ML20005E988
| ML20005E988 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 01/03/1990 |
| From: | Limroth D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20005E987 | List: |
| References | |
| 50-317-89-31, 50-318-89-32, GL-88-11, NUDOCS 9001120140 | |
| Download: ML20005E988 (14) | |
See also: IR 05000317/1989031
Text
i
.
--
.
.
.
U. S. NUCLEAR REGULATORY COMMISSION
Region I
E
Docket Nos.:
50-317
License Nos.: DPR-53
50-318
i
Report Nos.:
50-317/89-31
[
50-318/89-32
Licensee:
Baltimore Gas and Electric Company
Post Office Box 1475
Baltimore, Maryland 21203
Facility:
Calvert Cliffs Nuclear Power Plant, Units 1 and 2
Inspection at: Lusby, Maryland
Inspection
Conducted:
October 1 - November 30, 1989
,
Inspectors:
J. Beall, Senior Resident Inspector
S. M.eil, Projee Manager,
Approved by:
Bfh -
ddu
-
/b b
DafTd'F. Limfottfi ~ Acting Chief
Da~te
Reactor Projects Section No IA
Summary:
Areas Inspected: This was an unannounced special inspection of the licensee's
actions to assure adequate low temperature overpressure protection for the
reactor vessel.
Results:
An overall ' programmatic weakness was identified in the licensee's
tracking of corrective actions. . Examples were identified in which commitments
were overlooked and not completed.
Failure to fulfill commitments and imple-
ment timely corrective action may have resulted in a reduction in plant safety
in the area of reactor vessel low temperature overpressure protection. Appar-
ent violations of 10 CFR 50.60, 10 CFR 50.9, and 10 CFR 50 Appendix B and
Technical Specification 3.4.9.3 were identified.
.
9001120140 900103
7
ADOCK 05000317
G
PNU
.
.
.
-
.
. -
-
W
.
o
.
,
'
.
.
.
b
I
'
TABLE OF CONTENTS
l
!
Page
,
1.
Pe r so n s C o n t a c t e d . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
1
I
!
2.
Background..................................................
1
3.
Inspection Scope............................................
1
i'
4.
Chronology of LTOP Issue....................................
2
,
5.
Deficiencies in Meeting Original Commitments................
2
i
6.
Reanalysis Indicating Potentially Inadequate LTOP Not
I
Addressed.................................................
3
'
!
7.
Precursor Event Not Addressed...............................
4
8.
NRC Holds Meeting with Licensee Due to LTOP Concerns. . . . . . . .
5
,.
9.
Errors Identified in Licensee's Justification for
,
Pressurizer Manway Insta11ation...........................
5
'
10.
Conclusion..................................................
6
l
,
11.
Exit Meeting................................................
6
.
'
= Attachments
' Attachment A " Individuals Contacted During Inspection"
1
Attachment B. "LTOP Chronology"
'
Tables
L~
Table 1
Limiting Pass Addition Transient Administrative Controls
Table 2
Limiting Energy Addition Transient Administrative Controls
Table 3
Limiting Mass Addition Transient Events Analysis
'
,
i
-.
.
.
.
l
-
-
,
-
,
DETAILS
1.
Persons Contacted
During the course of this inspection, interviews and discussions were
,
!
conducted with various licensee personnel, including control room opera-
.
tors, design ennineers, and site managers. A partial list of individuals
contacted is presented in Attachment A.
2.
Background
+
The fracture toughness requirements for reactor coolant pressure boundary
materials are embodied in 10 CFR 50.60 and 10 CFR 50, Appendix G.
These
fracture toughness requirements include pressure-temperature (P-T) limits
which are established to provide an adequate margin of safety to protect
ferritic materials in the reactor coolant pressure boundary, including the
reactor vessel, from brittle fracture.
Overpressure protection measures
are provided to ensure that these limits are not exceeded during normal
operation or anticipated operational occurrences.
As reactor coolant system temperature decreases, the reactor vessel's
susceptibility to brittle fracture markedly increases, thus, in low
temperature ranges, the adequacy of overpressure protection is particu-
larly important.
To protect against reactor vessel brittle fracture
in the susceptible low temperature range, licensees were required to
provide additional Low Temperature Overpressure Protection (LTOP) measures
to prevent anticipated operational occurrences, such as an inadvertent
safety injection pump start, from causing pressure transients that would
result in conditions exceeding the associated P-T limits.
