used in
this report are listed in Appendix D.
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El.2 iiakeup and Purification System
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E1.2.1 System Description and Safety Functions
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During normal plant operations, the MU&P system provides makeup water and
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controls water quality and boron concentration of the Reactor Coolant System
(RCS), as well as seal coolant water for the RCS pumps.
During potential
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accident conditions, p' rtions of this system function in the high pressure
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injection (HPI) mode to pump borated water to the RCS for the small and
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= & intermediate break spectrum of loss of coolant accidents (LOCAs).
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'The components of the MU&P system which provide the HPI function include the
three high pressure injection pumps, the makeup tank, and a network of piping
and valves that provide injection of water into the discharge line of each
reactor' coolant pump.
The system is automatically actuated by signals from
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the engineered safeguards actuation system (ESAS) when a low RCS pressure
-(1600 psig or less) or.high reactor building pressure (4 psig or more).is
reached. Once the system is actuated, the injection valves to the RCS are
opened, the pumps are energized,'and stop check isolation valves connecting
the pump suction header of the HPI pumps with the Borated Water Storage Tank
(BWST) are opened so that' borated water is delivered to the RCS.
The HPI.
system along with the DHRS perform the emergency core cooling system (ECCS)
+ functions at THI-1.
In the event long term operation in the HPI mode is necessary, the design
provides for recirculation of water from the reactor building (RB) sump
through the RCS. This mode of operation, termed " piggyback" mode, is manually
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" aligned when the BWST is nearly depleted and the RCS pressure is higher thaa
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-the capability of the DHRS pumps.
In this mode, the operator realigns the
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systems to allow the DHRS pumps to transfer the sump water to the HPI pumps
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which then inject the water into the RCS.
El.2.2 Mechanical Design Review
El.2.2.1 Scope of Review
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The mechanical system design review consisted of a review of basic system
design and associated plant documents.
The. reviewed documents included:
sections of the Final Safety Analysis Report (FSAR), Technical Specifications
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(TSs),17 plant procedures, 26 system calculations, 6 technical assessment
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reports, the HPI and DH system design basis documents (SDBDs), system piping
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and arrangement drawings, flow diagrams, applicable sections of the training
manual, 6 modification packages, and the inservice testing (IST) program for
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this system.
In addition, several walkdowns of the system and interviews with
control room personnel were conducted.
E1.2.2.2 Findings
The system design documents reviewed by the team adequately supported the
design bases, except for the open items discussed in the following paragraphs.
a.
Effects of letdown line Break in the Auxiliary Buildina
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FSAR Appendix 14A, " Design Review for Consideration of Effects of Piping
System Breaks Outside Containment," Section 4.3, " Auxiliary, Fuel, and
Intermediate Area (Personnel Access) Buildings (Auxiliary Area)," described
analyses of full size breaks of the letdown line.in the auxiliary building and
concluded that, except for potential damage to cabling due to pipe whip, the
valves and cabling have been qualified for post-accident containment service
which was more severe than that resulting from the postulated break.
This
analysis was in. response to an Atomic Energy Commission (now NRC) request at a
meeting held June 28, 1973.
No approved document in support of this
(evaluation could be retrieved by the licensee.
(The l'icensee was able to retrieve an unverified and unapproved scoping
ca1culation dated June 27, 1973, of the estimated subcompartment pressure due
to.a letdown line break. This calculation did not provide an assessment of
- the potential local pressures and temperatures that could result due to a
letdown line break.
Each HPI pump as well as the makeup tank are well separated and locate'd in
. separate cubicles at E1.281' of the auxiliary building.
The suction and
discharge lines to the pumps are located in a compact valve gallery outside
the pump and tank cubicles.
Outside this gallery the suction piping is routed
in the auxiliary building to the BWST tank and the discharge piping to the
containment. To provide a degree of separation the injection piping from HPI
pump P-1C is routed from the valve gallery to the next elevation (El.305')'.
However, the two trains of pump suction piping from BWST are routed close to
each other outside the valve gallery and the two trains with stop check valves
join a common header from the BWST. The decay heat removal system piping to
.the HPI pump suction for " piggyback" operation is also located in this area.
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The high pressure (>2100 psig) portion of the letdown line is routed to the
entrance of the valve gallery which is a short_ distance away.
The' team was concerned about the adequacy of train separation in the above
arrangement because of the compactness of the valve gallery and the small
common area outside the valve gallery where safety related components of the
suction side of both trains were located.
Adequate environmental separation
of both HPI pump discharge lines appeared not to have been provided since
there are no environmental enclosures and an open stairway connects the areas
where the injection lines are located at EL 281' and EL 305'.
The team's
concern was whether a common harsh environment could disable both HPI
injection trains.
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The team noted that containment isolation valves MU-V-16A&B were in the
vicinity of the letdown line and both isolation valves were about 15-20 feet
from the containment structure. A number of issues were discussed with the
licensee on this aspect of the arrangement including:
the potential impact of
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a letdown line failure on the adjacent HPI injection lines: the ability of the
letdown isolation valve to close considering the dyn'amic effects of a pipe
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failure; the impact of letdown line break on safety related equipment,
particularly in areas where two trains of safety related components, such as
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valves MU-V-14A&B and DhiV-7A&B, are located; the impact of an open stairwell
(between EL 281' and 305') interconnecting the atmospheres associated with '
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both safety trains of the HPI; and the letdown break flow temperature increase
because of ineffective letdown cooling due to failure of the intermediate
cooling system or due to excessive letdown flow resulting from a break.
The team questioned whether the pipe whip and jet impingement effects of a
letdown line break in the auxiliary building would impair operation of the
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adjacent HPI injection lines or injection valves.
Also, the team questioned
whether the letdown isolation valve outside containment would be able to
operate to isolate the line after such a postulated break assuming that power
'is not available for the letdown line isolation valves inside containment due
to a single failure of the power source.
The licensee stated that Supplement 2, Part XI, Amendment 48 to the FSAR
indicated that the HPI containment isolation valves used the same motor
operators as were used inside the containment that were qualified for post-
accident containment environment. However, the motor insulation used for some
of the valves in the auxiliary building was specified as class B rather then
the higher temperature rating of class H which was used for safety-related
valves wi. thin the containment.
Although Appendix 14A of the FSAR discusses the effects of full breaks in .the
letdown line, Appendix 6B of the FSAR states that a LOCA was not postulated to
occur outside the containment for environmental qualification of components.
The licensee stated that the requirements to provide environmental
qualification of equipment subject to high energy line breaks (HELBs) under 10 CFR 50.49 had superseded the AEC request for additional information on the
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letdown line break because the HELB criteria covered only lines normally
operating at both greater than 275 psig and 200 deg F.
Lines meeting only one
of these criteria, such as the letdown line,. which was normally below 200 deg
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F, were required to be evaluated only for crack type breaks at the most
adverse locations.
The licensee could not provide an analysis of the
consequences of a crack in this line.
The master list that specifies EQ requirements for safety-related electrical
equipment indicated that, other than radiation, no harsh EQ requirements were
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specified for the portions of the HPI system outside the reactor building and
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within the auxiliary building.
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The licensee issued PFU 98-TI-120 to process a revision to FSAR Section 4.3 of
Appendix 14A to state that a postulated full diameter letdown line break
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outside the containment is not considered a high energy line break since the
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normal operating conditions do not meet the high energy criteria for the
purposes of EQ of safety-related electrical equipment.
The licensee also
issued a technical functions assigned action item (TFAA1) BT6560 to complete
the appropriate documentation and perform design verification of an evaluation
of the assumed letdown line break in the auxiliary building.
The team referred this issue to the technical review branch in the Office of
Nuclear Reactor Regulation (NRR) staff for review regarding the extent to
which TMI-1 was required to consider the effects of a letdown line break in
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the auxiliary building.
The staff review concluded that the THI-1 licensing
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basis for pipe breaks includes' the postulation of full diameter breaks in the
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letdown line between the containment penetration and the breakdown orifice as
described in Appendix 14A to the FSAR. Therefore, the design of safety-
related equipment in the affected areas should consider the conditions
resulting from these breaks.
(Inspection Follow-up Item 50-289/96-201-01).
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b.
Makeuo Pumo MU-p-1C Confiouration and Operation
During the normal standby mode the suction valve from the BWST and high
pressure injection valve for makeup pump MU-P-lC are in closed position, and
the pump is isolated from the suction and discharge headers that are connected
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.. to makeup pumps MU-P-1A&B and the makeup tank.
The makeup pumps and the
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valves receive an actuation signal at the same time. No interlocks or time
~ delays are provided to start the pump after the suction valve has opened.
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~pump will achieve full' speed within 6 seconds and opening of the suction line
. stop check valve is expected to be less than 13 seconds although the design
basis opening time for the suction valve is 22 seconds. The high pressure
injection' valve in the pump discharge line,is estimated to open fully ' suction
in 10
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seconds.
Inservice testing (IST) for this pump.is done with the pump
header cross-connect valves -ope.n so that the makeup tank is connected to the
pump. The licensee stated that only during a prepperational test in 1974 the
pump was tested with the suction header cross-connect valves closed.
The
licensee also stated that during the 1974 test the suction valve opened in
about 9 seconds. The team was concerned that the effect on the pump due to a
slow opening suction valve combined with a rapid s' tart of the pump and a fast
opening high pressure injection valve had not been analyzed.
The licensee
issued engineering work request (EWR) 786380 and TFAAI BT 6556.to conduct
further assessments and to revise the design basis stroke time of the suction
line stop check valve to 13 seconds.
(Inspec' tion Follow-up Item 50-289/96-
201-02)
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The team identified a potential for accumulation of noncondensibles, such as
hydrogen, released from the stagnant water in the suction line because of the
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physical configuration of the line. The suction piping from the stop check
valv'e to the pump is approximately 140 feet long with a long horizontal run
approximately 10 feet above the puinp suction.
Other than quarterly cycling of
the stop check valve and ensuring that the piping is filled and vented after
any maintenance work that required draining of the piping, no monitoring for
gas accumulation is performed. During the inspection, the licensee opened the
vent valve in the high point of this pipe section and verified that there was
no gas accumulation in the suction pipe.
The licensee stated that additional
monitoring of the MU-P-1C suction pressure would be instituted to ensure that
this line remains full during this operating cycle and revised the primary
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auxiliary operator logs accordingly. The team considered that a positive pump
suction pressure was not necessarily an indication of absence of gas
accumulation in the piping. The licensee issued EWR 786469 to develop a long
term solution to ensure that the line remained full of water.
(Inspection
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Follow-up Item 50-289/96-201-03)
c.
Makeuo Tank Operatina Limits
The hydrogen gas pressure and water level in makeup tank MU-T-1 are controlled
and monitored during plant operations.
Tank pressure versus level curves had
been established in procedures. OP 1104-2, "TMI-1 Operating Procedure Makeup &
Purification System," and 1101-1, " Plant Limits and Precautions," so that
operation of the tank could be controlled between two curves.
The upper curve
referred to as the " gas entrainment curve" provided th'e upper limit for
operating pressure and level to prevent gas entrainment and potential for gas
binding of the makeup pump. The analysis supporting this curve was provided
in calculation C-1101-211-5310-0047, Revision 0.
The team identified the
following concerns with this calculaden: tank volume / height relationship used
in the analysis was.not consistent with the tank characteristics provided in
procedure 0P 1101-1; the configuration of the piping from the stop check valve
V-14A to the pump suction header used in the calculation did not include two
elbows as well as some piping; and nonconservative maximum HPI pump flow rates
were used in the calculation.
The team was also concerned that the effects of
"any vortexing within the makeup tank had not been analyzed,. the appropriate
maximum BWST temperature had not been used, and the conservative minimum BWST
level was not considered.
The licensee issued EWR 786342 to revise procedure
~OP 1101-1, TFAAI BT6542 to address makeup tank vortexing, and TFAAI BT6554 to
revise the calculation.
The lower pressure / level curve is the "two pump NPSH" curve that specified the
limit abo've which the makeup tank pressure and level must be maintained to
assure that adequate net positive suction head (NPSH) is maintained for two
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pump operation during a loss of coolant accident (LOCA) condition.
The team
reviewed calculation C-1101-211-5360-003, Revision 1, that included the
supporting analysis for this curve and identified a concern that a
conservatively high limiting flow for two pump operation following an
injection line break was not used in the analysis.
In addition, this
calculation used inputs from calculation C-1101-211-5310-0047, Revision 0,
discussed in the previous paragraph. Therefore, the two pump NPSH curve in
the procedure could be incorrect. The team noted that,the requirements in 10
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CFR Appendix B, Criterion III and GPU Nuclear Operational Quality Assurance
Plan, Section 4.0, " Design Control," regarding measures for correctly
translating the design basis into procedures and instructions were not
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adequately established.
In response to the team's concern the licensee's
plant review group (PRG) discussed this issue on January 21, 1997, and
concluded that the system was operable but degraded.
The licensee initiated
procedure change requests (PCRs) 1-0S-97-0037 and 0038 to revise procedures'to
limit makeup flow and issued TFAAI BT6559 to revise the calculation.
(Unresolved Item 50-289/96-201-04)
l
d.
ECCS Actuation Time
i
FSAR Table 14.2-14 states that the ECCS delay time assumed in the LOCA
accident analysis is 35 seconds.
The licensee stated that this delay time is
composed of I second for instrumentation lag, 10 seconds for start of the
emergency power source if offsite power is not available, and 24 seconds for
'
system response (pump acceleration and valve stroke time).
FSAR Section
6.1.3.1 states that the system is designed to be in. full operation within 25
<
'
seconds after receiving an actuation signal, and surveillance procedure 1300-
3H, Revision 44, provides a design basis stroke time of 25 seconds for the
makeup pump recirculation isolation valves MU-V-36&37 and the injection
isolation valves MU-V-16A through D.
The team observed that the 25 second -
startup delay time added to the 11 second delay of the actuation signal
results in a total delay time of 36 seconds, which would be an unanalyzed
,
condition as stated in Section 4.2.2 of SDBD-T1-211, Revision 1.
This change
j
-
to the facility had not apparently been reviewed in accordance with 10 CFR 50.59. The licensee issued EWR 786255 to revise the design basis stroke
timesin procedure 1300-3H and issued plant FSAR update (PFU) 98-TI-124 to
revise the FSAR.
(Unresolved Item 50-289/96-201-05).
l
.e.
Additional Onen Items
=The following issues were discussed with the licensee but were not resolved
before the end of the'on-site phase of the inspection, and therefore, will be
l
followed up later:
The consequences of failure of non safety-related auxiliary steam
.
)
piping in the auxiliary building.
Failure of the auxiliary steam
-
piping due to a seismic event could result in degradation of safety
related equipment classified for mild environment. (Inspection
Follow-up Item 50-289/96-201-06)
,
The impact of loss of pressure in the makeup tank on the net positive
suction head for the HPI pumps due to letdown line break'or crack
combined with the failure of the check valve in the line to the
makeup tank.
