ML20090A602
| ML20090A602 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 02/24/1992 |
| From: | GENERAL PUBLIC UTILITIES CORP. |
| To: | |
| Shared Package | |
| ML20090A577 | List: |
| References | |
| NUDOCS 9203030026 | |
| Download: ML20090A602 (39) | |
Text
._
1
)
Three Mile Island Nuclear Plant, Unit 1 (THi-1)
Operating License No. DPR-S0 Docket No. 50-289 Technical Specification Change Request No. 204 1.0 Proposed Technical Specification Chance Reauest GPU Nuclear requests that the following pages of the THI-l Technical Specification be replaced as indicated below:
Replace Paaes:
it, vii. 3-18b, 3-1Bc, 3-18d, 3-18e, 3-41a, 3-41b, 4-29, 4-30, 4-31, 4-32, 4-33, 4-39, 4-42, 4-43, 4-44, 4-47, 4-58, and 4-68.
Replace Fiaures:
2.1-1, 2.1-2, 2.1-3, 2.3-1, 2.3-2, 3.1-1, 3.1-2, 3.1-2a, 3.1-3, 3.5-2m, 3.5-1, 3.5-2, 3.5-3, 3.11-1, 4.17-1.
Delete Paaes: viii, 4-34, 4-34a, 4-34b, and 4-72.
2.0.
Reason for Chance The changes included in this submittal are as follows:
2.1 CHANGES TO INCORPORATE EXEMPTION FROM THE 10 CFR 50, APPENDIX J REQUIREMENTS WHICH LINK THE SCHEDULE FREQt":NCY OF INTEGRATED LEAKAGE RATE TESTS TYPE A TESTS (ILRT), Wl?H THAT OF THE INSERVICE INSPECTION (151) PROGRAM REQUIRED BY 10 CFR 50.55A.
GPUN submitted a request for exemption from the requirements of 10 CFR Part 50 Appendix J, Section Ill D.l(a) on August 30, 1990.
Section Ill.D.l(a) requires the performance of a set of three Type A tests after the preoperational leakage rate tests, at approximately equal intervals during each 10-year service period, such that the third test of each set is conducted when the plant is shutdown for the 10-year plant inservice inspections required by Section 50.55a. The exemption would allow the third Type A Containment Integrated Leak Rate Test (ILRT) of the current 10-year service period (and subsequent periods) to be uncoupled from the 151 schedule.
This change is needed in order for the THI-l Technical Specifications to conform with the exemption granted by the NRC on February 25, 1991.
The link between ILRT and ISI schedules is not addressed in the current Bases for Specification 4.4; however, this change adds appropriate language to the Bases, describing exemption from the Appendix J requirement which connects the two schedules, in addition, paraphrased Appendix J wording is being administrative 1y removed from Technical Specification 4.4 to eliminate redundancy in the Technical Specifications and thus improve clarity.
9203030026 920224 PDR ADOCK 05000289 P
pyg l
C311-91-2130 2.2 CHANGES TO ALLOW USE OF AN EQUIVALENT VERIFICATION METHODOLOGY FOR ES VALVE SURVEILLANCE.
GPU Nuclear requests the addition of an alternate methodology for verification of valve travel in the performance of surveillance tests prescribed in Technical Specifications 4.5.1.1.b, 4.5.1.2.b.
4.5.2.4.b 4.5.3.1.b.2 and 4.5.3.2.b.
The alternate method, which GPU Nuclear finds is an acceptable mechanism for secondary verification of valve travel during the surveillance test, would utilize control board operating lights on a different control room aanel which is energized from a separate power supply and actuated
)y separate limit switch contacts. The plant computer uses this method for position indication, also.
By use of this alternate methodology, the intent of the specification is preserved.
2.3 EDITORIAL CHANGES AND CORRECTIONS TO IMPROVE CLARITY AND CONSISTENCY, AND TO CONFORM THE AB0VE CHANGES TO THE AMENDMENT APPROVING THIS TECHNICAL SPECIFICATIONS CHANGE REQUEST (TSCR 204).
Cha.iges in this category include the following:
a.
Page numbers are being added to all 8+1/2" x 11" figures throughout the Technical Specifications. Currently no page numbers are assigned, except for Figures 3.1-2a (p 3-9b) and 3.11-1 (p 3-56b).
b.
Redesignating lettered or numbered subsections with numbers or letters respectively to provide consistency of format within a given specification.
3.0 infety Evaluation Justifyina the Chanae 3.1 CHANGES NEEDED IN THE THI-l TECHNICAL SPECIFICATIONS FOR CONFORMANCE WITH THE EXEMPTION FROM REGULATORY REQUIREMENT COUPLING THE SCHEDULE FREQUENCY OF ILRT WITH THAT OF THE ISI PROGRAM.
In a letter dated February 25, 1991 the NRC approved an exemption for THI-l from a portion of the requirements stated in 10 CFR Part 50 Appendix J, Section III.D.!(a).
This section of Appendix J requires the performance of a set of three Type A tests, at approximately equal intervals during each 10 year service period, such that the third test of each set is conducted when the plant is shutdown for the 10 year plant inservice inspections required by Section 50.55a.
This exemption allows the third Type A test of the current 10 year service period (and subsequent periods) to be uncoupled from the ISI schedule. _
C311 91 2130 Performance of the third Type A test and the 10 year 151 in non-concurrent outages has no effect on safety inasmuch as the pur)oses of the Type A tests and the 10 year ISI are independent of eac1 other and the performance of one does not directly affect the other.
The operability of the Reactor Building containment and those components which are tested in accordance with 10 CfR 50, Appendix J will continue to be verified in accordance with acceptable schedules, methods, and acceptance criteria consistent in all other respects with Appendix J except for uncoupling the test schedule from that of the ISI Program.
Therefore, uncoupling the 10-year 151 and the third Type A test of a 10 year service period is justified.
3.2 CHANGES TO ALLOW USE OF AN EQUIVALENT VERIFICATION ME1H000 LOGY FOR ES VALVE SVRVEILLANCE.
The changes to Technical Specifications 4.5.1.1.b, 4.5.1.2.b.
4.5.2.4.b, 4.5.3.1.b.2 and 4.5.3.2.b provide an additional alternative methodology for performance of the verification of valve travel during ES valve tests.