,
3.
Inspection Scope
During this inspection period, licensee actions taken to resolve defici-
encies and concerns pertaining to plant P-T limits and low temperature,
'
overpressure protection (LTOP) systems, previously identified in NRC
Inspection
Report Nos.
50-317/88-05;
E0-318/88-06
and
50-317/88-12;
50-318/88-12, were examined.
The items initially examined included the
June 28, 1988 inadvertent lifting of a power operated relief valve (PORV)
on Unit 1, while heating up the reactor coolant system, and necessary
modifications to the P-T limits contained in Technical Specification
3/4.4.9, " Pressure-Temperature Limits."
.
!.
..
-
.
e
2
4.
Chronology of LTOP Issue
The Unit 2 Operating License (DPR-69) was issued on November 30, 1976 and
'
contained a license condition requiring a permanent, NRC-approved LTOP
,
system prior to startup from the first refueling outage.
In July 1977,
the licensee submitted the LTOP plan, applicable to both Units 1 and 2.
In August 1978, that plan was approved and incorporated into the Technical
Specifications of both units.
Both units operated for several fuel
,
cycles, which necessitated P-T curve revisions due to reactor vessel
radiation embrittlement.
During the last three years, the licensee
developed the revised P-T curves, the NRC issued new guidelines and
requirements based on industry experience (Generic Letter 88-11), and the
licensee took measures to respond to Generic Letter (GL) 83-11.
During this inspection, several deficiencies and omissions were identified
in the licensee's actions to assure adequate LTOP. This led to a meeting
with the NRC staff on November 27, 1989.
Additional deficiencies were
identified in the licensee's approach on the following day.
A more
detailed chronology is presented in Attachment B,
5.
Deficiencies in Meeting Original Commitments
The original LTOP measures, as evaluated by the NRC, utilized a combina-
tion of administrative controls, hardware improvements. Technical Specifi-
cations (TS) and operator training to ensure that the P-T limits estab-
lished to provide protection against reactor vessel brittle fracture would
not be exceeded.
The licensee's approach was intended to apply to the
period of operation from 0-10 Effective Full Power Years (EFPY), and was
not designed to provide protection after 10 EFPY of operation.
The licensee established the administrative controls based upon the
results of the analyses for the most limiting pressure transient producing
incidents at low RCS temperature conditions.
The two transients of
concern were:
1) an inadvertent high pressure safety injection (HPSI)
'
pump actuation (Limiting Mass Addition Transient (LMA)) and 2) the reactor
coolant pump start transient during plant conditions where steam generator
water temperature is higher than reactor vessel water temperature (Limit-
ing Energy Addition Transient (LEA)).
The principal LTOP administrative
controls to which the licensee committed in its July 21, 1977 submittal,
are described in Table 1 for the design LMA and in Table 2 for the design
LEA.
Similarly, the results of the design LMA events analyses are pro-
vided in Table 3.
l
i
,
-
.
.
.
3
L
The July 21, 1977, licensee submittal stated that most of the administra-
tive controls were in place, including those involving the HPSI system.
As documented in a later NRC inspection report (IR 50-317/88-05; 50-317/
s
88-06), certain LTOP-based administrative controls were found not to be in
'
place.
Specifically, the licensee had not prohibited either the testing
!
of the emergency core cooling system (ECCS) with the plant solid or the
startup of the shutdown cooling system when steam generator temperature
was above 220* F.
The licensee modified plant procedures to prohibit ECCS
testing while in a solid condition following NRC identification of the
'
deficiency in the above inspection report.
However, during the current
i
inspection, the inspector identified that certain of the original 1977
commitments had still not been met.
Sustained operation without the
committed controls indicates that adequate LTOP may not have existed
during the period and constitutes an apparent violation of 10 CFR 50.60,
" Acceptance Criteria for Fracture Prevention Measures for Light Water
Nuclear Power Reactors for Normal Operation."
(Details are provided in
the attached Tables referenced above.) Further, these actions constituted
the basis for the August 7,1978 NRC SER authorizing deletion of the LTOP
,
license condition (that is, the outage startup hold), this is an apparent
violation of 10 CFR 50.9(a), " Completeness and Accuracy of Information."
6.