(Inspection Follow-up Item 50-289/96-201-07)
The available net positive suction head for the HPI pumps when taking
=
suction from the BWST at low-low level or during potential vacuum
conditions in the makeup tank was not reviewed by the team because
i
calculations were not available for review before the end of the
inspection.
(Inspection Follow-up Item 50-289/96-201-08)
6
-
.
8
El.2.2.3 Conclusions
j
The SDBD for the system and the system modification documents were detailed
and Sell written, and were consistent with the FSAR.
The mechanical
modifications that were reviewed were appropriate for resolving the identified
problems and the modifications did not change the design bases of the system.
Although many calculations reviewed by the team were satisfactory, there were
weaknesses in calculations relating to the makeup tanks and pumps.
The
i
analyses supporting the MU&P tank pressure / level operating limit curves used
nonconservative inputs. The licensee imposed temporary limits on HPI flow
during post-accident conditions to resolve this issue pending revisions to the
analyses.
The simultaneous starting of makeup pump IC and opening of its
suction and discharge valves needed verification to ensure that the pump
operation is not adversely affected. Also measures are required to provide
'
assurance that the pump suction line would not get gas bound due to release of
noncondensibles in the water.
The acceptability of the environmental effects and pipe whip due to a letdown
'
line break in the auxiliary building was referred to the appropriate NRR
technical branch.
The NRR staff position was that the licensing basis
included consideration of consequences of full diameter breaks in the letdown
line between the containment penetration and the breakdown orifice as
described in Appendix 14A to the FSAR.
E1.2.3 Electrical Design Review
-
E1.2.3.1
Scope of Review
The team reviewed the electrical design for normal and emergency operation of
the MU&P/HPI pump motors, selected motor operated valves (MOVs) and air
operated valves, circuit breakers, fuses and interlocks. The team also
compared the FSAR descriptions with the SDBD, drawings, Technical
Specifications (TSs), operating procedures, and test procedures in order to
verify consistency in the documents.
In addition, the team reviewed the
calculations related to voltage drop, electrical loading, and coordination for
selected MU&P system components and associated electrical distribution system
components to determine the adequacy of the available voltages, equipment
loading, protective system coordination, and electrical isolation and
independence.
The team also reviewed four electrical modifications.
E1.2.3.2 Findings
,
The team verified that each train of the MU&P system was powered from i.
separate emergency power bus and that the electrical loading of the individual
components had been considered in the emergency diesel generator (EDG)
capacity calculations. The sequence and timing of loading of HPI pumps and '
valves onto the EDG was consistent with the FSAR.
'7
-
-
.
.__
_.
_
The system design. documents reviewed by the team adequately supported the
design, except for the open items discussed in the following paragraphs.
'125V DC System Voltaae Dron Analyses
a.
Calculation C-Il01-734-5350-004, "TMI-l DC System Calculation," Revision 1,
provides individual DC circuit voltage drop analyses for the redundant station
battery circuits.
The circuit for makeup isolation valve MU-V-18 was listed
in the calculation, but the voltage drop analysis for this valve was not
i
!
performed. The licensee prepared a preliminary analysis during the inspection
!
that demonstrated the adequacy of the available voltage at MU-V-18.
Numerous
l
DC circuits for other systems were listed in the calculation but the voltage
i
drops for these circuits were not calculated due to unknown data.
This
l
calculation also concluded that further investigation was required to
determine the adequacy cf the terminal voP. age at various DC equipment. The
licensee issued TFAAI BT6545 to update the calculation.
(Inspection Follow-up
Item 50-289/96-201-09).
l
b.
Makeup I' solation Valve MU-V-18 Control
FSAR section 6.1.3.1 stated that in the event of a small break LOCA a 70/30
t
HPI flow split. (70% to the reactor core, 30% out of the break) will assure
adequate core cooling.
This 70/30 flow split is ensured by closing air
operated valve MU-V-18, which is a safety-related valve that automatically
isolates on an engineered safeguards (ES) actuation signal, thus isolating the
makeup line flow to HPI leg "B" downstream of the cavitating venturi.
Valve
l
MU-V-18 is powered from 125 VDC distribution panel IM, which in turn is
supplied from either of the redundant 125 VDC main distribution panels lA or
IB via an automatic transfer switch.
This switch is locked out when an ES
signal is received.
The design of the control scheme for valve MU-V-18 is
'
l
such that a loss-of 125 VDC power to the valve would not close the valve if
I
control air was available. The valve can be remote manually closed if both
'
. power and control air are available. The FSAR analysis takes credit for
,
i
~ : partial isolation of the makeup flow if the operator closes the non safety-
'
- related makeup control valve MU-V-17.
The licensee stated that although MU-V-
l
-17 was not a safety-related component it was included in the inservice test
l
program, and its operational status was monitored continuously because the
l
valve was used for controlling the makeup flow during normal operation.
The
,' team questioned the ability of the system to maintain required injection flows
in the event of _an injection line break downstream of the last check valve.
The licensee stated that the loss of IA or IB 125 VDC power concurrent with an
,
HPI line break downstream of the last check valve on the injection line would
l
prevent the HPI system from performing its design ECCS function'if the power
source lineup and the valve positions in the cross-connect line'between pumps
A & B were not controlled.
If the system is not properly aligned, the loss of
a 125 VDC power train could result in failure of HU-V-18 to close, thus,
degrading one HPI train and the other HPI train would be inoperable due to the
loss of the 125 VDC' breaker control power from the same source.
The selection
of the A or the B 125 VDC power source for the IM 125 VDC distribution panel
must be based on the position of the valves in the makeup pump discharge
,
8
.
,
-
i
I
i
i
cross-connect piping.
A proper lineup will ensure that a 125 VDC system train
'
failure will not effect both HPI trains.
The licensee stated that this
problem was discovered before the inspection team arrived at the site.
The team asked the licensee to verify the current alignment in the plant.
The
licensee reported on January 8,1997, that the electri. cal power alignment was
,
incorrect and it was immediately corrected.
The licensee also issued
temporary orders to control room staff to perform checks of the cross connect
valve positions and DC power source alignment.every shift.
The licensee
issued PFU 98-T1-Il9 to revise the FSAR and PCRs 1-0S-96-0573 and 1-0S-96-0574
to revise operating procedures OP 1107-2 and OP 1104-2 respectively. The team
noted that the requirements in 10 CFR 50, Appendix B, Criterion V,
" Instructions, Procedures, and Drawings" regarding prescribing activities
affecting quality by documented instructions, procedures or drawings and
accomplishing the activities in accordance with these instructions,
procedures, or drawings were not apparently met. (Unresolved Item 50-289/96-
201-10)
'
'
c.
MCC Control Circuit Voltaae Dron Analyses
In the Technical Data Report (TDR) 995, "TMI-l Voltage Drop Study On Degraded
Grid Condition," Revision 3; the licensee assumed proper operation of
engineered safeguards motor starters based on the minimum voltage available at
the engineered safeguards MCC buses.
Section 1.3.1 of. System Description,
" Class IE Electrical Systems, Metropolitan Edison Company, Three Mile Island
Unit 1," dated March 23, 1970, states that the variation of voltage and
frequency in the system during any design basis event would not degrade the
performance of any load. On the basis of these statements it appeared that
i
voltage drops were considered in the original design.
However, because no
analyses to confirm these statements could be retrieved, the team was not able
to verify whether control circuits for both MU&P system and DHRS with long
. cable runs were analyzed to assure that the available voltages at the
equipment were acceptable.
The licensee had identified this issue during the
review of the TDR in January, 1996 and issued TFAAI BT6493 to perform detailed
analyses.
.
El.2.3.3 Conclusions
The electrical design for components that perform the engineered safeguards
functions of the MU&P system was ade_quate. Although the team did not have any
concerns with the available voltages at components selected for review (the
licensee demonstrated that the available DC voltage at the solenoid for MU-V-
18 was acceptable), there was no documented analyses to demonstrate that the
worst case electrical configuration had been evaluated and found acceptable.
Until recently the licensee had not evaluated the importance of proper
alignment of power source to valve MU-V-18.
.9
E1.2.4
Instrumentation and Control Design Review
El . 2'. 4 .1
Scope of Review
The instrumentation and control design review consisted of an assessment of
applicable sections of the FSAR, SDBD, and TS for the MU&P system.
These
'
documents were reviewed for consistency with emphasis on design bases for
instrument loops and associated control logic. Applicable component loop
analysis, calibration methodology, setpoint calculations, and compliance with
single failure criterion were also reviewed.
In addition, the team reviewed
calculations related to the replacement of Regulatory Guide (RG) 1.97
instruments. A total of 9 calculations and 2 modifications were reviewed.
El.2.4.2 Findings
The instrumentation and control requirements for the HPI funct. ions of the
system, such as measurement and indication of HPI flow and pressure and makeup
tank ievel and pressure were adequate.
The display of measured parameters and
provision of controls for pumps and valves in the control room were
appropriate for the system function.
The instrument setpoints for the makeup
tank and HPI flow were acceptable.
The system design documents reviewed by the team adequately supported the
design, e~xcept for the open item discussed in the following paragraphs.
,
'
a.
Make-un Tank level Instrument looo Error
,
The team reviewed the instrument data sheets in surveillance procedure 1302-
5.17, Revision 17, "Make-up Tank Level Instrumentation." The data sheet for
instrument loop MV14-LT specified a loop error (tolerance) of 1%.
However,
calculation C-1101-662-5350-049, Revision 0, estimated a loop error of 1.23%.
Although the loop error in the data sheet was more restrictive, the team
questioned the inconsistency in the two documents.
The licensee stated that
for all safety-related instrument loops a conservative loop error of 1% was
assumed.
Documentation justifying this assumption could not be retrieved.
'After further review of surveillance data sheets and calculations, the team
~ identified three other instances of inconsistency in the documents.
The data
sheets for control room indicator instrument loop MU-LI-778A and computer
point A0498 instrument loop MU14-LT specified a loop error of 11% each, but
calculation C-Il01-662-5350-049, Revision 0, determined that the loop errors
should be' limited to
0.64% and i 0.73% respectively.
Calculation C-1101-
624-5350-002, "Make-up Tank Level Error for Accident Conditions (MU-14-LT
Loop)," Revision 1, determined that the loop error should be i 0.57% for MU14-
LR in instrument loop MU14-LT, but the data sheet specified a loop error of
1.0.
The calculated allowable loop errors were more restrictive than the
instrument loop data sheets. The licensee issued TFAAI BT6544 to resolve the
team's concerns.
(Unresolved Item 50-289/96-201-11)
l
l
.
10
_
.
_
_
__ _
.
_
_
_. _
_
_
_ _ _
E1.2.4.3 Conclusions
'
The instrumentation and control design for the engineered safeguards function
of the MU&P system was adequate.
The team was concerned that surveillance
'
procedure data sheets that are used to calibrate the makeup tank
instrumentation, in some cases, did not use the more restrictive calculated
tolerances.
E1.2.5 System Interfaces
The team selected the following systems that interface with the MU&P system
.
and verified that the interface system design information for supporting the
function of MU&P system were appropriately considered: DHRS which provides
water from BWST during the HPI mode operation; reactor coolant system;
auxiliary building which houses the makeup system; and nuclear services'and
decay heat closed cycle cooling water systems which provide cooling water to
the lube oil coolers and motor coolers.
i
In addition to reviewing the interface design information for the above
systems, the team performed a walkdown of the instal _lation of the interfaces.
The team did not identify any concerns and concluded that the' design of the
interfaces were satisfactory.
E1.2.6 . System Walkdown
>
E1.2.6.1 Scope of Review
The system walkdown included examinations of the MU&P system piping and
-mechanical components within the auxiliary building, interface piping with the
.DHRS within the auxiliary building, installation of' instrumentation and
electrical components, and verification of consistency of selected portions of
-
- the system with plant drawings. The walkdown also included interviews with
plant operators in the control room and an examination of the instrumentation,
used to monitor the operating status of the MU&P tank conditions.
.
E1.2.6.2 Findings
he team identified the following discrepancies between the MU&P system flow
diagram 302-661, Revision 47, and the installed system:
Restricting orifices in pump recirculation lines were shown located
outside each pump room rather than within each room.
,
The symbol for venturi flow elements is illustrated backwards.
.
The strainer elements in the pump suction strainers had been removed
.
and this was not indicated on the drawing.
.
Instrument' MU14-LT and MU&P pump suction and discharge pressure
.
'
gauges were not located inside the tank / pump rooms as shown in the
i
flow diagram.
-
,
.
-
11
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__
-
.
. - _ _ _
_ _ _ _ _
_ __. _
_ _
The team also identified that the logic for alarm A78 for isolation valves DH-
V-1 & 2 was not included on drawing SS-209-887, Revision 2.
5
The licensee issued ECD C204671 and ECD C205641 to revise the affected
drawings.
,
The team identified during walkdown that engineering calculations for HPI
system performance had omitted two 90 degree elbows as well as about 10 feet
of piping in the line segment between the stop check valve and the suction
i
header for HPI pumps MU-P-1A and IB (see Section E1.2.2.2.c. of this report).
j
E1.2.6.3 Conclusions
The reviewed system drawings were ' consistent with the as-built system except
'
as noted above.
The plant equipment was generally well maintained and good housekeeping was
evident.
,
'
E1.2.7 FSAR and SDBD Review
The team reviewed the appropriate FSAR sections and SDBDs for the MU&P system
-
i
and the associated electrica'l and instrumentation and control systems.
The team identified the following discrepancies in the FSAR:
FSAR Table 14.2-18, Sheet 1, indicated that " total flow" for an HPI
line break.at an RCS pressure of 1800 psig is 347.5 gpm instead of
397.5 gpm.
FSAR Section 14.2.2.4.3.a indicated that an open issue existed with
=
regard to tripping the reactor coolant pumps during a SBLOCA but did
not state that a manual trip on a loss of subcooling margin was
acceptable.
FSAR Table 9.1-2 and the GMS-2 database contain conflicting design
. _ . . _
data for MU&P system components.
.
FSAR Section 8.2.3.1 b describes the Emergency Diesel Generator
.-
continuous rating as 2600kw instead of the correct rating of 2750kw
stated in the vendor manual, VM-TM-0191, Revision 29.
,
The above discrepancies had not been corrected and the FSAR updated to assure
that the information included in the FSAR contained the latest material as
required by 10 CFR 50.71(e).
(Unresolved Item 50-289/96-201-12)
- The licensee issued PFU 98-T1-113, PFU 98-T1-123, PFU-98-T1-127, and PFU-98-
T1-135 to revise the FSAR.
-
12
.
,4
e
_.
_
'
,
I
The team identified the following discrepancies with SDBD T1-211, Revision 1:
-
/
Section 4.1.3 (H) stated that train separation in the valve room was
-
8 feet. However, some valves and piping components are closer than 8
feet.
Reference 2.1.44 cited TDR 114, Revision 1.
However, TDR 114 has
-
been superseded by TDR 995.