The change serves to enhance safety and reduces potential operator error.
Safety is enhanced because the need to attach test instrumentation to spare contacts or to have operations personnel stationed locally is eliminated.
Yet the intent of the specification is preserved through the provision for a secondary means of verification of valve travel.
No additional possibility of component malfunction is created by this change, and no new accident scenario is created nor existing accident scenarios impacted.
This change provides for additional safety during performance of the test and eliminates the potential risk of failing to reassemble or im)roperly disconnecting test instrumentation previously used.
Tierefore, the alternative provides an overall improvement in safe plant operations.
3.3 EDITORIAL CHANGES AND CORRECTIONS:
For ease in reviewing the administrative changes proposed by TSCR No.
204, the changes being made by this request are discussed below on a page by page basis:
p 11 Reflects changes to pages being renumbered, p vii Table of Contents, page numbering added for Figures, p viii Deleted, original information moved to p vii.
p 3 9b Figure 3.1-2a title retyped for clarity p 3 18b Pages renumbered due to Figure pagination p 3-18e p 3-18d p 3 18e - - -,.
~-
'C311 91-2130 p 3 41a T.S. 3.6.8.2 references to Section 4.4.1.7.1 changed to 4.4.1.7.a.
p 3 41b Bases; comma added in paragraph 3.
p 4 29 Original Section 4.4.1.1.3 deleted, wording in T.S. paraphrased that contained in 10 CFR 50 Appendix J sections Ill.A.S.a.1 and b.1, and testing at Pt is not performed at IMI-1. Old 4.4.1.1.4 renumbered as 4.4.1.1.3 and moved from p 4 30 to 4 29.
New 4.4.1.1.3 revised to delete para >hrased Appendix J wording and use of the acronym ILRT to replace tie words " Integrated Leak Rate Test.'
p 4-30 Original Section 4.4.1.1.5 renumbered as Section 4.4.1.1.4.
New wording reflects the NRC exemption granted February 25, 1991, p 4-30 Original Section 4.4.1.1. 6 renumbered as Section 4.4.1.1.5 and Appendix J paraphrased wording deleted, as well as, acceptance criteria revised to reflect high pressure test symbols of 10 CFR 50 Appendix J lll.A.5.a.2 and b.2 rather than testing at Pt which is not performed at THI 1.
p 4 30 Original Section 4.4.1.1.7 renumbered as Section 4.4.1.1.6 and Appendix J paraphrased wording (Ill.A.I.a) deleted, as well as, revision to reflect testing at Pa rather than at Pt which is not performed at THI 1.
p 4-30 Original Section 4.4.1.1.8 renumbered as Section 4.4.1.1.7 and p 4-31 revised to reflect commitment to Appendix J reporting requirements, p 4-30 Section 4.4.1.2.1 reference to 8 4.1.2.5.f revised to 4.4.1.2.5.c.
p 4-30 Sections 4.4.1.2.2 and.3 are changed to reflect use of the acronym LLRT and revision of reference to 4.4.1.2.5.b changed to 4.4.1.2.5.a.
Paraphrased Appendix J wording in paragraphs b, c, and d deleted (See Appendix J lil.c.1).
p 4 31 Section 4.4.1.2.5 revised to reflect use of actrMm LLRT and exceptions to Appendix J frequency deleted where applicable, p 4-31 Section 4.4.1.4 revised to reflect use of the acronym ILRT.
p 4-31 Section 4.4.1.5 reference to 4.4.1.1.6 changed to 4.4.1.1.5, and revised to reflect the use of the acronyms LLRT and ILRT.
p 4-31 Section 4.4.1.6 Subparagraph numbering changed to letters for p 4-32 consistency within Specification 4.4.
i p 4-32 Section 4.4.1.7 Subparagraph numbering changed to letters for consistency within Specification 4.4.
[
l l
4
'C311-91-2130 p 4-32 Bases, wording added to reflect NRC exemption granted as well as p 4-33 use of the acronym ILRT and LLRT, new references added, p 4-39 Sections 4.5.1.1.b and 4.5.1.2.b revised to reflect an alternate methodology for secondary verification of valve travel during the surveillance test.
p 4-42 Section 4.5.2.2.c typographical correction: effects changed to affects.
p 4-42 Section 4.5.2.4.b revised to reflect an alternate methodology for secondary verification of valve travel during the surveillance test.
p 4-42 Bases, typographical corrections: storgage changed to storage, and lines c. tanged to line, p 4-43 Section 4.5.3.1.b moved from p 4-44 to 4-43 and Subparagraph 2 revised to reflect use of an alternate methodology for secondary verification of valve travel during the surveillance test, p 4-44 Section 4.5.3.2.b revised to reflect use of an alternate methodology for secondary verification of valve travel during the surveillance test.
p 4-44 Bases, wording added to clarify the flow for testing the de'.tvery capability of the Reactor Building Spray pumps, p 4-47 Editorial corrections of typographical errors in paragraph 4.6.3.a (1), misplaced wording relocated; and, typographical correction in paragra,nh 4.6.3.a (3): Verify.
p 4-58 T.S. 4.15 changes to the Main Steam ISI specification to reflect the THI-l inservice inspection program, applicable code requirements, and the correct weld identification numbers, in addition, the third paragraph of section 4.15.1 was deleted and the Tech. Spec. Bases expanded to include potential for future changes in the inspection interval. An appropriate reference to the UFSAR has been added.
Figures Currently page numbers are not assigned to most figures throughout the Technical Specifications. Assignment of page numbers will clarify the intended location of the figures.
This has been done for all figures except Figures 5.1, 5.2, and 5.3 which are 11" x 17" fold-outs in Section 5.0.
In summary, the changes described above are administrative in nature and are intended to make corrections that are justified and appropriate in accordance with NRC regulations.
No unreviewed safety question is involved in the changes as reflected in this Technical Specification Change Request.
C311-91 2130 4.0 Ro Sinnificant Hazards Considerati m GPUN has determined that this Technical Specification Change Request (TSCR) poses no significant hazards as defined by the NRC in 10 CFR 50.92. This change is considered to be administrative in nature and does not involve significant hazards consideration as evaluated below.