Reanalyses Indicating Potentially Inadequate LTOP Not Addressed
On January 21, 1987, the licensee staff submitted new P-T curves (reflec-
ting reanalysis) to the Plant Operations and Safety Review Committee
(POSRC) for approval.
The curves were developed to cover up to 12 EFPY
'
since Unit I was felt to be approaching 10 EFPY and would transition to
the more restrictive TS curves for 10-40 EFPY, During the current inspec-
tion, licensee representatives stated that the 10-12 EFPY curves were
approved for use by engineering personnel for activities such as determin-
ing relief valve setpoints. The 10-12 EFPY curves were not submitted for
NRC review and approval as were the TS curves. The in situ setpoints did
not prevent transient pressures above the TS 10-40 EFPY curves but were
intended to meet the less restrictive 10-12 EFPY curves not reviewed by
,
the NRC,
,
Another concern involved the memo to the POSRC that accompanied the
proposed 10-12 EFPY curves in that certain TS controlled parameters needed
to change.
Specifically, the maximum PORV setpoint that would provide
adequate protection for the 12 EFPY cooldown curve was 400 psia (TS 3.4.3.a allows the PORV setpoint to go up to 450 psig). Also, the maximum
differential temperature between the steam generator water and the reactor
vessel water for an RCP start, when RCS temperature is below 275
F, to
ensure adequate LTOP for the LEA is less than 30' F (TS 3.4.1.3 allows a
.
maximum differential temperature limit of less than 46* F). These changes
were made but no TS change was submitted.
.
i
l
l
l
'
L.
L
,
-
.
,
'.
I
i
4
,
!
'
Subsequently,
on March 26, 1987,
the
licensee completed a thermal-
hydraulic reanalysis to model RCS response to various LTOP transients.
In
this reanalysis, it was determined that significantly higher RCS pressures
would occur as a result of various mass addition transients.
These
pressures would exceed the P-T limits for both the 0-10 EFPY P-T curves
-
and the 10 - 40 EFPY P-T curves for significant portions of the RCS
,
temperature range over which LTOP is required.
These mass addition
<
transient results are included in Table 3.
>
The licensee took no mitigating or corrective actions at the time of the
'
reanalysis to address the potential inadequacy of in place LTOP measures.
These concerns were raised in a later NRC inspection (IR 50-317/88-05;
,.
50-318/88-06) which identified that reanalyzed LMA peak pressures were
l:
significantly higher than the design basis LTOP limits.
The licensee
again took no mitigating or corrective actions.
Failure to implement
corrective action in response to the reanalysis is a violation of 10 CFR 50 Appendix B Criterion XVI, " Corrective Action."
'
7.
Precursor Event Not Addressed
During a Reactor Coolant System (RCS) heatup on June 25, 1988, an inad-
vertent RCS relief valve lif t occurred.
System pressure was about 390
-
psia (setpoint 400 psia +/- 16 psi) and temperature was about 303* F.
This event was documented in Inspection Report 50-317/88-12; 50-318/88-12
which identified the cause of the event as licensed operator error.
Operators were too focused on maintaining pressure above the minimum
'
required to operate reactor coolant pumps and inadvertently allowed RCS
pressure to rise to the relief setpoint.
The inspection report noted that the licensee had ongoing efforts to
remove excessive P-T conservatisms to improve ease of plant maneuvering
.
and to reduce the temperature at which the RCS relief valve setpoint is
increased to its normal operating value (about 2400 psia) to below 330' F.
'
At 300
F, the P-T limit for LTOP is about 900 psia during a heatup. At
330' F, which is the temperature where the relief valve setpoint was
permitted to be changed, the P-T limit is 1200 psia during a heatup.
In
neither case would a setpoint of 2400 psia have afforded adequate over-
pressure protection.
The event, then, was a precursor reemphasizing the
need for adequate LTOP.
Followup inspection during the current period,
however, was unable to identify any evidence that licensee corrective
actions were responsive to this precursor event.
This event was a challenge to the LTOP system, which actuated automati-
cally, as designed.
No report was made to NRC regarding this event.
Failure to submit a Special Report concerning this event within thirty
days is an apparent violation of TS 3.4.9.3 action statement c.
i
.
--
m
.
m-
- .
- -
}
.
.
.
5
i
8.