,
El.3 Decay Heat Removal System
4
j
El.3.1
System Description and Safety Function
j
L
The engineered safeguards function of the Decay Heat Removal System (DHRS),in
l
conjunction with the Core Flooding System, is to provide inventory makeup at
intermediate to low RCS pressures and to ensure adequate core cooling for RCS
break sizes ranging from intermediate breaks to the . double-ended rupture of
the largest pipe.
The engineered safeguards function is referred to as low
'
pressure injection (LPI).
The DHRS also provides long term core cooling after
a LOCA. The system performs these functions by injecting borated water from
.
the BWST to the RCS, and by recirculating, cooling, and re-injecting reactor
!
The system is automatically actuated by signals
'
from the ESAS when a low RCS pressure or a high RB pressure is reached.
The
pumps start and operate in the bypass mode until the RCS pressure decreases
below the pump discharge pressure.
Valves are automatically actuated to
a
provide an injection path using borated water from the BWST. When the level
in the BWST reaches a low level, DHRS pump suction is manually transferred to
When the RCS is not depressurized sufficiently for LPI injection, the MUiP/HPI
,
pumps take suction from the LPI discharge for " piggy-back" operation during
--long term. cooling.
The Decay Heat Closed Cooling Water System provides
cooling water for the decay heat coolers, DHRS pumps and motors during
emergency operation.
E1.3.2 Mechanical Design' Review
.
i
E1.3.2.1 Scope of Review
i
- The mechanical system design review of the DHRS included the review of the
following. design and licensing documents: sections of FSAR, Technical
Specifications, 16 calculations, design basis document SDBD-TI-212, 7 system
modifications, 4 operating procedures, system flow diagrams, and other related
documents.
In addition, walkdown of the system.and interviews with control
room personnel were conducted.
Each of the selected modifications was reviewed t'o determine its impact on the
design and licensing basis, and to verify that the safety evaluation was
correct and consistent with the applicable procedures, the post-modification
testing was complete and appropriate, and the required plant documentation had
been updated to reflect the modification.
.
13
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.
e
-.
.- -
.
_
-
-
...
.
El.3.2.2 Findings
The, team verified that the design flow of each train of the DHRS was
consistent with the low pressure injection (LPI) flow required by the
'
emergency core cooling system (ECCS) analysis for a loss of coolant accident
(LOCA). The team noted the design features for " piggyback" operation of the
<
LPI and HPI system were in accordance with the functional requirements.
The'
i
.DHRS pump performance characteristics and the BWST volume storage requirements
were adequate.
The safety classifications of the system components specified
in the SDBD were appropriate and the applicable codes and standards were
-
specified.
The design pressures and temperatures specified for piping and
,
'
7
components were adequate. The system design basis document (SDBD) and flow
diagram were consistent with the FSAR descriptions of the system.
.
The system design documents reviewed by the team adequately supported the
design, except for the open items ~ discussed in the following paragraphs'.
a.
Switchover of DHRS Pumo Suction from the BWST to Reactor Buildina Sumo
Calculation C-1101-212-5310-050, "TMI-1 BWST Vortex Determination," Revision
0, dated June 1, 1994, evaluated the DHRS pump suction transition from the
s
BWST to the reactor building (RB) sump under post-accident conditions. The
l
'
i
team reviewed this calculation in detail and identified the following
i
concerns:
,
1
. Sectin 3.3 of the calculation stated that the BWST low-low level alarm
setpoint at which the switch over operations were to be initiated was
i
- '
at 6 feet-4 inches (6.3 feet).
This value did not account for
instrument uncertainties that could result in the alarm actuating at a
,
lower level. The Instrument Loop Data Sheet (1302-5.19, Sheet 2 of 7)
~
indicated that the minimum acceptable "as found" alarm level was at 5
feet-9 inches. The actual alarm level could be lower due to other
instrument errors.
. Section 3.6 of the calculation stated that the reactor building'
'
pressure was 21.62 psia prior to the switchover operation. A high
reactor building pressure reduces the flow from the BWST during the
period when the RB sump isolation valves DH-V-6A and B begin to open
and the BWST is isolated by closing valves DH-V-5A and B.
The team
noted that the reactor building pressure used in this calculation was
'
the maximum calculated value and was nonconservative for this analysis.
Depending on the accident scenario the switchover. from the'BWST could
,
take place at a lower reactor building pressure.
,
i
. Section 5.2 of the calculation used time delays of 8 seconds to 24
seconds between the BWST low-low alarm and the first operator action
for different scenarios. These time del.ay data were based on simulator
training results.
SDBD-TI-212, Section'4.8.1.17, states that the
current BWST low-low level alarm setpoint of 6.3 feet increased the
available operator response time to 30 seconds.
The team noted that
the calculation did not appear to have sufficient margin to support an
"
operator response time of 30 seconds. 'Also, the team questioned
14
i
!
l
'
whether it was conservative to use the operator action time data from a
demonstration on the simulator instead of from a LOCA test scenario
'
l
training. The licensee reviewed the data and concluded that because of
l
computer scan time the operator action times could have been
underestimated by about 25 seconds.
- The motor operated valve (MOV) stroke times used in the calculation
,
l
were less than the IST acceptance criteria values.
Future degradation
of the stroke time or modifications to the valves could cause the valve
stroke times to be longer, up to the IST acceptance criteria, and hence
the calculation would be nonconservative.
!
= The team questioned the value of a calculated constant in the equation
for required submergence of the outlet pipe in the BWST.
The constant
was outside the range recommended in the referenced technical paper,
and the BWST discharge flow rate used to calculate the constant was
incorrect.
On the basis of the above concerns the team concluded that there was a
potential for air entrainment in the DHRS pumps that could render the pumps
inoperable. The licensee p.erformed an operability evaluation of the DHRS
system on December 20 and 21, 1996, concluded that both trains of the DHRS
'
system and RB spray system (the RB spray pumps take suction from the same
piping from the BWST) were inoperable, and notified the NRC in accordance with
The licensee issued temporary change notices (TCNs) 1-96-0079,
0080, 0081, and 0082 to revise' abnormal transient procedures (ATPs) 1210-6,
. 1210-7, 1210-10, and alarm response procedure E-2-4.
The revised procedures
required the operators to begin transferring the DHRS pump suction from the
BWST to the reactor sump when the BWST level reaches the low level alarm
i
setpoint of 9 feet-6 inches.
When the BWST level reaches the low-low alarm
)
setpoint of 6 feet-4 inches the transfer would be completed by closing the'
BWST isolation valves.
In addition, several computer alarm setpoints were
revised to aid the operators in accomplishing the identified actions.
The
licensee stated that these changes were summarized in a briefing and in a
training handout'(No. 3210-96-180, dated December 24, 1996) presented to
operators. on. all five shifts.
The licensee informed the team that two demonstration scenarios were performed
.on the plant simulator. on December 23, 1996, to verify the adequacy of the
'
procedure changes. On January 20, 1997, the licensee issued LER 96-002
~
regardincj the discovery of a condition outside the design basis with respect
to accident procedures where potential air entrainment of the ECCS pumps might
have occurred during initiation of RB sump recirculation following a LOCA.
The licensee issued TFAAI No. BT 6542 to revise the calculation to resolve the
team's comments. The licensee also issued ECD C208801 to update SDBD-T1-212.
The design basis for the BWST switchover was not correctly translated into
procedures and instructions as required by 10 CFR 50, Appendix B, Criterion
III, " Design Control," and GPU Nuclear Operational QA Plan, Section 4.0,
!
" Design Control."
.
15
..
.
SDBD-T1-212, Revision 1, states that the initial setpoint for switchover
operations was 3 feet from the bottom of the BWST level sensing nozzle.
Calculation C-1101-214-5360-008, Revision 2, dated August 5, 1988, was
performed to consider a more realistic approach for the setpoint for BWST low-
low level and the calculation determined a new setpoint of 6.3 feet for the
alarm instead of 3 feet.
Technical Specifications, Section 3.3.1 states in
part, that the reactor shall not be made critical unless two decay heat
removal pumps and two reactor building spray pumps are operable.
However, for
a period of several years until December 21, 1996, the licensee apparently
operated TMI-1 outside of the design basis of'the switchover phase of the ECCS
system with potential for air entrainment of the ECCS pumps that could have
rendered them inoperable. (Unresolved Item 50-289/96-201-13)
b.
NPSH for DHRS Pumos
Calculation C-1101-212-5360-027, "NPSH Available for LPI and BS Pumps
following Large Break LOCA," Revision 0, dated October 2, 1990, addresses the
NPSH for the DHRS (LPI) pumps taking suction from the RB sump.
The team
reviewed this calculation and identified the following concerns:
In sections 6.1 through 6.9, the calculation determined that the NPSH
requirements for the DHRS pumps would not be met for an LPI flow of 3300 gpm
as indicated by flow instrumentation during post-10CA conditions (the required
'
LPI flow of 3150 gpm plus 150 gpm margin to allow for flow instrument
uncertainties) unless credit is taken for subcooling of reactor building sump
water (i.e., containment overpressure). The LPl flow was limited by operator
action to 3300 gpm as measured by flow instrumentation in accordance with
'
pro'edure ATP 1210-07.
The calculation also showed that the NPSH for the LPI
c
pumps would not be met if the measured flow was 3150 gpm taking into
consideration flow measurement inaccuracies.
The LPI flow of 3150 gpm at 0
psig reactor pressure was presented in FSAR Table 14.2-27 and was used in the
accident analysis.
As shown in Attachment _7 of the calculation, the NRC safety evaluation report
(SER) dated July 11, 1973, for THI-1, stated that the NPSH evaluation was
based on the assumption that the reactor building pressure was equal to the
vapor pressure of the sump water and was acceptable.
In this evaluation
containment overpressure was not considered.
In FSAR Update 9, dated July 1990, a statement was added to FSAR Section 6.4.2
to address sump subcooling.
In'the update it was stated that "an additional
calculation indicated that sufficient NPSH would be available for the maximum
flows, as limited by procedure, including instrument errors, recirculation
flow and taking credit for sump subcooling (containment overpressure)."
The licensee's 10 CFR 50.59 safety evaluation 115403-004, Revision 0, dated
February 20, 1990, associated with FSAR Update 9. addressed subcooling of the
reactor building sump.
As a result of removing the sodium thiosulfate tank
from the design and changing the BWST low-low level alarm setpoint, the.
calculated post-accident reactor building sump water level had decreased. The
safety evaluation recognized that the available DHRS pump NPSH would be less
than the required NPSH with a LPI flow of 3300 gpm (accounting for instrument-
16
.
e
.
_
_
_
_
_.
_
__
-
.
.
error) if no credit was taken for reactor building overpressure.
However, the
safety evaluation did not identify that, because the required NP!.H for these
pumps would not be met without taking credit for containment oveipressure, the
probability of occurrence of malfunction of these pumps previously evaluated
in the safety analysis report may be increased, and thus, a potential
unreviewed safety question was involved.
The licensee stated that the original SER evaluated the licensee submittal of
NPSH values that were based on design basis flows of 3000 gpm for LPI and 1500
gpm for reactor building spray system and did not consider instrument errors.
The licensee also stated that the LPI flow rate of 3300 gpm was a pump runout
condition, and therefore, it was beyond established design basis condition and
not part of the original SER.
The team noted that to assure a minimum flow of
3150 gpm required by the accident analysis, it was appropriate to consider in
the licensee's 10 CFR 50.59 safety evaluation that the operator would limit
the LPI flow to 3300 gpm per procedure ATP 1210-07. A measured flow of 3300
gpm could result in actual flow in the range of 3150 gpm and 3450 gpm assuming
an instrument error of 150 gpm. However, the licensee's safety evaluation
did not consider the NPSH conditions at 3300 gpm taking into consideration
instrument errors.
This change to the operation of the system could
potentially increase the probability of malfunction of the DHRS pumps due to
i
inadequate NPSH if no containment overpressure was considered.
Therefore, a
potential for unreviewed safety quest' ion as defined in 10 CFR 50.59 was
involved.
(Unresolved Item 50-289/96-201-14).
c.
Technical Specifications
The team reviewed THI-l Technical Specifications related to the DHRS to verify
that information related to the system was consistent with the design basis
and other documents. The team identified the following errors and
,
inconsistencies:
+
- -Technical Specification 3.3.1.1.f stated: "The two reactor building
=
~
sump isolation valves (DH-V-6A/B) shall be either manually or remotely
~ ~ ~
operable." This requirement was not appropriate because only remote
1
operation of the valves would support the design basis assumptions for
'
--
the-transition time of the DH pump suction switchover from the BWST to
the RB sump during post-accident conditions.
In addition, manual
operation of these valves during an accident may not be possible
because the DHRS pump rooms may not be accessible due to high radiation
conditions.
The licensee stated that THI-l had not been operated with these valves in
.
a condition where they were only manually operable.
The valve
surveillance tests required the valves to be tested for remote operation.
The licensee issued TFAAI No. BT6540 to revise the Technical
Specifications.
(Inspection Follow-up Item 50-289/96-201-15)
-
17
-
.
.
= FSAR Sections 6.4.3, 6.4.4 and Table 6.4-3 stated the design basis
leakageintheaugiliarybuildingfromtheDHRSandbuildingspray(BS)
system as 2255 cm /hr (0.6 gal /hr).
FSAR Section 14.2.2.5 and Table
'
14.2-20 documented a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> dose for engineered safeguards leakage of
2255 ml/hr of 0.037 rem.
Technical Specifications 4.5.4, " Decay Heat Removal System Leakage,"
allowed 6 gal /hr leakage from the DHRS and referred to FSAR Section 6.4.4
and Table 6.4-3.
The TS bases stated that dose was 0.39 rem from the 6
gal /hr leakage and used FSAR Section 14.2.2.5(d) as a reference for this
j
dose. There were no Technical Specifications for BS System leakage.
The
team noted that leakage control in the DHRS was more important than for
the BS system since the BS system operates for a few hours after the
'
accident while the DHRS operates over the entire duration of the
accident.
The discrepancy between the FSAR and TS was identified by the licensee as
inspection observation 212-61 in report TDR 1092 during a self-a'ssessment
of the DHRS performed in 1992.
The preliminary safety significance
review documented on memorandum C320-92-1287 stated that the observation
was not safety significant because the dose of 0.39 rem "still represents
a negligible portion of the total two-hour thyroid dose of 189 rem."
No
action was taken to resolve the discrepancy.
The licensee stated the FSAR would be revised as part of licensing action
item (LAI) 95076.
The team noted that the LAI wa's initiated May 25,
1996.
The above discrepancies had not been corrected and the FSAR
updated to assure that the information included in the FSAR contained the
latest material as required by 10 CFR 50.71(e).
(Unresolved Item 50-
289/96-201-16)
d.
Desian Basis Document
SDBD-TI-212, " Decay Heat Removal System," Reissue 1, dated November 12, 1996,
Section 4.2.1.3.1 stated that the minimum NPSH available for the DHRS pumps
when drawing from the BWST varied from 115 to 60 feet and that the required
NPSH at 4500 gpm was estimated from test data to be 60 feet.