4.1 Operation of Three Mlle Island Nuclear Station, Unit 1, in accordance with this TSCR would not involve a significant increase in the probability or consequences of an accident previously evaluated because the proposed Technical Specification change does not modify or create any accident initiating condition.
This change provides administrative changes and corrections to the Technical Specifications; conforms the Technical Specifications to the NRC granted exemption from the ISI schedule requirements of 10 CFR 50.55a; and, provides additional flexibility in the performance of surveillance test confirmations of valve travel. The changes do not result in any condition contrary to the requirements of 10 CFR 50 Appendix J.
Therefore, the changes do not impact on nuclear safety or safe plant operations.
4.2 Operation of Three Mile Island Nuclear Station, Unit 1, in accordance with this change would not create the possibility of a new or different kind of accident from any accident previously evaluated because the proposed Technical Specification change does not modify or create any accident initiating condition.
The proposed changes will result in Technical Specification requirements that meet or exceed the requirements of 10 CFR 50 Appendix J.
4.3 Operatior af Three Mile Island Nuclear Station, Unit-1, in accordance with this change would not involve a significant reduction in a margin of safety because the margins of safety, as described in existing Technical Specification bases, remain unaffected by this change request; further, the margins of safety deh 9d in the SAR are not impacted by this change.
The Commission has provided guidelines pertaining to the application of the three standards by listing specific examples in the Federal Register (48FR14870).
This proposed change is considered to be in the same category as examples (i), (ii), or (iv) of the " Amendments Not likely to Involve Significant Hazards Consideration" from that listing.
- Thus, operation of the facility in accordance with the prc, posed amendmant involves no significant hazards considerations.
5.0 Imolementation it is requested that the amendment authorizing this change become effective upon issuance and shall be impirmented within thirty (30) days of receipt.
6-
TABLE OF CONTENTS itC110D EA22 2
SAFETY llMITS AND LIMITlhG S4[ETY SYSTEM _SETTINM 2-1 2.1 Safety Umits.- Reactor Core 2-1 2.2 hafety Limits. Reictor System Pressure 2-4 2.3 Liniitina Safety System Settinas. Protect 10J1 IntitCecDiti.1911 2-5 3
LIMITING CONDITIONS FOR OPERATION 3-1 3.0 General Action Reauirementi 3-1 3.1 Reactor Coolant SY11tD 3-la 3.1.1 Operational Components 3-la 3.1.2 Pressurization, Heatup and Cooldown Limitations 3-3 3.1.3 Minimum Conditions for Criticality 3-6 3.1.4 Reactor Coolant System Activity 3-8 3.1.5 Chemistry 3-10 3.1.6 Leakage 3-12 3.1.7 Moderator Temperature Coefficient of Reactivity 3-16 3.1.8 Single Loop Restrictions 3-17 3.1.9 Low Power Physics Testing Restrictions 3-18 3.1.10 Control Rod Operation 3-18a 3.1.11 Reactor Internal Vent Valves 3-18c 3.1.12 Pressurizer Power Operated Relief Valve (PORV) and Block Valve 3-18d 3.1.13 Reactor Coolant System Vents 3-18f 3.2 title.gg._and Purification & Chemiral Addition Systemt 3-19 3.3 f.meraency Core Coolina. Reactor Buildina Emeraency Coolina and Reactor Buildina Snray Systems 3-21 3.4 Decay Heat Removal Canability 3-25 3.4.1 Reactor Coolant System Temperature Greater than 250*F 3-25 3.4.2 Reactor Coolant System Temperature 250*f or less 3-26 3.5 Instrumentation Systems 3-27 3.5.1 Operational Safety Instrumentation 3-27 3.5,2 Control Rod Group and Power Distribution Limits 3-33 3.5.3 Engineered Safeguards Protection System Actuation Setpoints 3-37 3.5.4 Incore Instrumentation 3-38 3.5.5 Accident Monitoring Instrumentation 3-40a 3.5.6 Chlorine Detection Systems 3-40f 3.6 Reactor Bu11dina 3-41 3.7 Unit Electrical Power SY11ED 3-42 3.8 fuel Loadina and Refuelina 3-44 3.9 Rele.ted 3-46 3.10 Miscellaneous Radioactive Materials Sources 3-46 3.11 Handlina of Irradiated fuel 3-55 3.12 RgAg. tor BuildiDa Polar _ Qang 3-57 3.13 Secondary System Activity 3-58 3.14 f.Innd 3-59 3.14.1 Periodic Inspection of the Dikes Around IMI 3-59 3.14.2 Flood Condition for Placing the Unit in Hot Standby 3-60 3,15 Air Treatment System 1 3-61 3.15.1 Emergency Control Room Air Treatment System 3-61 3.15.2 Reactor Building Purge Air Treatment System 3-62a 3.15.3 Auxiliary and fuel Handling Building Air Treatment System 3-62c 3.15.4 fuel Handling Building ESF Air Treatment System 3-62e 11 Amendment No. 57, 71, 7S, 97, 95, JJP, J22. J M, 149
LIST OF FIGURES i
FIGURE IJ1g pyg
{
2.1-1 Core Protection Safety Limit THI-l 2 4a 2.1 2 Core Protection Safety Limits TM1 1 2 4b j
2.1 3 Core Protection Safety Bases THI-l 2 4c l
2.3 1 THI l Protection System Maximum Allowable Setpoints 2 11 2.3 2 Protection System Maximum Allowable Setpoints for 2 Axial Power Imbalance TMI l 3.1 1 Reactor Coolant System Heatup/Cooldown Limitations 3 Sa (Applicablethru10EFPY) 3.1-2 Reactor Coolant inservice Leak and Hydrostatic Test 3-5b (Applicable thru 10 EFPY) 3.1 2a Dose equivalent 1 131 Primary Coolant Specific-Actual 3 9b Limit vs. Percent of RATED THERMAL POWER 3.1 3 Limiting Pressure vs Temperature Curve for 3-18b 100 STO cc/ Liter H,0 3.5 2A thru DELETED l
3.5 2L 3.5-2M LOCA Limited Maximum Allowable Linear Heat Rate 3 36b 3.5 1
.Incore Instrumentation Specification 3-39a Axial Imbalance Indication 3.5 2 Incore Instrumentation Specification 3 39b Radial Flux-Tilt Indication 3.5-3 Incore Instrumentation Specification-3-39c 3.11 1 Transfer Path to and from Cask Loading Pit 3-56b 4.17 1 Snubber Functional Test Sample Plan 2 4 67 5-1 Extended Plot Plan TMI 52-
' Site Topography 5_ Mile Radiur 53 Gaseous Effluent Release Points and Liquid Effluent Outfall locations vil Amendment Nos. JJ. J7, 1), JJ, 9), JS, J), 71, 19), 199, //S, J/#,
{
119,Jpt,159 g
l
4.