NRC Holds Meetina With Licensee Due to LTOP Safety Concerns
Generic Letter (GL) 88-11. "NRC Position on Radiation Embrittlement of
Reactor Vessel Materials and Its Impact on Plant Operations," was issued
on July 12, 1988.
The licensee conducted reanalysis of P-T limits and
submitted a TS change request on October 26, 1989 in response to GL 88-11.
The inspector reviewed the analyses and noted that the results were
,
similar to the March 26, 1987 analysis results.
That is, certain tran-
sients could result in unacceptably high RCS pressures (see Section 6).
Discussions with the licensee indicated that the 10-12 EFPY curves of P-T
limits were being used for RCS relief valve setpoints. The inspector also
noted that not all administrative controls were in placo. At that time,
the Unit 1 pressurizer manway was not installed, providing a large RCS
vent such that no overpressure transient was possible.
Unit 2 was
defueled. The licensee stated that Unit 1 pressurizer manway reinstalla-
tion was imminent but agreed to first meet with the NRC to discuss LTOP
concerns prior to its installation.
The
licensee met with
the NRC
staff at
the NRC:NRR offices on
November 27, 1989. At that meeting, the licensee indicated an investiga-
tion was underway to determine why the originally committed administrative
controls were not currently in place. This investigation was not expected
to be completed until' February, at the earliest. The licensee stated that
two additional actions were being taken to ensure adequate Unit 1 brittle
fracture protection.
These actions were the lowering of the PORV lift
pressure setpoint to 384.4 psia and the throttling of HPSI pump flow to a
maximum of 350 gpm.
One of the original LTOP administrative control
requirements was for the HPSI header isolation valves to be physically
locked shut.
The licensee stated that the valves were administratively
locked shut and that no single failure could cause the valves to open and
defeat the HPSI flow throttle requirement.
Based upon the licensee's commitments and reevaluation of Unit 1 LTOP, the
NRC staff agreed to the reinsta11ation of the Unit 1 pressurizer manway
and the subsequent pressurization of the Unit 1 RCS.
The licensee also
committed to submit the documentation of LTOP adequacy prior to heating
the Unit 1 RCS above 325* F, the PORV enable temperature.
The Unit 1
pressurizer manway was installed on November 27, 1989.
9.
Errors Identified in Licensee's Justification for Pressurizer Manway
Installation
The inspector conducted an on-site verification of interim LTOP measures
i
on November 28,1989 and identified two additional deficiencies.
The
i
first deficiency, identified concurrently by the licensee, was that no
'
I
such provision as " administrative 1y locked shut" applied to the HPSI
header isolation valves.
Neither physical controls (such as locks), nor
administrative controls (such as tags), were in place to prevent or
restrain valve rotion,
l
i
.
.
.
-
--
,
m
r
.
.
'
9
i
6
,
l
The second deficiency involved the identification, by the inspector, of a
.
single failure scenario involving an unthrottled HPSI pump (LMA tran-
sient).
The alternate boration path available to the operator involved
'
using the HPSI pump which had its power supply breaker racked in and
throttling HPSI flow to 350 gpm.
The eight HPSI header isolation valves
receive automatic full open signals during certain accident situations.
Of the two possible initiating signals (low pressurizer pressure and high
containment pressure), only one was blocked (low pressurizer pressure).
The other signal (high containment pressure) was not blocked. An inad-
vertent or spurious high containment pressure signal initiated during HPSI
use for boration would have caused all HPSI valves to leave the shut or
throttled position and go full open.
The same signal would also have
started the HPSI pump which had the power supply breaker racked in if the
pump was inadvertently removed from the " pull to lock" position.
The licensee acknowledged the inspector's concerns and placed the hand
switches for the HPSI header isolation valves in the " pull to override"
position.
This position prevents any electronic signal from initiating
automatic valve motion while allowing the operator to reposition the valve
from the control board.
The errors identified in the licensee's justification for the reinstalla-
tion of the pressurizer manway are additional examples of weakness in the
licensee's resolution' of the LTOP issue.
10. Conclusion
The failure to meet original operating license LTOP commitments, to
address reanalysis results showing LTOP deficiencies, and to address an
LTOP precursor event, reduced the level of protection for a low tempera-
ture overpressure event.
These failures represent apparent violations;
details are found in the individual report sections.