The 60 feet
value was extrapolated from the manufacturer's pump performance curve that was
included in calculation C-1101-212-5360-026.
The pump test data ranges from 0
to 4121 gpm.
Calculation C-1101-212-5360-026, " Acceptance Criteria for Testing of the
DH-V-14A/B Valves," Revision 0, addressed the maximum pressure drop in the LPI
pump suction line from the BWST during valve testing assuming a minimum level
of 6.3 feet in the BWST. ' The calculation determined the available NPSH for
train A and train B DHRS pumps at a flow of 4500 gpm was 63.75 feet and 61.70
feet respectively, and 51.81 feet for both pumps at the specified acceptance
criteria for the pressure drop across valves DH-V-14A/B during the test.
The
calculation used a value of 10 feet for NPSH required at 3000 gpm flow for the
DHRS pumps and concluded that the available NPSH was acceptable.
The team
questioned how the required estimated NPSH of 60 feet would be assured for the
DHRS pumps at a flow of 4500 gpm.
In response, the licensee stated that
18
.
_ _
l
l
.
adequate NPSH was available to the DHRS pumps under runout conditions until
the,BWST level drops to about 20 feet, and the operating procedures require
the operator to limit the LPI flow to 3300 gpm to protect the pumps.
The
licensee initiated ECD No. C208801 to revise SDBD-T1-212, Section 4.2.1.3.1.
e.
Testina of Check Valves
,
' FSAR Section 6.1.4 stated: "The check valves of the low pressure
a
injection system wi.11 be leak tested e.ach time that the valves are
'
moved from their closed position and before the reactor is returned to
power at their maximum operating pressures per Technical Specification 4.5.2."
The correct TS reference is TS Section 4.2.7.
The only LPI
system check valves in TS Table 3.1.6.1, which lists the valves covered
by TS Section 4.2.7, are DH-V-22A&B. Also, the testing frequency
stated in the TS was different from that in the FSAR.
The testing
frequencies implemented in the TS and surveillance procedure 1300-3T
were less restrictive than that specified in the FSAR.
The licensee
issued PFU 98-T1-111 to revise the FSAR.
The licensee's memorandum 5310-94-024 was the basis for not including
leak testing of check valves DH-V-14A&B in the IST program.
These
valves are in each suction line from the BWST to the DH pumps.
Leakage
i
'
of post-accident reactor building sump water through these valves in
conjunction with leakage through the isolation, valve or the failure of
'
the isolation valve to close could result in increased offsite dose.
The referenced memo contained an analysis that concluded that the post-
accident differential pressure across the check valves would not be in
the direction that could cause leakage into the BWST and thus these
valves do not perform a safety function in the closed position. The
team questioned whether the analysis used the appropriate RB pressure
that could occur during sump recirculation.
The pressure used in the
analysis (6.92 psig) 'was from another calculation for determinirig
parameters for environmental qualification (EQ). This pressure was not
the most currently calculated value. Additionally, the team questioned
.
i
if the EQ calculation represented the most conservative condition for
the maximum reactor building pressure during recirculation.
A new
calculation (C1101-212-E610-053, Revision 0) completed during the
inspection determined a maximum reactor building pressure of 10.53
,
psig during recirculation. At this pressure, valves DH-V14A&B are
exposed to a differential pressure that challenges the reverse flow
prevention capability of these valves. The licensee's plant review
group (PRG) met to consider this matter and established interim
operating restrictions on BWST temperature, RB operating pressure, and
RB fan cooler availability to preclude occurrence of conditions after
an accident that would challenge the closed position of the check
valves.
The licensee issued TFAAI BT6549 to determine the final
disposition regarding testing these valv'es. The design basis for the
valves was apparently not correctly translated into procedures and
instructions as required by 10 CFR 50, Appendix B, Criterion III,
" Design Control," and GPU Nuclear Operational QA Plan, Section 4.0,
" Design Control ."
(Unresolved Item 50-289/96-201-17A)
19
.
4
.
- The DHRS pump discharge cross-connect valves (DH-V38A&B) are normally
closed.
Procedure ATP 1210-6&7 required the discharge cross-connect'
valves to be opened to provide flow through both' injection lines if a
,
DHRS pump failed.
The team noted that if the DH pump discharge check
valve for the non-operating pump (DH-V16A or B) leaked when the cross-
connect valves were open, backflow through the idle pump would reduce
injection flow. The team questioned why these check valves were not in
the IST program for leakage testing in the closed position since they
perform a safety function in the closed position.
The licensee agreed
that these valves perform a safety function in the closed position, and
established the current operability of the valves in the closed
position by performing a special test during the inspection. The
licensee also issued EWR 78254 to add testing of the valves in the
closed position in the IST program. This is the second example of
design basis for the valves not being correctly translated into
procedures and instructions.
(Unresolved Item 50-289/96-201-17B)
- There are two floor drains in each pump vault, one has a swing check
valve and the other a ball check valve; both' valves prevent backflow.
The lice'nsee stated that preventive maintenance procedure U-17
periodically inspected the check valves.
However, the procedure
covered only the swing check valves.
,
In the documents dispositionihg IE Circular 78-06 and IE Information Notice 83-44 that discussed potential damage to safety equipment due to
common mode flooding and backflow through floor drains, the licensee
stated.that procedure U-17 was used to inspect the check valves to
assure operability.
l
The ball check valves were not included in procedure U-17 and the
i
licensee stated that these valves had not been inspected.
The licensee
i
inspected the ball check valves during this team inspection and found
them operable.
The licensee issued PCR l-MT-96-4008 to revise
procedure U-17.
This is the third example of design basis for check
valves not being correctly translated into procedures and instructions.
(Unresolved Item 50-289/96-201-17C)
i
'
f.
Timeliness of Action on Self-Assessment Open Items
The licensee performed a safety system functional inspection (SSFI) of the
DHRS in 1992.
The licensee's report TDR No. 1092, " Low Pressure Injection
System . Safety System Functional Inspection," was approved on January 12,
1993.
The team selected about 30 open items to verify the status of
corrective actions.
In the following instances, the licensee delayed actions
on the open items that had been pending for up to four years:
- SSFI Observations 212-1 through 11 concern design basis piping analysis
and related matters. A safety significance evaluation was performed by
the licensee and documented in Memorandum 5320-92-151, Revision 1,
dated December 9, 1992. This evaluation concluded that there was no
safety concern with these SSFI observations because compared with the
current designs the original calculations used conservatively high
20 '
'
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_ _ _
.
-
.
_ . _ __-
-
_ _ _ _ _
-
_ _ _ _ .
_ . _ . _ _
'
- .
,
.
!
value for seismic response input and low allowable piping stress
v
values. The team noted that the design basis calculations and related
<
'
matters addressed by the SSFI observations still require resolution.
In response to the team's question, the licensee issued closure memos
for 5 of the 11 items in November and December 1996.
The remaining
observations were not closed.
The team reviewed the TFAAI forms for
those items not yet closed and observed that the " Concurred Completion
Date" had been changed numerous times.
SSFI observations 212-12 and 212-13 relate to potentially non-
conservative RB pressure and sump liquid temperatures used in the LPI
pump NPSH calculation. A preliminary safety significance review
documented on Memorandum 5150-92-0069, dated October 27, 1992, stated
that the containment analysis calculations to determine the worst
>
combination of low RB pressure and h gh sump liquid temperature would
be redone once observation 212-22 (SDBD description of BWST volume) was
resolved. Observation 212-22 was resolved by Memorandum 5450-93-0059,
dated July 28, 1993. Calculation C-1101-212-E610, Revision 0, "THI-1
RB Conditions for NPSH Calculations Using GOTHIC" was completed
December 5,1996 and results of a draft LPI NPSH calculation were
reported in Memorandum E610-96-0028, dated December 5, 1996.
This
memorandum stated th'at the available LPI pump NPSH was adequate using
the RB pressure and sump liquid temperature from . calculation C-1101-
212-E610.
'
-
'
- SSFI Observation 212-42 questioned why the DHRS pump vent valves, DH-V-
75A&B and DH-V-76A&B, were not in the EQ program.
These valves have
methyl-vinyl silicone rubber seats that are subject to aging
degradation. The licensee determined that these valves did have a
safety function and' issued memorandum 5450-92-0065 on October 20, 1992,
,
to add these valves to the EQ program.
The team determined that these
valves were not yet included in the EQ program. The licensee performed
w
an evaluation which determined that the valve seats were within their
service life, issued ECD C204175 to add these valves to the EQ program
i
and issued EWR 786478 to ensure the existing parts for, the valves that
_
are in the warehouse meet all appropriate EQ requirements.
'
- SSFI Observation 212-61 regarding the discrepancies between FSAR and TS 4.5.4 is discussed in this report and identified as unresolved-item 50-
289/96-201-16.
The above items are examples of inadequate measures to assure that conditions
adverse to quality are promptly corrected as required in 10 CFR 50, Appendix
B, Criterion XVI, " Corrective Action."
(thresolved Item 50-289/96-201-18).
g.
Reactor BuildinaLSumo
The licensee responded to NRC Bulletin 93-02, " Debris Plugging of Emergency
' Core Cooling Suction Strainers," with letter C311-93-2086 dated June 10, 1993,
'
which stated that procedural controls ensure that any temporary sources of
fibrous material brought into containment will not jeopardize proper Emergency
Core Cooling System suction from the RB sump during a LOCA.
The licensee
'
21
,
.
a
,q
.
stated that the procedural controls referred to in the letter were the
folJowing: Operating Procedure 1102-1, " Plant Heatup to 525'F," Step 55 of
Enclosure 1; Procedure AP 1008, " Good Housekeeping"; Surveillance Procedure
1303-11.16, "DH Leak Test," Steps 8.2.38 and 8.2.39; and General Maintenance
Procedure 1401-18, " Equipment-Storage Inside Class 1 Buildings."
However, none of these procedures specifically control fibrous material
brought into the RB.
The licensee stated that procedure OPS-S98, " Reactor
Building Entry Data Requirements, Surveillances and Inspections," would be
revised to require explicit inspection for fibrous material.
h.
Reactor Buildino Sumo Screens
The team questioned whether the RB sump screens had been analyzed to verify
that they were designed to withstand pressures due to reverse flow when the
RCS drop line is opened during p%st-accident conditions to control boron
precipitation in the reactor core.
The licensee stated that an analysis of
the acceptable RCS pressures below which the RCS drop line could be opened to
the RB sump without damage to the RB sump screens had not been made and issued
EER No. 785982 to initiate such an analysis.
(Inspection Follow-up Item 50-
289/96-201-19).
E1.3.2.3 Conclusions
The SDBD for the system and the system modification documents were detailed
and well written.
The team concluded that the mechanical modifications that
were reviewed were appropriate for resolving the identified problems and the
modifications did not change the design bases of the system.
The team identified a significant weakness in design control with regard to
the improper selection of BWST level for initiating, under post-accident
. conditions, the manual switchover of suction for the DHRS pumps from the BWST
to the reactor building sump on the basis of c calculation that included many
nonconservative inputs. As a result, the licensee concluded that the ECCS
could not be shown to be operable and issued temporary changes to plant
procedures to restore the system to an operable condition.
,
The licensee implemented temporary procedures to preclude conditions after an
.
accident that could challenge the back flow prevention capability of check
valves DH-V-14A&B in the suction line from the BWST.
The teem also identified
other check valves that were not being tested in the closed poaition.
A safety evaluation for a proposed revision to the FSAR did not identify that
an unreviewed safety question was apparently involved when calculations showed
that the available NPSH for the DHRS pumps were less than the required value
considering the actual flows used in the accident analysis and instrument
errors.
The licensee's corrective actions for the open items identif ed by the
licensee conducted SSFI of the DHRS had not always been timcty.
l
22
'
The licensee had taken immediate actions where needed in response to some of
the, team's findings and had initiated corrective actions to resolve the team's
other findings.
E1.3.3 Electrical Design Review
E1.3.3.1 Scope of Review
,
The team reviewed the electrical design for normal and emergency operation of
the DH/LPI pump motors, selected motor operated valves (MOVs), circuit
breakers, fuses and interlocks.
The team also compared the FSAR system
descriptions with the SDBD, drawings, Technical Specifications, operating
procedures, test procedures in order to verify consistency in the documents.
In addition, the team reviewed the calculations for voltage drop, electrical
i
loading, and coordination for selected DH/LPI system components and associated
electrical distribution system components to determine the adequacy of the
system voltage, equipment loading, protective system coordination, and
electrical isolation and independence.
The team reviewed a total of 10
electrical calculations and 2 modifications.
E1.3.3.2 Findings
The team verified that each train of the DHRS was powered from a separate
emergency power bus and that the electrical loading of the individual
'
components, such as pumps and valves had been considered in the EDG capacity
determination.
The sequence and timing of loading of LPI pumps and valves
.
onto the EDG was consistent with the FSAR.
1
The system design documents reviewed by the team adequately supported the
design, except for the open items discussed in the following paragraphs.
a.
Periodic Testino of Molded Case Circuit Breakers
FSAR Section 8.1, " Design Basis" stated that the electrical system design
-satisfied IEEE report No. NSG/TSC/SC4-1, " Proposed IEEE Criteria for Class IE
Electrical Systems for Nuclear Power Generating Stations," June 1969.
Section
6.3 of this IEEE document requires that electrical system components that are
not exercised during normal operation be demonstrated to be operable, and
specific tests and frequency be included.in the maintenance program.
General
Maintenance System, Revision 2, database (GMS-2) specifies the testing
frequency for molded case circuit breakers at TMI-1. The team identified that
molded case circuit breakers in safety-related motor control centers (MCCs) 1A
ES ESF VENT and IB ES ESF VENT had not been tested since their installation in
1986.
In 1993, the licensee added these breakers to GMS-2 and scheduled the
tests for 1997 and 1998. Also, in 1993, the feeder circuit breaker located in
IB ES VALVES Unit 7A for MCC IB ES ESF VENT was deleted from GMS-2. As of
this inspection the circuit breaker in IB ES VALVES Unit 7A was still within
its four year maintenance cycle.
The team identified that the requirements of
10 CFR 50, Appendix B, Criterion V, " Instructions, Procedures, and Drawings,"
regarding inclusion of appropriate quantitative acceptance criteria in
23
.
procedures had not been followed for testing frequency of the breakers in MCCs
lA ES ESF VENT & IB ES ESF VENT resulting in the above breakers not being
tested.
(Unresolved Item 50-289/96-201-20).
b.
Emeraency Feedwater Pumo Electrical Load
In calculation C-1101-424-5310-054, "EFW Pump Brake Horsepower During a
'
LBLOCA/ LOOP Event," Revision 1, it was assumed that the operator would limit
emergenc/ feedwater (EFW) pump flow to 600 gpm, and on that basis the maximum
brake horsepower of the pump was calculated for use in estimating the
emergency diesel generator capacity. Operating Procedure 1106-6, " Emergency
Feedwater," Revision 73, cautions operators not to exceed 600 gpm/ pump. The
available instrumentation in the control room indicate the EFW flow to each
This indicated flow does not include the recirculation and
,
. bearing cooling flows provided by the EFW pump.