2400 2200 ACCEPTABLE m
{
OPERATION s
) 2000 C
t g
UNACCEPTABLE n
OPERAT!0N 1800 7
v 1600 580 600 620 640 660 Reactor Outlet Temperature, OF CORE PROTECT!0N SAFETY LIMIT TMI-1 Figure 2.1-1 2-4a Amendment No. W. 142
\\
~
Ther-tal Power Level. '.
t 120
(-43.8.112) 1 (37.8,112)
ACCEPTABLE 4 PUMP OPERATION 100
_-43.8.89.3) 2
( 37. 8.09. 3_)
(
( 58.5.30.4)
ACCEPTABLE 80
($3,0.30.4) 3 & 4 PUNP OPERATION
(-43.8.62.0) 3 (37.8.62.0) i
(-30.5.37.8)
ACCEPTABLE 60 (53.0.57.8)
I 2.3. & 4 PUMP i
OPERATION 40 1
I
(-58.4.30.4)
.(53,0,30,4)
- 20 t
i e
i e
i 1
i i
i i
e i
i i
1 70 50 30 10 0
10 20 30 40 50 60 70 80 Axial Power Imbalance. ".
Curve Reactor Coolant Flow (lb/hr) 6 1
139.8 x 10 6
2 104.5 x 10 6
3 68.8 x 10
.b
.4 i
CORE PROTECTION SAFETY LIMITS TMI-I Anendment'No. 11, 28 )#, 4,
M. Up, Wy, 142 2-4b l
2400
~
2200 1
gCT
~
g E
2 y
2000 E
a.
T:
3 E
S 1800 ff 1600 580 600 620 640 660 Reactor Outlet Teaperature. OF Reactor Coolant Flow Curve (1bs/hr)
Power Pumps Operating (Type of Limit) 1 139.8 x 106(1005)*
112%
FourPumps(DNBRLimit) 2 l'04. 5 x 100(74.7%)
89.4%
Three Pumos (Quality Limit)
~
3 68.8 x 100(49.2%)
62.0%
OnePumpinEachLoop(QualityLimit)
- 106.5% of Cycle 1 Design Flow CORE PROTECTION SAFETY BASES
~
2-4c Amendment flo. 59 179, 142_
2500 P = 2355 psig 2300 ACCEPTABLE OPEPATION 2100 t
u P = 1900 psig g
1900 U
UNACCEPTABLE 8
OPERATION t
a 1700 1500 i
i i
540 560 580 600 620 540 Reactor Outlet Temperature. OF 1
,5 TMI-1 PROTECTION SYSTEM MAXIMUM ALLOWABLE SETPolNTS l
Amendment No. 13. 17, 28. 39. 45.
Figure 2,3-1
- 78. I26. J M 142 2-11
l Thermal Power Level, %
--120
( 30.0.108)
(24.5.108)
ACCEPTABLE l
g = 1.900 j 4 pugp
.100 g
2 = -1.854 m
m l OPERATION l
+
- (-30.0,60.6 )
(24.5.80.6 )
\\
ACCEPTABLE 80
(-50.0,70.0) lPfRf7 (45.0,70.0)
N l(-30.0,53.1 )
6f24.5,53.1) l l ACCEPTABLE -
2,3, & 4-(45.0.42.6 )--
(-50.0.42.6 )
l PUMP
. 40 l
I OPERATION l
1 l-i g
I l
o o
(-50.0,15.1 ) f pl 20 9 I
(45.0.15.1) l l
e el el e
i i
i i
i i
e ii i e
i i
i i
- 70 50 30 10 0 10 20 30 40 50 60 70 80 Axial Power Imbalance, %
6 PROTECTION SYSTEM MAXIMUM ALLOWABLE SETPOINTS FOR AXIAL POWER IMBALANCE l-TMI 1 Amendment No. 17,29,39.-AB,45, Fiqure-2.3-2 l
50.I20,)?p,142 2-12 p
l
3 8
8 E
1 2_
=
k c
~
I
.w i5mE[g
$,f et P
s ssaarissetava 5
r5~Ij i
i !
.E tj
~~
.las-7 I.
c
. I Ed-3 II 0y es Ei
- {'
g naassaassesss 2
a jtgl_-"' I s g g j =
g g
~
g 85 g
= sg lsEl
.IE 5
v a\\m k g:
b I ~!jsj i.:~3s N =
,E
~k*gs 3
i. l = =, s 1.g 4 2 J
1 t
f 8_!,a-i's?i-
- 233, g
v t
-I s3
=la r-as.Ia-8-g_5 s
-r n
n h
a svY.Ka*
h 2
$$3E 05 5C a
a hj
- i
=
gE25..w I s
=
x Igw we8
")*gg (p.
y i 2E35
$5 m 8
- ug Egy E3ww!:_
= ~
dif_r rtcus:
T32
(
5 gg33 Ea 4 4 s..
i i
g i
R-
_I
_a _!
E_ _a R
3 8
a
~
~
(dti 641 toH $33 19) 61sd 'aanssaad wists wetoog setssay pastolpui Amendment flo. /I,134 (4/28/77)
. 5a
9 8
ee t
~
o 5
R 2
8 r
e-e EE 5
I!M V.e$
R'I
$e 5: =l~
35 i 5355555 s ! ".5l
?
$3
, s j*Il
!.F t
}$
5~.
~
- i t.
25,
.-l j-na s:~:p^a
- t.5=15 ig, i
E 1
==
a v
5.Iat-gsE!
c.:
IE 8
8 ej2j
~
t
_. e 8 Ej
[g!s'!cif"-i j
i ---r-t.