,
11. Exit Meeting
,
An exit meeting was held on November 22, 1989. Attendees are listed in
Attachment A,
F
7-
.
-
.
,
,
..
ATTACHMENT A
Individuals Contacted During Inspection
Baltimore Gas and Electric Company
A. B. Anuje, Supervisor, Quality Audit Unit
- T. L. Cook, Senior Engineer, Nuclear Engineering Unit
G. C. Creel, Vice President, Nuclear Energy
- P. T. Crinigan, General Supervisor, Chemistry
- C. H. Cruse, Manager, Nuclear Engineering Services Department
- R. E. Denton, Manager, Quality Assurance Department
,
- D. S. Elkins, Senior Engineer, Nuclear Engineering Unit
- -
- R. P. Heibel, General Supervisor, Quality Assurance
- J. R. Lemons, Manager, Nuclear Outage Management
- W. J. Lippold, General Supervisor, Technical Services Engineering
- B. S. Montgomery, Principal Engineer, Licensing
P- A. Pieringer, Supervisor, Independent Safety Evaluation Unit
.
- R. B. Pocha, Contract Employee Licensing
- L. B. Russell, Plant Manager
- J. E. Thorp, Senior Engineer
- J. 9. Wood, Senior Auditor. Quality Assurance
Nuclear Regulatory Commission
- J. E. Beali, Senior Resident Inspector
- R. A. Capra, Director, Project Directcrate I-1, NRR
- C. Y. Cheng, Chief, EMTB, NRR
- T. E. Collins, Section Chief, SRXB, NRR
- B. J. Elliot, Senior Engineer, EMTB, NRR
- M.
A.- Hunemuller, Operations Engineer, LPEB, NRR
- R. C. Jones, Jr. , Acting Chief, SRXB, NRR
- C. Y. Liang, Senior Engineer, SRXB, NRR
E. Throm, Engineer, RES
- P. N. Randall, Engineer, EMTB, NRR
D. F. Limroth, Project Engineer, Region I
l
- S. A. McNeil,' NRR Project Manager, Calvert Cliffs
V. L. Pritchett, Resident Inspector, Calvert Cliffs
- J. T. Wiggins, Chief, Reactor Prohects Branch No. 1, Region I
- Denotes those present at the exit meeting on November 22, 1989.
- Denotes those present at the public meeting on November 27, 1989.
l
W
L
- ,
'
.
MA
ATTACHMENT B
LTOP Chronology
November 30, 1976
Unit 2 OL (DPR-69) issued, contains license condition
requiring permanent, NRC-approved LTOP system prior to
startup from first refueling outage.
July 21, 1977
Licensee submits LTOP plan, containing a combination of
administrative
controls,
hardware
improvements,
TS
revisions and operator training.
August 7, 1978
NRC issues SER documenting completion of staff review of
licensee submittal
and
removing LTOP Unit 2
license
condition.
January 21, 1987
Licensee staff submits new P-T curves to site Plant Opera-
tions -and Safety Review Committee (POSRC) for review and
approval.
The curves are for the Unit I reactor vessel
for the period of up to 12 EFPY, The POSRC did not take
issue with the proposed approach. No TS submittal is made
and the curves are not NRC reviewed.
March 26, 1987
Licens'ee completes a reanalysis indicating that P-T limits
could be substantially exceeded despite existing LTOP
measures in certain mass addition transients.
May 11, 1988
NRC issues inspection report (50-317/88-05; 50-318/88-06)
which identifies that the transition to the more restric-
tive 10-40 EFPY curves poses problems needing licensee
resolution with respect to LTOP. The report also documents
that some of the analysis results (see March 26, 1987
above) need to be addressed by the licensee,
June 28, 1988
Unit 1 PORV lifts at about 390 psia during an RCS heatup
due to operator error which allowed pressure to drift high.
The PORV setpoint was 400 psia (+/- 16 psi) as required for
LTOP.
This was a precursor event demonstrating the need
for adequate LTOP.
July 12,1988
NRC issues Generic Letter (GL) 88-11, "NRC Position on
Radiation Embrittlement of Reactor Vessel Materials and Its
Impact on Plant Operations."
The GL provides new guide-
lines and requires reanalysis of P-T limits.