The team was concerned that
if the operator limited the EFW flow to 600 gpm as indicated in the control
room, the actual flow through the pump would be higher resulting in a higher
electrical load on the EDG power demand.
The licensee issued WR 786484 to
'
revise the procedure to provide additional operating instructions.
El.3.3.3 Conclusions
4
The electrical design of components that perform engineered safeguards
function.of DHRS was adequate.
The team concluded that the electrical loading
of the DHRS components on the EDG had been appropriately considered. The team
identified a concern with control of GMS-2 database regarding requirements for
periodic testing of a few molded case circuit breakers.
The operator
instructions regarding EFW flow control needed additional clarifications so
that the EFW electrical load on the EDG could be maintained within limit.
El.3 4
Instrumentation ~and Control Design Review
- El.3.4.1 Scope of Review
The instrumentation and control design review consisted of an assessment of
applicable sections of the FSAR, SDBD, and TS for the decay h. eat removal
system. These documents were reviewed for consistency with emphasis on design
bases for instrument loops and associated control logic.
Applicable component and loop analysis, calibration methodology, setpoint
calculations, compliance with single failure criterion, and performance test
analyses;were part of the scope of review.
The team reviewed a total of 4
calculations and 3 modifications related to instrumentation for the DHRS.
E1.3.4.2 Findings
The instrumentation and control requirements for the LPI functions of the
system, such as measurement and indication of LPI flow and pressure and BWST
. level were adequate.
The display of measured parameters and provision of
controls for pumps and valves in the control room were appropriate for the
system function. The instrument setpoints for LPI flow and BWST level were
acceptable, however, the use of BWST low-low level alarm for initiating DHRS
.
24
pump switchover operations potentially could cause air entrainment in the DHRS
pumps as discussed in Section El.3.2.2.a. of thi; report
The system design documents reviewed by the team adequately sucported the
design, except for the discrepancies and open items discussed in the following
paragraphs.
a.
BWST Level Instrumentation
The team questioned the bases for the static head corrections for differences
in elevation between the bottom of the BWST and the level transmitters in data
sheets El-1 through 4 for level transmitters DH-LT-808 and DH-LT-809 in
surveillance procedure 1302-5.19, " Borated Water Storage Tank Level
Indicator," Revision 18.
The licensee could not retrieve documentation in
support of the static head correction for the level transmitters.
However,
the licensee provided engineering evaluation request (EER) 88-070-E which
contained two field sketches showing the distances from the bottom of the BWST
to the center of the level transmitters.
The licensee stated that the
dimensions in these sketches were used for static head corrections in the
level transmitter data sheets.
The sketches had no drawing numbers assigned,
did not show plant elevations or survey marks, and there were no review and
approval signatures.
Since one inch of tank level corresponds to
approximately 530 gallons, inaccuracy in the static head correction would
result.in errors in the assumed quantity of inventory'in the BWST and
exacerbate the problems of DHRS pump switchover discussed in Section
El.3.2.2.a. of this report.
In addition, the team noted that the EER dealt
with static head correction for level switch DH-DPS-914 (low level alarm)
only. The licensee issued TFAAI.BT6541 to update the applicable documentation
for the level transmitters.
The team noted that the requirements in 10 CFR 50, Appendix B, Criterion III, " Design Control," re'garding design control
measures for verifying or checking the adequacy of design was not followed for
the static head correction for BWST level instrumentation.
(Unresolved Item
50-289/96-201-21)
b.
BWST Level Instrument Drift
After the review of several instrument loop accuracy calculations, the team
noted inconsistent treatment of drift errors in these calculations.
For
example: calculation C-1101-662-5350-059, Revision 0, which established the
loop accuracy of BWST level instrumentation stated an arbitrary value'for
drift as one-half of the accuracy.of the loop components; calculation C-Il01-
.212-5350-051, Revision 0, "BWST Level Instrument Drift," empirically
determined drift; GPUN Engineering Standard ES-002, Revision 4, " Instrument
Error Calculation and Setpoint Determination," addressed drift as a variable
'to be considered while performing error analysis, but it did not offer any
guidance on how to calculate drift; and calculation C8706-021 "R.G.1.97'RMT
Transmitter Loop Accuracy" did not address drift error in the loop analysis,
but formed the initial basis for the loop accuracy of the BWST level
instruments.
Licensee memo 5450-88-0023 in calculation C8706-021, stated that
.
the BWST level instrument accuracy requirements were on hold pending
25
completion of two other evaluations. The memo also stated that the basis for
the loop accuracy requirements was included in TDR-883, Revision 1.
The team
revjewed this TDR and did not find the loop accuracy bases.
The team noted instrument loop error concerns similar to the ones discussed in
Section El.2.4.2 of this report.
Calculation C-Il01-662-5350-59 stated that
the existing surveillance tolerance for LT-808 and LT-809 was i 2%. A memo
dated August 8,1988, attached to the calculation stated that to be consistent
with the usual approach which provided margin and allowed for additional
drift, the surveillance accuracy should be lowered to 1.5%.
The memo stated
that supporting calculations were being developed.
The licensee was unable to
provide these calculations. The discussion in the memo was not consistent
with the present surveillance data sheets which specify a 1.0% loop accuracy
which are more restrictive. Additionally, safety evaluation SE-000-214-001,
Revision 2, Sectior, 3.3.2.l(d) stated an allowance of
2% for level error for
the low-low alarm generated by BWST level transmitters DH-LT-808 and DH-LT-
809.
These inconsistencies had not been resolved by the end of the
inspection.
(Inspection Follow-up ..em 50-289/96-201-22)
c. BWST low level Sianal for Oper t?r Action
As stated in Section El.2.2.3 c: this report, the temporary procedure changes
implemented in December,1996 cequired operators to open valves DH-V-6A/B on a
low BWST loel alarm iro.. k el switch DH-DPS-914.
The team noted that GMS-2
.
database iisted the level suitch as a Category 1 device, but it interfaced
with the non safety-related main control room annunciator panel.
The panel is
powered from a vital power source. To provide redundancy for the DH-DPS-914
alarm, the licensee proposed the use of pir.nt computer alarm A0486 which has
an input from DH-LT-0809 instrument loop.
This loop is safety-related; the
plant computer is not.
In response to the team's concern that critical
operations were initiated utilizing non safety-related alarms the licensee
issued TFAAI BT6558 to investigate this concern.
(Unresolved Item 50-289/96-
201-23).
E1.3.4.3 Conclusions
.The instrumentation and control design for the engineered safeguards function
of the DHRS was adequate. The team was concerned with the as-installed
condition of a BWST level transmitter that was not in accordance with
procedures (see Section 1.3.6.2 of this report).
Unverified static head
correction documentation for the BWST level instruments as well as
.
inconsistencies in the loop error calculations were weaknesses in design
I
control.
Reliance on non safety-related instrument components for initiating
critical operations required further review.
E1.3.5 System Interfaces
The team selected the following systems that interface with the DHRS and
verified that the design information for supporting DHRS for each of these
systems was appropriately considered in the design: Decay Heat Closed Cycle
Cooling Water System which provides cooling water to the decay heat removal
coolers, as well as the motor and bearing coolers for the DHRS pumps; Reactor
26
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.
.
_
_ _
_
.
.
._
_
.
_ _ _ _ _ _ .
!
!
Coolant System; reactor building sump from which the DHRS pumps takes suction
under post-accident conditions;
Building Spray system which shares common
suc' tion lines with the DHRS; and MU&P system which shares a common BWST
suction line with the DHRS and operates in the " piggyback" mode with the DHRS
providing suction water from the reactor building sump.
!
In addition to reviewing the interface design information for the above
l
systems, the team performed a walkdown of the installation at the interfaces.
The team did not identify any concerns and the design of the interfaces with
the DHRS were satisfactory.
l
El.3.6 System Walkdown
El.3.6.1 Scope of Review
i
The system walkdown included examinations of the DHRS piping and mechanical
components within the auxiliary building, portions of the interface system
'
piping, installation of instrumentation and electrical components, and
verification of consistency of selected portions of the system with plant
drawings.
The walkdown also included interviews with plant operators in the
control room.
'
l
El.3.6.2 Findings
'
'
a.
Discrepancies in DHRS Flow Diaaram
The team identified the following discrepancies 'between the DHRS flow diagram
302-640, Revision 70, and the as-built system:
The minimum flow orifice (FE 231B) in the recirculation line shown in
the drawing had been removed in 1980 but was still shown on the
drawing.
Also, a local differential pressure indicator (with
~
associated valves) was installed between pressure connections PX 5938
.
and 5948.
The differential pressure indicator was not shown on the
flow diagram.
.
._
The 1-1/4 inch minimum flow recirculation line was shown in the drawing
connected to the DHRS cooler outlet upstream of thermowell TX 540.
However, the line was connected immediately upstream of valve DH-Vl98.
Temperature element DH2-TE2 was shown in the drawing located
immediately downstream of the cooler outlet flange.
However, the
temperature element was installed immediately upstream of valve
DH-Vl9B.
I
The drawing showed an erroneous notation for the air supply to valve
DH-V-61A/B.
Also, the team noted that transmitters DH1-DPTl/2 should
i
be deleted from the drawing because they had been removed.
1
!
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.
i
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L
27
.
.
- . -
-,
-
- - -
_ _ -
- - - -
. - - -
'
.
l
.
The licensee added the above items to ECD No. C204670 to revise the drawing.
b.
Undocumented Modification
!
During the walkdown the team observed a Sse and metal tubing attached to the
1
open end of the discharge of relief va we US-V-63B apparently to direct the
,
relief valve discharge to a floor drain. The team questioned if this was an
'
,
acceptable configuration. Any attachments to the relief valve have the
potential to aenrsely affect the design function of the relief valve.
The
licensee stated that this configuration had not been reviewed by engineering
'
and was not acceptable, and had the attachments to the relief valve removed.
,
Installation of the configuration observed by the team was not in accordance
3
with licensee procedure 1000-ADM-7350.05, " Configuration Change," which
required that configuration changes be documented and reviewed by engineering.
The licensee initiated ENMCF 96-357 to develop any required actions to prevent
similar occurrences.
(Inspection Follow-up Item 50-289/96-201-24).
'
c.
BWST Level Transmitter Enclosure and Freeze Protection
The team noted that the cover plate of the enclosure for BWST level instrument
DH-LT-808 open and the fasteners for the cover missing.
No work was in
progress that could explain the reason for the cover being .open. The
enclosure is a metal barrier that provides separation between redundant
transmitters and process tubing.
The enclosure was installed in accordance
with Section 3.3 of SP-9000-44-001, " Instrument and Control Instrument
Insta*,lation," Revision 0, which specifies the use of protective barriers
where separation criteria could not be met.
The licensee reinstalled the
cover plate and initiated PCR 1-MT-97-8501 on December 18, 1996, to revise
procedure 1302'-5.19 to ensure that the enclosure cover was maintained closed.
{
The team observed that the heat tracing for BWST safety-related level
instrument DH-LT-808 was left coiled within the sheet metal enclosure.
The
. heat tracing was not wrapped around the sensing lines or the transmitter.
This configuration was not in accordance with the vender drawing (ET-30250,
Revision 2) and licensee procedure 1420-HT1, " Heat Trace Repair and
Replacement," Revision 11. Although there had been no history of freezing of
the transmitter or its associated tubing, the team was concerned that the
transmitter had the potential to freeze if left in the as-found condition.
The licensee issued work request (WR) 785993 on December 11, 1996, to correct
this installation.
The team noted that the requirements of 10 CFR 50,
Appendix B, Criterion V, " Instructions, Procedures, and Drawings," regarding
accomplishing activities affecting quality in accordance with instructions,
. procedures, or drawings had not been met. (Unresolved Item 50-289/96-201-25)
d.
Color Codino of Electrical Eauipment
The team noted that the non-segregated phase bus duct supplying the IE 4160V
ES BUS from the auxiliary transformer was painted red instead of gray which is
.the proper color for non nuclear-safety related electrical equipment. Also,
the nuclear safety-related motor control center 1B ES ESF VENT was painted
plantation green, instead of the nuclear safety-related green (Keeler and Long
. green # 2338).
This MCC and the redundant IA ES ESF VENT MCC were not
28
i
j
9
.
- - - - .
-
.-,
.
-
-
.-
4
$
included in the list of safety related equipment in FSAR Section 8.2.2.10
"
h.(1) or in SDD 772-A, " System Design Description for THI-l Nuclear Generating
i
Station Electrical Cable and Raceway Routing," Revision 2.
The licensee
-
issued WRs 786327, 786329, and 786337 to apply labels on the MCCs indicating
'
the correct safety classifications.
.
.
'
e.
Water Leak in Auxiliary Buildina
The team observed water dripping from the ceiling about two feet behind
nuclear safety-related motor control center IA ES VALVES.
The licensee
investigated the leak, determined that it was a roof leak, issued WR 786244
to provide engineering direction for repairs, and issued job order (J0) 130289
.
to repair the leak.
'
E1.3.6.3 Conclusion
,
!
The system flow diagram was consistent with the as-built system, except for
'
4
the identified-discrepancies.
.
i
- _
The team noted weaknesses in the control of field activities in the as-built
4
condition of transmitter DH-LT-808 that created a potential for freezing of
the transmitter and connected tubing and was not in compliance with physical
separation requirements. Also, allowing installation field modifications to a
relief valve discharge without engineering evaluation and proper modification
documentation was of concern.
-
E1.3.7 FSAR.and SDBD Review
The team reviewed the appropriate FSAR sections' for the DHRS and for the
!
associated electrical and -instrumentation and control systems.
The team identified the following discrepancies in the FSAR:
FSAR Section 6.1.2.1.b.1 stated that the maximum flow through the DHRS
!
pump bypass line is 125 gpm i 5 gpm when operating at a shutoff head o.f
425 feet. Also, the system design basis document SDBD-T1-212, Revision
1, stated the same recirculation flow rate. The flow as determined in
'
calculation 1101-212-5360-008, Revision 0 was 150-155 gpm,
FSAR Table 9.5-2 contained several entries of two design temperatures
a
-
and/or pressures for DHRS pumps and coolers. The licensee stated that
'
these were incorrect.
- FSAR Section 8.2.2.10.9 stated that nonsegregated, metal enclosed.4160
V bus ducts were used for major circuit runs from the unit auxiliary
-
transformers to 4160 V and 6900 V buses.
The equipment bill of
materials showed a voltage of 7.2 KV for the bus ducts.
tracing load as BS-T-2B; however, thi.s load designation was for the
!
sodium hydroxide tank heat tracing.
.
29
.
._
--
.
-
..
_ - - -
- . _- -
- . - -
_
_-.-
..
- .
.
e
The above discrepancies had not.been corrected and the FSAR updated to assure
4
t'1at the information included in the FSAR contained the latest material as
required by 10 CFR 50.71(e).