Eifl.8g3jjipssa sa
=
=s-asv!--_8-,,,, I=gs-ssi-g _i
~
u g
i a
-~=
ss 3 m
- mss
=-
t a
a
-a ge Esy Is -
.h
- EI 35l5'[E E
In
- g se.g n-i aa,a g
I:-v al sis ltslc.l8aislii;
= - my vv i !!i-s ss;5 32 ae es i
i i
8 g
g g
[
]
[
]
3
$ $*g e
~
~
~
(del 641 tog $3s st) 61sd 'sanssaJd wets /s 1u:1003 actstay patt31put Amendment No.
134
/28/77) 3-5b l
~..
,p..
i.u..
'.u i
g
=
p g......
3 v 250 s --,
i 3
_ _.y.
. _ 1_
p.a._.
i n
e
- +
.,, '. p.a w 5
z E
. k' -
l' d.l'
_\\,
H,,'
-F t.'.'UN ACC E PT A B LF U
-..b,.i
. i 1'
OPERATION
=,
6 u
_ a..
.... m.2.,
....:.._..,..-H-AHi- +t.H+ :+H-- - + -
.- w c
r.
s i,.
=
,,,i,
. i....
.=,,,
i,n,
~
~~
~
5.
ala 4.i i.
'. t '
. s 8
,'.i, s, ', '
\\ '
u
. \\: -...2"--
e 6.
i e Aq Ei ii1 i.
. # 4.
i, i
e i i
g i
... 6
. 6..
rg e
g
...i L..
~ ~ ~ ~
g I"*
\\(' ' r <
4 ' '
a.
.. i, E
' * * * ~ "
?"""
4 "I. i t
Fo i.i i \\"-"
. e i
- Fi g
. 1.
i ig.i
.ii.6
=
i.
.. 1I
,,,,,,,4 4
z ACCEPTA81.E t
OPERATION w
s<
50
.___2___
g
=
_ q.,..
a l-w
....L
.. y.1.. '. ;. _
_. 4.... y.._.....
p x
._.r-------
0 20 30 40 50 60 70 80 90 100 PERCENT OF RATED THERMAL POWIR FIGURE 3.1-2a Dose equivalent 1-131 Primary Coolant Specific Activity 1.imit Versus Percent of RATED THERMAL POWER (with the Primary Coolant Specific Activity >1.0 pCi/ gram Dose Equivalent 1-131).
t i
l l
Amendment No.108 3-9b e
-~-
v
-+rs x
m.
v,mo n
rn,-p~-
y v v n-e-m--
,c e
m-----
vv.--n-
_. ~
',)
2000 1800 t
I
~ 16 00 I. 1400 i
3 1200 2
5 E 1000 30 IDO g
r 2
3 600 3
2
?
4 00 -
7 200 7
0 i
0 100 200 300 400 500 600
-700 Innicaten Reactor Coolant System Tamparature *f LIMITlHG PRES $URE V5 TEMPERATURE CURVE FOR 100 STD CC/ LITER H O 2
THREE MILE ISLAND NUCLEAR STATt0N UNIT I 3-13b FIGURE 3,1-3
3.1.11 REACTOR INTERNALS VENT VALVES anglicability Applies to Reactor Internals Vent Valves Ob.iective To verify that no reactor internals vent valve is stuck in the open position and that each valve continues to exhibit freedom of movement.
Soecifications 3.1.11.1 The structural integrity and operability of the reactor internals vent valves shall be maintained at a level consistent with the acceptaitte criteria in Specification 4.16.
4 3-18c Amendment No. 42 (8-16-78)
~
3.1.12; Pressurizer Power Operated Relief Valve (PORV) and Block Valve Apolicability Applies to the settings, and conditions for isolation of the PORV.
Ob.iective To prevent the possibility of inadvertently overpressurizing or depressurizing the-Reactor Coolant-System.
Soecification 3.1.12.1 The_ PORY shall not-be taken out of service, nor shall it be isolated i
from the system (except that the PORY may be isolated to limit leakage to_ wi_ thin the limits af Specification 3.1.6) unless one of the following is in effect:
a.
High Pressure injection Pump breakers are racked out or MU-V16A/B/C/D and MU-V217 are closed, b.
. Head of the Reactor Vessel is removed.
c.
Tavg is above 332*F.
3.1.12.2 The PORY settings shall be as follows, within the tolerances of i 25 psi and i 12'F:
Above 275'F - 2450 psig Below 275'F - 485 psig 3.1.12.3 If the reactor vessel head is installed and Tavg is $332*F, High Pressure Injection Pump br akers shall not be racked in unless:
a.
MU-V16A/B/C/D and MU-V217 are closed, and b.
Pressurizer level is 5 220 inches.
If pressurizer level is > 220 inches, restore level to 1 220 inches within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
3.1.12.4 PORY and Block Valve The PORY and the associated block valve shall be ~0PERABLE during HOT STANDBY, STARTUP, AND-POWEP OPERATION:
a.
With the PORV inoperable, within I hour-either restore the PORV to OPERABLE status or close the associated block valve and remove power from the block valve; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
b.
With the PORV block valve inoperable, within I hour either restore the PORV block valve to OPERABLE status or close the PORV (verify closed) and remove power from the PORV; otherwise, be in at least HOT-STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the fallowing-30 hours.
3-18d Amendment No. 56, 78, 149
i c.
With either the PORV or block valve inoperable, restore the inoperable valve to operable status prior to startup from the next cold shutdown unless the cold shutdown occurs within 90 Effective Full Power Days (EFPD) of the end of the fuel cycle.
If a cold shutdown occurs within this 90-day period, restore the inoperable valve to operable status prior to the startup for the next fuel cycle.
Eiui.11 If the PORV is removed from service, sufficient measures are incorporated to prevent severe overpressurization by either eliminating the high pressure sources or flowpaths or assuring that the RCS is open to atmosphere.
In order to prevent exceeding leakage rates specified in T.S. 3.1.6., the PORY may be isolated.
The PORV setpoints are specified with tolerances assumed in the bases for Technical Specification 3.1.2.