September 29, 1989 Licensee reanalysis in response to GL 88-11 again shows
i
that P-T limits can be exceeded during certain mass addi-
tion
The
results
are
similar
to
the
March 26, 1987 reanalysis results,
i
t.
i
e.
.
L.
..
,
Attachment B
2
i
<
L
i
i
October 26, 1989
Licensee submits TS amendment in response to GL 88-11.
November 27, 1989
The licensee meets with NRC staff to discuss status of LTOP
corrective measures in response to the findings of the
ongoing NRC inspection (50-317/89-31; 50-318/89-32).
On
'
the basis of the licensee's actions and commitments, the
'
staff agrees to the reinstallation of the Unit 1 pressur-
g
izer manway.
November 28, 1989
Inspector identifies deficiencies in the licensee's correc-
tive measures; namely, the valves to be used for manual
throttling still have the capability to fully open auto-
'
matica11y.
HPSI pump use with full open valves is the
major LTOP risk scenario.
The licensee then placed the
,
affected valves in " pull to override" to resolve the
concern.
,
R
l
l
i
'
.
-
.
.
,
TABLE 1
Limiting Mass Addition Transient Administrative Controls
LTOP Controls Design (7-21-87)
In Place (3-31-88)
In Place (11-22-89)
HPSI Pumps
<320F: 1 Disabled *
<27SF: 2 Disabled
<300F: 2 Disabled
3220F: 2 Disabled
Third caution tagged Third caution tagged
<160F: All Disabled with switch in pull- with switch in pull-
~
to-lock.
Pump use
to-lock.
Pump use
permitted at all
permitted at all
temperatures.
temperatures.
RCS solid and
(Not Determined)
Valve position
Isolation
<200F: valves
uncontrolled
Valves
locked shut
Charging
RCS solid: Those
(Not Determined)
RCS solid and <200F:
Pumps
not required are
Two caution-tagged
disabled. Only
with switches in
one normally
pull-to-lock,
required.
ECCS Testing
Prohibited *
Not Prohibited
Prohibited
(RCS Solid)
Pressurizer
3 60% steam
Same as design
Same as design
Steam Volume
volume: precaution
to minimize time
with RCS solid
' Disabled by racking out master breaker, caution tagging and placing handswitch
in pull-to-lock,
.
.
..
<
TABLE 2
Limitino Energy Addition Action Transient Administrative Controls
Desian (7-21-77)
In Place (3-33-88)
In Place (11-22-89)
Prohibited
Allowed
Allowed - limits on
' Initiation with
initiation if RCS
pressure <270 psia
temp >220 F.
or RCS temperature
<300 F.
Prohibited
Not Determined
Prohibited - RCP
pump start with
start not allowed
secondary to
with RCS solid.
primary temp
differential
> 5* F (RCS
lolid).
Pressurizer
Disabled and
Not Determined
Disabled and Tagged
'
Heaters
Tagged
'
(RCS Solid)
y %
g)
'
+
.
l,
_-l.
j
4
,
.
f
l.
l
-
i
TABLE 3'
l
'
Limiting Mass Addition Transient Events Analyses-
f
j
.
Design
Reanalysis
Reanalysis
P-T Limit-
P-T Limit
!
(
(7-21-77)-
(3-26-87)
(9-29-89)
0_-10 EFPY
10-40 EFPY
l
460 psia.
424 psia-
390 psia
'--
--
Setpoint
'
'
I HPSI &
540 psia.
N/A
N/A
540 psia.
300 psia
.!
3 CCG Pumps
(T = 160F)
l
(RCS Solid)
.
'
'
t
1 HPSI &
N/A
740 psia
770 psia
380 psia
300 psia-
3 CCG Pumps
-
,
-
(T = 100F)-
i
(RCS Solid)
'
.;
,.
1 HPSI &'
N/A
740 psia
770 psia-
610 psia
300 psia
-[
.
3 CCG Pumps
U
(T = 180F)-
i
-(RCS Solid)
-i
'
2 HPSI &s
870 psia
N/A
N/A
860 psia
300 psia
l
3 CCG Pumps
(7-21-77
-(T = 220F) _
submittal.
.
[
.(RCS Solid)
stated 870)
L
.
.
'2 HPSI &'
N/A-
850 psia
N/A
1230 psia -
360 psia
-3'CCG
(T = 280F)
(RCS bubble)
'
i
k
>
I
>
I
s
r
,
l
J
>