(Unresolved Item 50-289/96-201-26)
There were several discrepancies in the table on page 99 of SDBD-T1-642,
"ESAS."
The numbers for the actuation setpoints, setpoint function, and TS
limit numbers for RCS Pressure did not correspond with one another.
The licensee issued PFU 98-TI-126, PFU 98-T1-129 and ECD C204652 to resolve
the above discrepancies.
Section 4.1.3.7, " Single Failure," of SDBD-T1-212 states that the definition
of active component has been interpreted to exclude self-actuating components
for which there is adequate positive force to assure they function (i.e.,
This definition and its application in safety-related systems
had not been resolved by the end of the inspection.
(Inspection Follow-up
Item 50-289/96-201-27)
El.4 Control of Calculations
During the inspection the team reviewed several engineering calculations and
other documents for both the MU&P system and the DHRS as discussed in the
{
previous sections of this report. The team had two design control concerns
regarding these documents.
-
Licensee engineering procedure EP-006, " Calculations"' specifies the
i
requirements for preparation, review, and approval of calculations.
Technical
data reports (TDRs) are prepared in accordance with procedure 5000-ADM-
73416.01, and are not required to be design verified or maintained in
'
currently accurate status.
The team identified the following memoranda, TDRs,
and plant engineerir;g evaluation requests (EERs) that were used to perform
safety-related calculations that did not comply with EP-006:
= Memorandum 5310-92-366, dated December 22, 1992, " Evaluation of TMI-1
HPI SSFI Observation No. 211-10" included a calculation to resolve a
concern from the licensee's SSFI of the MU&P system regarding dead
heading of makeup pumps under various combinations of op.erating pumps.
This calculation was neither formally reviewed nor approved.
. Memorandum 5310-92-024, dated March 1, 1994, "DH-V-14A/B, DH-V-5A/B,
BS-V-52A/B and BS-V-2A/B valves and IST Program." incorportted an
analysis to justify not leak testing check valves DH-V-14A&3 as part of
the inservice testing program as discussed in Section 1.3.2.2.e. of
this report.
This analysis was in error.
In response to the team's
concern the licensee implemented temporary operating restrictions to
,
ensure that after a postulated accident these check valves would not be
i
exposed to a differential pressure that would challenge their leak.
'
tightness.
1
'
30
.
?
.
-~
I
TDRs reviewed by the team that contained safety-related analyses
.
included TDR No. 836, Revision 6, dated January 31, 1995, " Evaluation
,
of Loading for the Emergency Diesel Generator and Engineered Safeguards
(ES) Buses" and TDR No. 995. Revision 3, dated January 18, 1996,
" Voltage Drop Study for Degraded Grid Condition." The licensee had
,
identified this issue and initiated quality deficiency report (QDR)
962012 on July 25, 1996, to document the inappropriate use of TDRs 836
and 995 for design calculations.
EERs 88-060-E,"BWST Level Alarm Setpo' int Change" dated July 29, 1988,
=
and 88-070-E, " Calibration of DH-DPS-914," dated august 10,1988,
contained setpoint correction calculations for specific gravity.
The above examples indicated that the requirements in 10 CFR 50, Appendix B,
i
Criterior. III, " Design Control," regarding verifying or checking the adequacy
of design apparently were not being met.
Also, the requirements for
calculations in licensee procedure EP-006, " Calculations," were not being
followed.
(Unresolved Item 50-289/96-201-28)
The team's second concern with control of calculations was that several
calculations and analyses were not the latest documents because subsequent
calculations performed the same, or very similar, analyses.
The older
'
analyses were not identified as superseded. Additionally, changes to data
from other sources that were used as input to calculations were not
.
'
consistently incorporated into the completed calculations.
Examples observed
by the team included:
TDR 114, Revision 1, " Adequacy of Station Electric Distribution System
Voltages," had been superseded by TDR 995, Revision 0, dated February
21, 1990, " Voltage Drop Study on Degraded Grid Conditions." TDR 114,
Revision I continued to be referenced in SDBD-TI-211, Revision I while
TDR 995 was not referenced in the SDBD.
The current revision of TDR
995, Revision 3, dated January 18, 1996, did not state that it
superseded TDR 114.
'
Calculation C-Il01-212-5360-027, "NPSH Available of LPI and BS Pumps
a
,
Following Large Break LOCA," Revision 0, dated October 2, 1990, stated
that calculation C-1101-214-5300-008, " Reactor Building Pump Flot Based
on NPSH Available," Revision 2, did not consider conditions that have -
changed.
It appeared that calculations C-Il01-212-5360-027, Revision 0
and C-1101-212-5310-050, Revision 0. replaced calculation C-1101-214-
5360-008. The supersedd status of the calculation was not noted in the
.
records.
.
- Calculations C-Il01-211-5320-001 "THI-l HPI Flow Analysis," C-Il01-211-
5320-002, "TMI-1 Makeup System Flow Analysis," and C-Il01-211-5360-004,
"TMI-l HPI Flow Analysis," documented HPI flow analyses during the
months of June and September 1983'.
It appeared that the last
calculation superseded the other two.
.
31
'-
'
.
_
_ . . .
.- . _ _ __ __. . _ _ _ _ . _ . _ . _ . _ _ _ _ _ _ . - _ _ _ . ._ _ _ . _ . . _ . _ _ _ . . . . __
,
.
The licensee issued quality deficiency report (QDR) No. 972001 to correct the
l
above deficiencies in calculation and design control.
,
/
X1
Exit Meeting
After completing the on-site inspection, the team conducted an exit meeting
with the licensee on January 31, 1997, that was open for public observation.
During the exit meeting, the team leader presented the results of the
inspection. A list of persons who attended the exit meeting is contained in
'
Appendix B.
.
,
9
4
e
i
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L
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l
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!
>
.
.
s
..
l
.
, . _ _ _ - .
-
4
4
l
l
!
!
!
'
,
l
32
j
?
2
l
,
I
r
w
m
e.w..
,
-w
,
m
_. _
_
__ _ _ _ - _ _ _
_ . _ _ . - . _
_ _ _
__ ,
)
-
APPENDIX A
'
-
.
Ooen Items
l
4
This report categorizes the inspection findings as unresolved items and
inspection follow-up items in accordance with the NRC Inspection Manual,
Manual Chapter 0610. An unresolved item (URI) is a matter about which more
information is required to determine whether the issue in question is an
acceptable item, a deviation, a nonconformance, or a violation.
The NRC
Region I office will issue any enforcement action resulting frorc the review of
3
the identified unresolved items. An inspection follow-up (IFI) item is a
'
matter that requires further inspection because of a potential problem,
j
because specific licensee or NRC action is pending, or because additional
l
information is needed that was not available at the time of the inspection.
1
.
Item Number
Findina
Title
Type
4
1
50-289/96-201-01
IFI
Letdown line Break in the Auxiliary Building
(Section El.2.2.2.a)
,
50-289/96-201-02
IFI
Evaluation of Simultaneous Start of MU&P Pump
,
MU-P-lC and Suction and Discharge Valves
,
(Section El.2.2.2.b)
50-289/96-201-03
IFI
Evaluation of Gas Accumulation in Suction
]
Piping for MU&P Pump MU-P-lC (Section
- !
El.2.2.2.b)
1
.
!
I50-289/96-201-04
Adequacy of Makeup Tank Pressure / Level Curves
j
l
(Section E1.2.2.2.c)
1
'50-289/96-201-05
Design Basis Valve Stroke Times in Surveillance
- 7
Procedure (Section El.2.2.2.d)
.
50-289/96-201-06
IFI
Consequences of Failure of Auxiliary Steam
.
Piping (Section E1.2.2.e)
l
.
50-289/96-201-07
IFI
Loss of Pressure in MU&P Tank due to Letdown
i
'
Line Break (Section E1.2.2.e)
50-289/96-201-08
IFI
M&UP Pump NPSH When Taking Suction From.BWST or
Makeup Tank (Sect. ion E1.2.2.e)
50-289/96-201-09
IFI
Incomplete DC System Voltage Drop Calculations
-
(Section E1.2.3.2.a)
i
50-289/96-201-10
Alignment of MU-V-18 DC Power Supply (Section
.
'
El.2.3.2.b)
'
A-1
i
_
-
-
. .--
-
--
.
--- . , . - . - - -
. . - _ . .
. - . . - - -
l
50-289/96-201-11
Makeup Tank Level Instrument Loop Tolerances
(Section El.2.4.2.a)
50-289/96-201-12
FSAR Discrepancies (Section El.2.7)
i
50-289/96-201-13
Adequacy of BWST Setpoint for DHRS Pump
Switchover to RB Sump (Section El.3.2.2.a)
50-298/96-201-14
Adequacy of Safety Evaluation of an FSAR Change
(Section E1.3.2.2.b)
.50-289/96-201-15
IFI
Technical Specification Discrepancy (Section
El.3.2.2.c)
50-289/96-201-16
Discrepancy Between FSAR and Technical
Specifications Regarding DHRS Leakage (Section
l
El.3.2.2.d)
'
50-289/96-201-17A
Leak Testing'of DHRS' Pump Suction Check Valves
(Section E1.3.2.2.e)
50-289/96-201-178
Leak Testing of DHRS Pump Discharge Check
Valves (Section E1.3.2.2.e)
50-289/96-201-17C
-Inspection of DHRS Pump Vault Floor Drain Check
Valves (Section E1.3.2.2.e)
1
-50-289/96-201-18
Timeliness of Action on SSFI Open Items
(Section E1.3.2.2.f)
50-289/96-201-19
IFI
Evaluation of Reactor Building Sump Screen for
Reverse Flow (Section E1.3.2.2.h)
50-289/96-201-20
Testing o' Molded Case Circuit Breakers
~
'
(Section U.3.3.2.a)
'
50-289/96-201-21
Static Head Correction.for BWST Level
Transmitter (Section El.3.4.2.a)
50-289/96-201-22
IFI
BWST Level Instrument Drift (Section
E1.3.4.2.b)
50-289/96-201-23
Selection of BWST Low Level Alarm for Operator
Action (Section El.3.4.2.c)
.
50-289/96-201-24
IFI
Undocumented Modification (Section El.3.6.2.b)
50-289/96-201-25
BWST Level Transmitter Enclosure and Heat
l
Tracing (Section El.3.6.2.c)
,
L
50-289/96-201-26
FSAR Discrepancies (Section El.3.7)
l
'
l
A-2
.
t
.
, ~ , . *
.
..
50-289/96-201-27
IFI
Definition of Single Active Failure (Section
El.3.7)
,
50-289/96-201-28
Control of Calculations (Section El.4)
t
e
,
t
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.
A-3
.
.
cm-
l
l
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- _ _= -
-
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.
. . . - . _ .
- . _ - . . - - . - - _ . -
- .
.
_-
h
i
APPENDIX B
1
EXIT MEETING ATTENDEES
NAMI
ORGANIZATION
D.J. Distel
GPUN, Licensing
'
P.D. Milano
NRC, Acting PD1-3, NRR/DRPE
S.K. Malur
NRC, Team Leader, NRR/PSIB
,
D.P. Norkin
NRC, Section Chief, NRR/PSIB
i
J.T. Wiggins
NRC, Director, DRS, Region 1
E.M. Kelly
NRC, Branch Chief,DRS/SEB, Region 1
i
B.C. Buckley
NRC, Sr. Project Manager, NRR/DRPE
S.L. Hansell
NRC, Resident Inspector,.TMI-1-
!
T.J. Kenny
NRC, DRS, Region 1
L.A. Karinch
GPUN, TMI Communications
'
P. Walsh
GPUN, Director, Equipment Reliability
J. Wetmore
J. Kneubel -
GPUN, Manager, Regulatory Affairs
GPUN, Site Director
G.R. Skillman
GPUN, Director, Configuration Control
A.T. Asarpota
GPUN, Manager, Modification's
E.P. O'Donnell
GPUN, Director Engineering Support
S. Maingi
Bureau of Radiation Protection, PA
.
H. Wilson
GPUN, Supervisor, Maint. Assessment
.
~ T.M. Dempsey
GPUN, Engineering Support
i
L.R. Freeland
Duquesne Light Co.
- W. McSorley
GPUN, System Engineer
j
H. Heilineier
'
M. Wells
GPUN, Communications
'
M.G. Kapil
GPUN, Manager, Elec. Power & Inst.
D. Hull
GPUN, EP&I Engineer
T. Noble
GPUN, Engineering
i
R. Hernan
'
NRC, Project Manager, NRR/PD2-3
M.'J. Ross.
GPUN, Director, 0&M
N.A. Sheehan
NRC, Public Affairs
-
!
E. Hammond
GPUN, NS&C Staff
l
P. Campbell
Winston & Strawn
J.R. Pearce
GPUN, Engineering
,
,
i
8
e
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B-1
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4
e
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.- -
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<
ATTACHMENT C
LIST OF DOCUMENTS REVIEWED
,
-
1.
DRAWINGS
Drawina No.
Title
4
i
'SS-201-001
" Terminal Box # T371," Rev 8
201-251
" Arrangement Diagram 125/250V DC Main Distribution Panel
.
1A", Rev 9
.
201-252
" Arrangement Diagram 125/250V DC Main Distribution Panel
-
IB", Rev 10
201-261
" Arrangement Diagram Engineered Safeguards 125/250V DC Main
Distribution Panel IM",.Rev 12
l
201-039, Sh 1
"480V Control Center General Notes & Index", Rev 12
,
201-039,.Sh 2
"480V Control Center General Notes & Index", Rev 6
201-040
"480V Control Center Location And Orientation", Rev 14
i
201-043, Sh 1
"480V Control Center 1A Engineered Safeguards", Rev 30
201-043, Sh 2
"480V Control Center 1A Engineered Safeguards", Rev 21
-
201-043, Sh 3
"480V Con' trol Center IA Engineered Safeguards", Rev 27
201-044, Sh 1
"480V Control Center 1B Engineered Safeguards", Rev 28
201-044, Sh 2
"480V Control Center IB Engineered Safeguards", Rev 25
201-044, Sh 3
"480V Control Center'1B Engineered Safeguards", Rev 27
201'052, Sh 1
."480V Control Center 1A ES Valves", Rev 40
201-052, Sh 2
"480V Control Center IA ES Valves", Rev 26
201-053, Sh 1
"480V Control Center IB ES Valves", Rev 39
'
'
201-053, Sh 2
"480V Control Center IB ES Valves", Rev 26
- 201-062, Sh 1
"480V Control Center IA ES Screen House", Rev 20
201-062, Sh 2
"480V Control Center 1A ES Screen House", Rev 25
ea E201-063,-Sh 1
"480V Control Center 1B ES Screen House", Rev 24
r
- 201-063,'Sh 2
"480V Control Center IB ES. Screen House", Rev 23
"=
_201-069, Sh 1
"480V Control Center IC ES Valves", Rev 31
-
201-069, Sh -
"480V Control Center 1C ES Valves", Rev 28
-
.201-076
"480V Control Center IA ES ESf Vent Bldg", Rev 2
+ 201-077
"480V Control Center IB ES ESf Vent Bldg", Rev 3
206-011
" Main One Line and Relay Diagram", Rev 35
206-022
"One Line and Relay Diagram 4160V.ES Switchgear", Rev 19
20b 032
"One Line and Relay Diagram ES Screen Hse React Bldg 480V
SWGR", Rev 12
'
206-051
"One Line and Relay Diagram 250/125V DC Sys & 120V AC Vital
_
Instrumentation", Rev 22
208-264
" Electrical Elementary Diagram 480V Switchgear", Rev 1
. 208-288
" Ele ~ctrical Elementary Diagram 480V Switchgear (ES) (IP-
4A)", Rev 2
.