With RCS temperatures less than 332'F and the makeup pumps running, the high pressure injection valves are closed and pressurizer level is maintained less than 220 inches to prevent severe overpressurization in the event of any single failure.
Both the PORV and the PORY block valve should be operable during the HOT STANDBY, STARTUP, AND POWER OPERATION. If either the PORY or the PORY block valve are inoperable, the PORV discharge line should be isolated to prevent potential uncontrolled RCS depressurization.
For protection from severe overpressurization during HPI testing, refer to Section 4.5.2.1.c.
3-18e Amendment No. 78, 149
. ~.
-20 1
.a
=
18 J
[. / * &&
16 t
3
'/
}
/
~l 14
=
I
=
I 12 0-1000 mwd /stu After 1000 mwd /mtu-l l
10 I
1 i
i t
0 2
4 6
8 10 12 Antal Location From Bottom of Core, ft.
LOCA LIMITED MAXIMUM ALLOWABLE LlHEAR HEAT RATE TMI-1 Dendment No. lM 152 3-36b
9 p
9 9
w M
a m a' 7
N nvvtin
[o
\\
f s
/
,~
r
o
\\
\\
/ /
TC: :f ' A.
CCFE p
eu g
j N-O
- y e/
N
~
4
~~~
~~ %
4 h
W
/
/
s gg o
u m.
ist U
/
x
=
N
/
~
d 8 M
=
~
/-
N
~
[Q
/
O
\\
/
s M
+
\\
\\
/
x
\\
/
s;tICy art:t
.-m:.
'~
H 1' ;
~---%
[
h N
1 INCORE INSTRUMENTATION SPECIFICATION l
AXIAL IMB AL ANCE INDICATION THREE MILE ISLAND NUCLEAR STATION LWIT 1 3-39a FIGURE 3.5-1
L 4
b I
RADIat
,=t i
l i
i 1~ N
,+' +-l-+ s+s N
picm smna
/
IN THt $ PL ANE N!.i i
x x
i EME N
N h
/
\\
i
(
x
/
//
=
~
E
\\
N ss
,,c,,; 5y,,,
s
,s.m.
s,
(+ \\
i
/
N l
-+
/
\\
/
N INCORE INSTRUMENTATION SPECIFICATION RADIAL FLUX INDICATION THREE MILE ISLAND NUCLEAR STATION UNIT 1 3-39b FIGURE 3.5-2
. - ~
l.
i
/
s sus
/ 'I
-i
/
TT O
\\
g E
D s
/
m IM f[
z N
e
~
ii h
4 >
a,,
M
- 6
/j 5
kq x' -
/
s 5
EM J+, -
Ta+sN.Q\\
M o
s ss
/
/
tl.
gN A06
- 410
~
/
N
/
/
lHCORE INSTRUMENTATION SPECIFICATION THREE MILE ISLAND NUCLEAR STAT'ON UNIT I FIGURE 3.5-3 e-w
n 3.6.8.1 If inoperability is due to reasons other than excessive combined leakage, close the associated valve end within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> verify that the associated valve is OPERABLE. Maint in the associated valve closed until the faulty valve can be declared OPERABLE.
If neither purge valve in the penetration can be declared OPERABLE within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, be in HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
3.6.8.2 If inoperability is due to excessive combined leakage (See Specification 4.4.1.7.a), within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> restore the leaking valve to OPERABILITY or l
perform either a or b below, a.
Manually close both associated reactor building isolation valves and meet the leakage criteria of Specification 4.4.1.7.a and l
perform either (1) or (2) below.
(1)
Restore the leaking valve to OPERABILITY within the following 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
(2) Maintain both valves closed by administrative controls, verify both valves are closed at least once per 31 days and perform the interspace pressurization tes; of Specification 4.4.1.7.a j
every 3 months.
In order to accomplish repairs, one containment purge valve may be opened for up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> following successful completion of an interspace pressurization test, b.
Be in H1T SHUTDOWN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
346.9 Except as specified in 3.6.11 below, the Reactor Building purge isolation valves (AH-Vl-A&D) shall be linited to less than 31* and (AH-Vl-B&C) shall be limited to less than 33' open, by positive means, while ourging is conducted.
3.6.10 During STARTUP, HOT STANDBY, and POWER OPERATICN:
a.
Containment purging shall not be performed for temperature or humidity control.
b.
Containment purging is permitted to reduce airborne activity in order to facilitate containment entry for the following reasons:
l (1) Non-routine safety-related corrective maintenance.
I (2) Non-routine safety-related surveillance.
i (3)
Performance of Technical Specification required surveillance.
(4)
Radiation Surveys.
l (5)
Engineering support of safety-related modifications for l
pre-outage planning.
(6)
Purging prior to shutdown to prevent delaying of outage commencement (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to shutdown).
3-41a Amendment Nos. E7, JpE
d 5
c.
Containment purging is permitted for Reactor Building pressure
- control, d.
To the extent practicable, the above containment entries shall be scheduled to coincide, in order to minimize instances of purging.
3.6.11 When the reactor is in COLD SHUTDOWN or REFUELING SHUTDOWN, continuous purging is permitted with the Reactor Building purge isolation valves opened fully.
Bases The Reactor Coolant System conditions of COLD SHUTDOWN assure that no steam will l
be formed an hence no pressure will build up in the containment if the Reactor Coolant System ruptures.
The selected shutdown conditions are based in the type of activities that are being carried out and will preclude criticality in any oCCurrenC9.
A condition requiring integrity of containment exists whenever the Reactor Coolant System is open to the atmosphere and there is insufficient soluble poison in the reactor coolant to maintain the core one percent subcritical in the event all control rods are withdrawn.
The Reactor Building is designed for an internal pressure of 55 psig, and an external pressure of 2.5 psig greater than the internal pressure.
Due to industry reports of elastomer degradation in containment purge valve seats, unique action requirements are now designated to help preclude common mode l
failure of both valves in series.
An increased frequency of leak rate testing is also incorporated to help assure timely discovery and resolution of any seat degradation.
An analysis of the impact of purging on ECCS performance and an evaluation of the radiological consequences of a design basis accident while purging hne been completed and accepted by the NRC staff. Analysis has demonstrated that a purge isolation valve is capable of closing against the dynamic forces associated with a LOCA when the valve is limited to a nominal 30* open oosition.