208-292
" Electrical Elementary Diagram 480V Switchgear (ES) (1S-
4A)", Rev 2
208-211
" Electrical Elementary Diagram, 4160V Switchgear(ES)(1D6)
DH-P-1A", Rev.2
'208-212
" Electrical Elementary Diagram, 4160V Switchgear (ES) (1E7)
DH-P- 1B", Rev 4
.C-1
-
.
-
.
_.
.
-
_ _ -
.
. _ ~ -
.
._-
. -
.
__ _
.
208-213
" Electrical Elementary Diagram 4160V Swii.chgear (ES) (107)
'
MU-P-1A", Rev 6
208'-214
" Electrical Elementary Diagram 4160V Switchgear (ES) (lE8)
MU-P-1C", Rev 10
4
208-215
" Electrical Elementary Diagram 4160V Switchgear (ES) (108)
MU-P- 1B", Rev 8
'
208-216
" Electrical Elementary Diagram 4160V Switchgear (ES) (1E9)
MU-P-1B", Rev 9
208-431
" Electrical Elementary Diagram, 480V Control Center 1 A-ESV,
,
3
'
Unit 3C (DH-V-7A) & 1B-ESV, Unit 3C (DH-V-78)", Rev 3
208-432
" Electrical Elementary Diagram, 480V Control Center 1 A-ESV,
Unit 3A (DH-V-5A) & 1B-ESV, Unit 3A (DH-V-5B)", Rev 3
>
208-433, Sh 1
" Electrical Elementary Diagram, 480V Control Center 1 A-ESV,
j
Unit 1C (DH-V-4A)", Rev 11
'208-433, Sh 2
" Electrical Elementary Diagram, 480V Control Center 18-ESV,
4
Unit 1C (DH-V-4B)", Rev 0
.
208-434, Sh 1
" Electrical Elementary Diagram, 480V Control Center 1 A-ESV,
Unit 3B (DH-V-6A)", Rev 4
208-434, Sh 2
" Electrical Elementary Diagram, 480V Control Center 18-ESV,
Unit 3B (DH-V-6B)", Rev 2
'
208-435
" Electrical Elementary Diagram 480V Control Center -
MU-V-1A/1B", Rev 6,
208-437
" Electrical Elementary Diagram 480V Control Center
18-ESV-Unit 4D Letdown Cooler A Outlet Valve MU-V-2A", Rev 7
208-438
" Electrical Elementary Diagram 480V Control Center l A-RWD
-Unit 6B - Valve MU-V-8", Rev 8
'208-439
" Electrical Elementary Diagram 480V Control Center 1B-RWD
-Unit 6C - Valve MU-V-12", Rev 3
208-440, Sh 1
" Electrical Elementary Diagram 480V Control Center 1A-ES
-Unit 7D - BWST Outlet Valve MU-V-14A", Rev 3
208-440, Sh 2
" Electrical Elementary Diagram 480V Control Center 18-ESV
-Unit 4A - BWST Outlet Valve MU-V-14B",-Rev 1
208-441
" Electrical Elementary Diagram 480V Control Center 1 A-ESV
-Unit 4D - MU-V-25" Rev 8
-_208-442, Sh 1
" Electrical Elementary Diagram 480V Control Center 1 A-ESV -
Unit 4B - MU-V-16A" Rev 11
_.208-442, Sh 2
" Electrical Elementary Diagram 480V Control Center 1A-ESV -
Unit 4C - MU-V-16B", Rev 4
~208-442, Sh 3
" Electrical Elementary Diagram 480V Control Center 1B-ESV -
Unit 4B - MU-V-16C" Rev 4
208-442,; Sh 4
" Electrical Elementary Diagram 480V Control Center 1B-ESV -
Unit 4C - MU-V-16D", Rev 0
208-452
" Electrical. Elementary Diagram 480V Control Center 1C-ESV -
Unit 3A - DH-V-1", Rev 7
208-453
" Electrical Elementary Diagram 480V Control Center 1C-ESV. -
Unit 3B - DH-V-2", Rev 8
208-454
" Electrical Elementary Diagram 480V Control Center 1C-ESV -
Unit 4B - DH-V-3", Rev 3
208-512
" Electrical Elementary Diagram 480V Control Center
lA-Radwaste- MU-V-33A-D and MU-V-38", Rev 2
'208-523
" Electrical Elementary Diagram 480V MCC - MU-P-2A/B/C", Rev
6
,
.
C-2
.,
_ _
__.
_
__
._
_ _
. - _ _ _
. _ _ _ . . .
'l
i
208-562, Sh 2
" Electrical Elementary Diagram 480V Control Center 1B-ES
-Unit 2A, MU-P-3C", Rev 2
" Electrical
Elementary Diagram 480V Control Center 1A-ESV -
Unit 108, MU-P-4A", Rev 2
208-648
" Electrical Elementary Diagram 480V Control Centers 1C-ESV -
Unit lE, MU-P-4B", Rev 4
208-649
" Electrical Elementary Diagram 480V Control Centers 18-ESV -
Unit 1D, MU-P-4C", Rev 5
208-690, Sh 1
" Electrical Elementary Diagram 480V Control Center lA-ESV -
Unit 2D, MU-V-36", Rev 5
.
208-690, Sh 2
" Electrical Elementary Diagram 480V Control Center 18-ESV -
Unit 2D, MU-V-37", Rev 3
l
.
l
208-691
" Electrical Elementary Diagram 480V Control Center lA-ESV -
'
Unit SD, MU-V-39" Rev 2
208-721
" Electrical Elementary Diagram 480V Control Center
.MU-V-217",
Rev 4209-020" Electrical Elementary Diagram - DC &
,
l
Miscellaneous
- MU-V-4, MU-V-6A, MU-V-1 18", Rev 5
1
209-021
" Electrical Elementary Diagram - DC & Miscellaneous -
MU-V-10",
Rev 5
209-022, Sh 1
" Electrical Elementary Diagram - DC & Miscellaneous -
MU-V-3", Rev 10
209-025
" Electrical Elementary Diagram DC Control for Valves,
MU-V-26", Rev 5
209-080
" Elementary Diagram - MU-V-27", Rev.6
209-129
" Electrical Elementary Diagram, DC & Miscellaneous,
DH-V-61B", Rev 3
209-145
" Electrical Elementary Diagram DC & Miscellaneous, MU-V-51",
Rev 1
302-001
" Symbols F1ow Diagram", Rev 13
302-003
" Component and System Identification Index", Rev 37
302-051
" Aux. Steam Flow Diagram"
c 31303C
" Nelson Electric Drawing - Panel 2A and 28"
31303C
" Nelson Electric Drawing - Electric Trace Schedule 3003-1"
L32-262
" Westinghouse ratings for circuit breaker trip and close
l
coils"
4692-51-629-0
" Heater Tracing Panel Schedule - Panel 2A"
_4692-51-631-0
" Heater Tracing Panel Schedule - Panel 2B"
SS-209-050
" Electrical Elementary Diagram DC and Miscellaneous", Rev 3
i
SS-209-531
" Electrical Elementary Diagram Engineered Safeguards", Rev
11
SS-209-631
" Electrical Elementary Diagram Engineered Safeguards", Rev
13
'SS-209-129
" Electrical Elementary Diagram DC and Miscellaneous", Rev 3
SS-209-348
" Electrical Elementary Diagram Waste Handling System", Rev 3
209-022, Sh 2
" Electrical Elementary Diagram DC and Miscellaneous Normal
Makeup Valve MU-V-18", Rev 2.
209-887
" Electrical Elementary Diagram Alarms (AF) Vertical Panel
'(PL), Rev 2
GS-213-689
" Bull. 930 Form E Automatic Transfer Switch", Revision C
E-213-021
" Electrical Heat Tracing Misc. Sodium and Borated Water
4
Storage Tank Piping" Revision 7
>
~GS-219-565
"DC Automatic Transfer Switch", Rev A
.
C-3
i
_ _ _ _ _ .
.
.
.
229-001
" Substation One Line Diagram", Rev 26
229-002
" Substation One Line Diagram", Rev 17
302-001
" Flow Diagram Symbols", Rev 1
302e002
" Component & System Index", Rev 37
Panel 2A
" Heat Trace Panel Schedule" Rev 4,
Panel 2B
" Heat Trace Panel Schedule" Rev 4
3/303C
" Electric Heat Trace Schedule" Rev 3
WD/31303C
" Heat Trace Wiring Diagram
-
302-640
" Decay Heat Removal Flow Diagram, Rev 70"
'
302-645
" Decay Heat Closed Cycle Cooling Water Flow Diagram, Rev 32"
302-660
"Make-up & Purification Flow Diagram, Rev 34"
302-661
"Make-up & Purification Flow Diagram, Rev 47"
.
4
1
0
e
C-4
.
.
I
2.
LICENSEE CALCULATIONS
densee Calculation No.
Title
C-1101-211-5300-030, Rev 0
"MUF4 Mod- radiological effect" 3/25/91
C-1101-211-5300-031, Rev 0
"MUF4 Mod- radiological effect-area 15
3/25/91
C-1101-211-5310-047, Rev 0
" Makeup Tank Drawdown During LOCA"
C-1101-211-5320-001, Rev 0
"HPI Flow Analysis" 6/13/83
C-1101-211-5320-002, Rev 0
" Makeup System Flow Analysis" 6/28/83
C-1101-211-5320-004, Rev 0
"HPI Flow Analysis" 9/6/83
C-1101-211-5320-020, Rev 0
" Effects of loss of cooling in letdown
cooler on downstream piping due to an
Appdx R fire" 12/18/87
C-Il01-211-5320-021
" Effects of loss of cooling in letdown
cooler on downstream piping due to an
Appdx R fire" 1/13/88
C-1101-211-5320-023, Rev 0
" Evaluation of change of MU24FE of Makeup
Flow to RCS"
C-Il01-211-5320-025, Rev 0
" Evaluation of HPI cross connect piping at
RCS temp" 9/8/89.
C-1101-211-5320-051, Rev 0
"Recirc pipe
"HPI thermal (MU-P-1 A)" 9/15/93
C-1101-211-5320-044, Rev 0
upgrade" 1/19/93
C-1101-211-5320-045, Rev 0
"HPI thermal upgrade" 1/19/93
C-1101-211-5320-046, Rev 2
"RCP seal injection leakoff" 1/17/95
C-1101-211-5450-006
" Capacity of makeup flow" 4/25/85
C-1101-211-5450-012, Rev 0
" Makeup & Letdown for DH Removal 0 10%"
12/2/83
.
C-1101-211-5450-018, Rev 0
" Makeup and letdown sys. used to remove DH
0 5%" 12/2/83
-
C-1101-211-5450-029, Rev 0
"THI-1 Reactor Building Sump Temperature
'
Daring a SBLOCA where HPI Piggy Back Mode
is Required"
C-1101-211-5350-054, Rev 0
" Error for Seal Injection and HPI Flow"
C-1101-211-5350-057, Rev 0
" Makeup Tank Level Drift"
C-1101-211-5360-001, Rev 0
" Drain down of MU-T-1 during MU pump
operation from BWST" - (Superceded by calc
5310-047)
C-1101-211-5360-003
"NPSH for MU Pumps"
C-1101'-211-5360-009, Rev 1
"HPI flows for selected RCS pressures"
7/30/84
C-1101-2il-5360-014, Rev 0
"MU pump NPSH-High Suction Temp" 11/12/86
C-1101-211-5360-019, Rev 0
" Makeup flow at low RCS pressure" 12/7/87
C-1101-211-5360-024, Rev 0
"HPI piping pressurization due to valve
leakage" 8/11/89
C-1101-211-5360-026, Rev 0
"HPI Pump NPSH in Piggy-back Mode"
C-1101-212-5300-043, Rev 1
" Load Of The.ECCS. Pumps During LBLOCA"
C-1101-212-5310-044
" Vortex Limits for LPI and BS Pumps in the
RB"
4
C-1101-212-5310-050, Rev 0
"TMI-1 BWST Vortex Determination"
C-1101-212-5350-051, Rev 0
"BWST Level Instrument Drift"
.
C-5
'
.
_
l
C-Il01-212-5360-001
"DH System Performance Under Degraded
Conditions",
C-l'101-212-5360-002
" Pressure Requirements for DHR Operation
with both DHR Pumps Running at 3000 GPM"
C-Il01-212-5360-004
"DHR System Drop Line NPSH Concerns"
C-Il01-212-5360-008
" Decay Heat Removal System Resistance"
i
C-Il01-212-5360-020
"BWST Gravity Feed During Loss of DHR"
C-1101-212-5360-026
" Acceptance Criteria for Testing of the
DH-V-14A/B Valves"
C-1101-212-5360-027
"NPSH Available for LPI and BS Pumps
Following large Break LOCA"
C-1101-212-5360-038
" Gravity Flow through the Drop Lines in
the DHR System"
C-1101-212-5450-039
" Excess NPSH Available for LPI and BS
Pumps for LB LOCA EQ Analysis"
C-1101-624-5350-002, Rev 0
"MU Tank Level Error for Accident
'
Conditions (MU-14-LT Loop"
C-1101-662-5350-015
"TMI-1 Decay Heat Removal Flow Evaluation
Instrumentation Loop Accuracy"
C-1101-662-5350-049, Rev 0
"TMI-l Make Up Tank Level Error for
Accident Conditions (LT.-778 Loop)".
-
C-1101-662-5350-059, Rev 0
"RG 1.97 BWST Level Loop Tolerance"
C-Il01-730-5350-001, Rev 1
"GL.89-10 Mov's Degraded Grid Voltage Drop
Calculation",
.
C-1101-733-5350-003, Rev 2
"TMI-l Class IE 480v Unit substation
Settings For Conversion To Solid State
Trip Units"
C-1101-734-5350-003, Rcv 1
"TMl-1 Station Battery A Capacity
Calculation",
C-1101-734-5350-004, Rev 1
"THI-1 Dc System Calculation"
C-1101-735-5350-002, Rev 1
"TMI-l 120v Vital Ac Distribution System
i
Coordination"
C-1101-735-5350-003, Rev 0
" Vital Ac Panel VBA Voltage Drops"
C-1101-741-5351-003, Rev 0
" Relay Settings For Diesel Generator Up To
Voltage And Thermal Overload Relays"
C-1101-823-5450-001, Rev 2
"TMI-1 LBLOCA EQ Temperature Profile using
the Gothic Computer Code"
C-1101-862-5360-002, Rev 0
"TMI-1 EDG Fuel Requirements"
C-3340-95-001, Rev 0
"HPI Flow Test-Estimate of Random Errors"
C-3340-96-001, Rev 0
"MU-T-1 Pressure vs_ Level"
C-8706-021, Rev 1
"THI-1 New RG 1.97 RMT Transmitter Loop
Accuracy's"
TI-5360-212-006
"DHR Capability", dated 5/25/82.
C3064A-322-052
" Unit 1 Restart Modification Sizing of
External Oil Reservoirs for DHR Pumps"
DC-218-002, Rev 1
" Miscellaneous Voltage Drop / Ampacity
Calculations For Cycle 6 Cable Routing"
DC-412384-5, Rev l'
" Screen House Voltage Drops Using
Rockbestos Cable"4192-077, Rev 0
Gilbert Associates Calculation 6/27/73
C-6
.
..
--
.- --
-.
...
-
-
.
..
..
-
.
3.
TECHNICAL DATA REPORTS
l
'
TDR 836, Rev 6
" Evaluation Of Loading For The Emergency Diesel Generator
,
And Engineered Safeguards (ES) Buses",
TDR 995, Rev 3
"TMI-1 Voltage Drop Study On Degraded Grid Condition",
TDR 1055, Rev 1
" Direct Current Power System Underrated Fuses",
-TDR 883, Rev 1
"TMI-l Equipment Qualification-Performance Evaluation of
Instrument Loops"
i
TDR 1034, Rev 1
"High Pressure Injection System - Safety System functional
l
Inspection"
(Inspection
Observations
211-
01,03,08,09,10,22,25,26,27,29,30,35,36,37,and 38)
TDR 1092, Rev 0
" Low Pressure Injection System - Safety System Functional
Inspection" (Inspection Observations
21201.02,03,04,05,06,07,08,09.10.11,14,15,16,17,18,19,20,21,
4
22,23,24,27,34,36,39,42,44,47,50, and 61)
!
4.
OPERATING PROCEDURES
1101-1, Rev 58
" Plant Limits and Precautions"
1102-12, Rev 20
" Hydrogen Addition and Degassification"
3
1104-2, Rev 104
"Make0p & Purification System'
1104-2, Rev 105
" Makeup and Purification System" Secondary - Auxiliary
Operator Log, Rev 56
1107-4, Rev 153
" Electrical Distribution Panel Listing"
'
1302-5.17, Rev 17
"Make-Up Tank Level Instrumentation"
1302-5.18, Rev 26
"HPI/LPI Flow Channel Calibration"
.
1302-5.19, Rev 18
" Borated Water Storage Tank Level Indicator"
2
5.
EMERGENCY AND ABNORMAL OPERATING PROCEDURES-
2
.1202-2, Rev 41
" Loss of Station Power"
1202-2, Rev 42
" Loss of Station Power"
1202-12, Rev 41
" Excessive Radiation Levels"
.
1202-29, Rev 51
" Pressurizer System Failure"
4
'
-.1202-35, Rev 30
" Loss of Decay Heat Removal System"
1202-37, Rev 48
"Cooldown from Outside the Control Room"
1203-15, Rev 20
" Loss of R.C. Makeup / Seal Injection"
1203-16, Rev 38
" Reactor Coolant Pump and . Motor Malfunction"
1210-6, Rev 22
"Small Break LOCA Cooldown"
1210,-7, Rev 22
"Large Break LOCA Cooldown".
1210-1031, Rev
" Abnormal Transients Rules, Guides, and Graphs"
6.
ALARM RESPONSE PROCEDURES
.
THI-1 Alarm Response Procedure, Map B Rev 9," Main Annunciator Panel B"
.
C-7
.
_ _ . _ _ , _ _ _ _
_..___ _ _ _
_ _ _ _ _ _ _ _ _ _._ . _ _ _ _
,
l
.
t-
.
7.
' SURVEILLANCE PROCEDURES
i
d'
1302-5.31A, Rev 15
"4160V D and E Bus Degraded Grid Undervoltage Relay
'
System Calibration"
.
'
1302-5.318, Rev 13
"4160V D and E Bus Loss of Voltage Relay system
Calibration"
.
!
I
1302-5.31C, Rev 9
"4160V ID Bus Loss of Voltage / Degraded Grid Timing
"
Relay Calibration"
'
i
1302-5.31D, Rev 9
"4160V IE Bus Loss of Voltage / Degraded Grid Timing
Relay Calibration"
,
4
1303-11.10, Rev 31
"ES System Emergency Sequence And Power Transfer Test"
,
i
1303-4.16, Rev 79
" Emergency Power System"
1303-11-11, Rev 24
" Station Battery Load Test"
,
t
)
E-62. Rev 21,
" Molded Case Circuit Breaker Testing"
l
1300-3L, Rev 10
" Disassembly / Inspection of Valves for IST"
-
- .
.
.
"HPI Test"
1
j
1303-11.8, Rev 31
,
4
!
1300-3H, Rev 44
"IST of MU Pumps and Valves"
i
l
1330-3R, Rev'31
"IST of Valves During Shutdown and Remote Indication
Check"
.
l
l1300-3T, Rev 19-
" Pressure Isolation Test of CF-V4A/B, SA/B and DH-
)
i
V22A/B"
~
'
i
,
j
71302-5.17, Rev 17
"Make-up Tank Level Instrumentatio1
-
l
1302-5.18, Rev 26
"HPI/LPI Flow Level Instrumentation"
.
j
1302-5.19, Rev~18
" Borated Water Storage Tank Level Indicator"
l
1303-11.16, Rev 34
"DH Leak Test"
l-
1303-11.54, Rev 13
"LPI Test"
t
8.-
MISCELLANEOUS PROCEDURES
i
2
.ES-002, Rev.4
" Instrumentation Error Calculation and Setpoint
Determination"
- 1420-HT-1, Rev 11
" Heat Trace Repair and Replacement"
j
C-8
-
,
-
,.
~, A
a
'
2~
.
l
-
Operations Plant Manual, Section B-5, Rev 14, " Makeup & Purification"
Startup Testing, TP207-7, TCN-1 " Makeup & Purification System E.S. Test"
U-17, Rev 3
" Inspection of Zurn Floor Drain Check Valves"
l
STP No. 1-95-0021
" Decay Heat Removal Cooler Thermal
Performance / Instrumentation"
9.
VENDOR MANUAL
VTM-TM-0191
"THI-l Vendor Manual Fairbanks Morse (Colt Industries)
Emergency Diesel Generators", Rev 29
10.
SYSTEM DESIGN DESCRIPTION
SDD 772-A
" System Design Description for TMI-l Nuclear Generating
Station Electrical Cable and Raceway Routing", Rev 2
System Description
" Class IE Electrical Systems, Metropolitan Edison
Company, Three Mile Island Unit 1", Issue Date 3-23-70
j
11.
SPECIFICATION
'
" Specification for Installation of Electrical
Equipment", Rev 5
" Specification for Electrical Work", Three Mile Island
Nuclear Station Unit 1, Revised 6-1-73
'
" Instrument and Control Equipment Installation,
Revisions 1 and 2
-
m
,
M 12.
QUALITY DEFICIENCY REPORT
-QDR-DRD-1676-81
QDR 912017
_ QDR 912028
13.
NONCONFORMANCE REPORT
.
NCR 80-22
'
1
'14.
AUDIT FINDING _
AF S-TMI 04-04
C-9
.
.
.
-
-.
_
-
. - -
. _ - . . _ _ - . .
. =
- - -
- . .
. - - -
- -
15.
LICENSEE EVENT REPORT
i
.
j
LERs 93-07, 90-02, and 91-006
!
16.
MATERIAL NONCONFORMANCE REPORT
920017
940005
.
i
950009
950010
960036
4
.
17.
PLANT MODIFICATIONS
S-ECM-007 SDD-212B, Rev. 1
.
T1-CCD-113202-001, DH-V-4
Pressure Locking Modification
!
j
SDD-212F, Div. 1, Rev. 1
DHR & CF Systems Check Valves Leak Detection
l
System
,
3
S-ECM-044, Rev. 2
DHR Pumps Remote Vent Valve Actuation
T1-IS-412394-002, Rev. O
Reach Rod Extension on Decay Heat Valve
'
DH-V64
DCN 0042
Remove DH-P-1A/B Recirculation Orifices
T1-MM-412543-001, Rev 0
" Mini Mod Installation Specification for
MU-P-1B
Lube
Oil
Pumps
Power
Supply
Changeover"
-
CM-0845
" Change to Makeup Pump Low Lube Oil Trip
'
Circuit"
4
T1-MM-128087-142, Rev 0
" Installation Specification for MU-V-36 and
,
MU-V-37 Control Relocation"
"TI MM 128155 001 , Rev 1
" Solenoid Valve and Limit Switch Upgrade
- -
-
Modification"
.
TI-MM-123142-001, Rev 0
" Replacement of Decay Heat Valve 4A/B and
SA/B Motor"
TI-MM-123202-001
"Deenergize EQ Limitorque Space Heaters"
' MDD-T1-662E, Rev 2
"R.G.1.97 (Rev 3) Compliance Mods for TMI"
T1-IS-128041-001
" Rerouting Low Side Sensing Line for MU
Tank Level. Transmitter"
.
C-10
.
-
4
s-
j
'
T1-IS-412491-003
"R. G. 1.97 Instrument Loop Upgrade Mods"
T1/IS-412443-001
"MU-24-FE/FT
MU
System
Flow
Meter
.
Replacement"
-
1
T1-IS-412468-001
" Environmentally Qualified Decay Heat and
HPI Flow Transmitters"
i
TI-MM-412633-001
HPl Pipi.ng Support Upgrade
CM-0039
Makeup Pump Trip on Loss of Lube Oil
CM-0031
MU-V16A,B,C,0 Control Circuit Change
CM-105
Replace Letdown Block Orifice
CM-0004
Replace MU22-P11, P12 & P13
18.
INDIVIDUAL PLANT EXAMINATION
IPE Submittal Report (partial), March,1993
1
.
19.
GPU MEMORANDUM
5320-92-151, Rev. 1
5310-94-086
5310-94-258
-
5310-92-210
.
5310-93-187
5310-92-330.
'
5450-93 59
"
- 5810-93-0096
3300-92-0160
~5310-92-325 ~
5450-92-0065
'
5450-92-0069
5450-94-009
,
,
C320-92-)287
5310-93-209
'3310-92-1104
5360-92-266
5310-92-366
-
5310-93-124
3210-93-0003
'
5310-93-021
5450-94-0003
5310-93-098
5350-92-214
-
-
-
5350-93-143
'
C-11
.
,
e
+
_ _ _
_ _
__
. _
_ .__ _
_ . -_
_
_
_
_
5
5310-93-020, Rev. 1
,
L
5450-93-0023
'
,
5310-93-169
5310-94-024
20.
GPUN LETTER
GPUN letter 5211-84-2179
1
l
21.
TECHNICAL SPECIFICATIONS
T. S. 3.3
Emergency
Core
Cooling,
Reactor
Building
Emergency
Cooling and Reactor Building Spray Systems
T. S. 3.4
Decay Heat Removal Capability
T. S. 4.5.2.2
Low Pressure Injection
i
T. S. 4.5.4
Decay Heat Removal System Leakage
22.
FSAR SECTIONS
'
.
Section 1.4
Principal Architectural and Design Criteria
-
Section 6.0
Engineered Safeguards
Section 6.1
Section 7.0
Instrumentation and Control
Section 8.0
Electrical Systems
Section 9.5
Decay Heat Removal System
Section 9.6
Cooling Water Systems
4
'Section 14.2.2.3
Large Break Loss of Coolant Accident
-
SYSTE'M DESIGN BASIS DOCUMENTS
23.
-
--
m SDBD-T1-211-
" System'
Design
Basis
Document
for
Makeup
and
{
j
Purification System," Rev 1,11/12/96
!
- SDBD-T1-212
" Decay Heat Removal System," Reissue 1, 11/11/96"
l
' SDBD-T1-642
" Engineered Safeguards Actuation System, Rev 1, 1/8/91"
i
L
,
i
24.
SAFETY EVALUATIONS
SE-115403-008
Revision
of
Long-Term Cooling
Methods
for
'
1
12/6/94.
.
!
SE-115302-050
Impact of 92 *F River Water Temperature, 6/8/93.
.
l
25.
OTHER DOCUMENTS:
'
Test results for In-Service Testing of the DH system pumps and valves (1300-38)
- 3/88 through 10/96
~
,
i
'
C-12
,
.
y
$r-
g
.
_
,
_
__ ,
.
--
.-
.
. -
. - .
-
~ .-
.
i
APPENDIX D
/
i
LIST OF ACRONYMS
"
Alternating Current
Atomic Energy Commission
'
ANSI
American National Standards Institute
American Society of Mechanical Engineers
ATP
Abnormal Transient Procedure
BS
Building Spray
Boiling Water Reactor
BWST
Borated Water Storage Tank
CFR
Code of Federal Regulations
Direct Current
DH
Decay Heat
DHRS
Decay Heat Removal System
,
Emergency Core Cooling System'
Engineering Change Document
Engineering Evaluation Request
Emergency Feedwater
ENMCF
Event or Near Miss Capture Form
Environmental Qualification
Engineered Safeguards
,
.
Engineered Safeguards Actuation System
Engineering Work Request
.
.TMI-1 Updated Final Safety Analysis Report
- GMS-2
General Maintenance System (Rev 2) (software name)
'
GPUN
General Public Utilities Nuclear Corporation
High Pressure Injection
' IE
Inspection & Enforcement
IEEE
Institute of Electrical and Electronic Engineers
' IEN,
Inspection & Enforcement Notice
- IFI
Inspection Follow-up Item
IN
Information Notice
. IP
Inspection Procedure
In Service Testing
J0
Job Order
LAI
Licensing Action Item
Loss of Coolant Accident
Low Pressure Injection
Motor Control Center
Motor Operated Valve
- MU&P
Makeup and Purification
Net Positive Suction Head
- NRC
Nuclear Regulatory Commission
Procedure Change Request
'
PEER.
Plant Engineering Evaluation Request
. PFU
Plant FSAR Update
Plant Review Group
[
D-1
.
4,*
s
,
.
Reactor Building
RCS,
Regulatory Guide
Small Break Loss of Coolant Accident
SDBD
System Design Basis Document
Safety Evaluation Report
Safety System Functional Inspection
TCH
Temporary Change Notice
-
Technical Data Report
.
TFAAI
Technical Functions Assigned Action Item
TMI-l
Three Mile Island Unit 1
.
TS
TMI-l Technical Specifications
Unresolved Item
VDC
Volts Direct Current
Work Request
<
.
O
i
i
,-
D-2
.
O