Allowing purge operations during STARTUP, HOT STANDBY, and POWER OPERATION (TS 3.6.10) is more beneficial than requiring a cooldown to COLD SHUTDOWN from the standpoint of (a) avoiding unnecessary thermal stress cycles on the reactor coolant system and its components and (b) reducing the potential for causing unnecessary challenges to the reactor trip and safeguards systems.
l The recombiner unit is capable of controlling the expected hydrogen generation associated with 1) zirconium water reactions 2) radiolytic decomposition of water and 3) corrosion of metals within containment.
The recombiner is designed in accordance with the recommendations of Regulatory Guide 1.7, " Control of Combustible Gas Concentrations in Containment following a LOCA," Harch 1971, the acceptance criteria of the Standard Review Olan (SRP) 6.2.5, and NUREG 0578, July 1979.
In addition to the installed hy, 99en recombiner, a second recombiner including all piping, electrical, and structural provisions is available on site, t
3-41b Amendment Nos. E/, JEE
4.4 Reactor Buildina 4.4.1 Containment Leakage Tests Acolicability Applies to containment leakage.
Ob.iective To verify that leakage from the Reactor Building is maintained within allowable limits.
Soecification 4.4.1.1 Intearated leakaae Rate Test (ILRT) 4.4.1.1.1 Dasign Pressure Leakage Rate The design integrated leakage rate, (Lq)ht percent of the building atmosphere at I
, from the Reactor Building at the 55 psig design pressure, P, is 0.1 weig d
that pressure 9er 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
4.4.1.1.2 Allowable Integrated Leakage Rate The maximum allowable integrated leakage rate, (La), from the Reactor Building at the calculated peak Reactor Building internal pressure of 50.6 psig (P }
a associated with the design basis accident, shall not exceed 0.1 weight percent of l
the building atmosphere at that pressure per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
4.4.1.1.3 Conduct of Tests j
a.
During the period between the initiation of the containment inspection and the performance of a periodic ILRT, no repairs or adjustments shall
[
be made unless the inspection reveals structural deterioration which could affect the containment structural integrity or leak-tightness.
Such structural deterioration shall be corrected before performance of the test and a description of the deterioration and the corrective action taken shall be reported as part of the test report submitted in accordance with 10 CFR 50 Appendix J.V.B.
l b.
The test duration shall be at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> unless experience from at l
least two prior tests provides evidence of the adequacy of a shorter test duration, c.
Missile shielded lines outside the secondary shield will not be vented.
l d.
All containment components normally pressurized by the penetration pressurization system shall be at atmospheric pressure during the ILRT.
4-29 Amendment Nos. A3 1
i
l 4.4.1.1.4 Frequency of test The containment leakage rates shall be demonstrated at the following test schedule:
a.
Three Type A tests shall be conducted in accordance with 10 CFR 50 Appendix J, Section Ill D.I.(a), except as noted in 4.4.1.1.4.b and as stated in the Bases, b.
Where an exemption from the frequency specified by 10 CFR 50 Appendix J has been granted by the NRC, the schedule specified by the exemption shall apply.
4.4.1.1.5 Acceptance Criteria for Periodic ILRT l
for initial and periodic ILRT at P ai lam shall be less than 0.75 La-l 4.4.1.1.6 Corrective Action and Retest j
If the test data from a completed leakage rate test does not meet the acceptance criteria, the ILRT need not be repeated provided local leakage rate measurements I
are made at pressure P befoie and after repair to demonstrate that the leakage a
rate reduction achieved by the repairs reduces the overall measured integrated leakage rate to an acceptable value.
4.4.1.1.7 Report of Test Results l
Reporting of test results shall be in accordance with 10 CFR 50 Appendix J.V.B.
requirements.
4.4.1.2 Local Leakaae Rate Tests (LLRT) l 4.4.1.2.1 Scope of Testing LLRT of penetrations and valves identified in the FSAR shall be performed in accordance with 10 CFR 50 Appendix J except as provided in 4.4.1,2.5.c.
4.4.1.2.2 Conduct of Tests LLRT shall be performed pneumatically at a pressure of not less than P, with the a
exception that the access hatch door seal test shall normally be performed at 10 psig and the test every six months specified in 4.4.1.2.5.a shall be performed at a pressure not less than P -
a 4.4.1.2.3 Acceptance Criteria The combined leakage froa all penetrations and valves subject to LLRT shall not exceed 0.60 La (the maximum allowable leakage rate at Pa)-
4.4.1.2.4 Corrective Action and Retest a.
If at any time it is determined that the criterion of 4.4.1.2.3 above is exceeded, repairs shall be initiated immediately, b.
If conformance to the criterion of 4.4.1.2.3 is not demonstrated within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> following detection of excessive local leakage, the reactor shall be shutdown and depressurized until repairs are effected and the local leakage meets the acceptance criterion as demonstrated by retest.
Amendment No. 27 4-30
I 4.4.1.2.5 Test frequency LLRT shall be performed at a frequency as required by 10 CFR 50 Appendix J, except that:
a.
The entire personnel and emergency airlocks shall be tested once every l
six months. When the airlocks are opened during the interim between six month tests, the airlock door resilient seals shall be tested within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of the first of each of a series of openings.
This requirement exists whenever containment integrity is required, b.
An interspace pressurization test (See T.S. 4.4.1,7.a) shall be l
performed for reactor building purge isolation valves every 3 months.
This requirement is not in effect during cold shutdown.
c.
Where an exemption from the frequency specified by 10 CFR 50 Appendix J l
has been granted by the NRC, the frequency specified by the exemption shall apply.
4.4.1.3 Isolation Valve Functional Tests Every three months, remotely operated reactor building isolation valves shall be stroked to the position required to fulfill their safety function unless such operation is not practical during plant operation.
The valves not stroked every three months shall be stroked during each refueling period.
4.4.1.4 Annual Inspection A visual examination of the accessible interior and exterior surfaces of the containment structure and its components shall be performed annually and prior to any ILRT to uncover any evidence of deterioration which may affect either the l
containment's structural integrity or leak-tightness.
The discovery of any significant deterioration shall be accompanied by corrective actions in accordance with acceptable procedures, nondestructive tests, and inspections, and local testing where practical, prior to the conduct of any ILRT.
Such repairs I
shall be reported as part of the test results.
l 4,4.1.5 Reactor Building Modifications
(
l Any major modification or replacement of components affecting the reactor building integrity shall be followed by either an ILRT or an LLRT, as appropriate, and shall meet the acceptance criteria of 4.4.1.1.5 and 4.4.1.2.3, respectively.
4.4.1.6 Operability of Access Hatch Interlocks a.
At least once per six months the operability of the personnel and l
emergency hatch door interlocks and the associated control room-l annunciator circuits shall be determined, if the interlock permits both doors to be open at the same time or does not provide accurate status l
indication in the control room, the interlock shall be declared inoperable.
4-31 Amendment Nos. D
b.
During' periods when containment integrity is required and an interlock I
is inoperable, each entry and exit via that airlock shall be locally supervised by a member of the unit operating, maintenance, or technical staffs, to assure that only one door is open at any time and that both doors are properly closed following use.
A record of supervision and verification of closure shall be maintained during periods of interlock inoperability in an appropriate station log.
c.
If'an interlock is inoperable for more than 14 days following l
determination of inoperability, use of the airlock, except for emergency purposes, shall be suspended until the interlock is returned to operable status.
4.4.1.7 Operability of Purge Valves a.
A periodic pressurization of the purge valve interspace to 50.6 psig per l
Specification 4.4.1.2.5.b shall be performed to help assure timely detection and resolution of valve and/or actuator degradation.
The acceptance criteria is that total local leakage, when updated for the new purge valve leakage, shall be less than 0.60 L.
See Specification l
a 3.6.8 for further action.
b.
The rubber seats on purge valves shall be visually examined and I
durometer tested each refueling interval to detect degradation (e.g.
cracking, brittleness, etc.) and to assure timely cleaning, lubrication, and seat replacement.
Bases (1) 9 e performance of periodic ILRT and LLRT during the plant life provides a current assessment of potential leakage from the containment in case of an accident that would pressurize the interior of the containment.
In order to provide a realistic appraisal of the integrity of the containment under accident conditions, "as found" local leakage results must be documented for correction of the ILRT results.
Containment isolation valves are to be closed in the normal manner prior to LLRT or ILRT. Containment Isolation Valves are addressed in the UFSAR (Reference'2).
The minimum of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> was specified for the ILRT to help stabilize conditions and thus improve accuracy and to better evaluate data scatter.
The frequency of the periodic ILRT is keyed to the refueling schedule for the reactor, because these tests can best be performed during refueling shutdowns.
Surveillance tests for measuring leakage rates are consistent with the requirements of 10 CFR 50, Appendix J with the following exemption. The third test of each Type A testing set need not be conducted when the plant is shut down for the 10-year plant inservice inspections (Reference 3).
The operational 1
readiness of the containment is proven by the ILRT, and in accordance with license requirements, when completed pursuant to the frequency stated in i
Technical Specification 4.4.1.1.4.
l 4-32 Amendment Nos. O, U, J M, JEJ l
9 f
The specified frequency-of periodic ILRT is based on three major considerations.
l First is the low probability of leaks in the liner, because of conformance of the complete containment to a 0.1 percent leakage rate at 55 psig during pre-operational testing and the absence of any significant stresses in the liner during reactor operation.
Second is the more frequent testing, at P, of those l
a portions of the containment envelope that are most likely to develop leaks during reactor operation and the low value of leakage that is specified as acceptable from penetrations and isolation valves (0.60 La).
Third is the tendon stress j
surveillance integrity program which provides assurance that an important part of the structural integrity of the containment is maintained.
More frequent testing of various penetrations is specified as these locations are more susceptible to leakage than the reactor building liner due to the mechanical closure involved.
The basis for specifying a total leakage rate of 0.60 La from j
those penetrations and isolation valves is that more than one-half of the allowable integrated leakage rate will be from these sources.
Valve operability tests are specified to assure proper closure or opening of the reactor building isolation valves to provide for isolation or functioning of Engineered Safety Featu m systems.
Valves will be stroked to the position required to fulfill their safety function unless it is established that such l
testing is not practical during operetion.
Valves that cannot be full-stroke tested will be part-stroke tested during operation and full-stroke tested during each normal refueling shutdown.
Periodic surveillance of the airlock interlock systems (Reference 4) is specified l
to assure continued operability and preclude instances where one or both doors are inadvertently left open. When an airlock is inoperable and containment integrity is required, local supervision of airlock operation is specified.
Purge valve interspace pressurization test operability requirements, inspections, and durometer testing provide a high degree of assurance of purge valve performance as containment isolation barriers.
Reference (1) UFSAR, Chapter 5.7.4
" Post Operational Leakage Rate Tests" (2) UFSAR, Tables 5.7-1 and 5.7-3 (3) NRC Letter dated February 25, 1991 (C311-91-3033)
(4) UFSAR, Table 5.7-2 l
i
(
l l
l l
l 4-33 (Pages 4-34, 4-34a, and 4-34b deleted) l
4'.5 EMERGENCY LOADING SE0VENCE AND POWE.R TRANSFER. EMERG MCY CORE COOLING SYSTEM & REACTOR BUILDING COOLING SYSTEM PERIODIC TLSTING 4.5.1 Emergency Loading Sequence Anolicability:
Applies'to periodic testing requirements for safety actuation systems.
Ob.iective:
To verify that the emergency loading sequence and automatic power transfer is operable.
Specifications:
4.5.1.1 Seouence and Power Transfer Test a.
During each refueling interval, a test shall be conducted to demonstrate that the emergency loading sequence and power transfer is operable, b.
The test will be considered satisfactory if the following pumps and fans have been successfully started and the following valves have completed their travel on preferred power and transferred to the emergency power as evidenced by the control board component operating lights, and a-second means of verification, such as: the station computer, verification of pressure / flow, or control board indicating lights initiated by separate limit switch contacts.
-M. U. Pcmp
-D. H. Pump and D. H. Injection Valves and D. H. Supply Valves
-R. B. Cooling Pump
-R. B. Ventilators
-D. H. Closed Cycle Cooling Pump
-N. S. Closed Cycle Cooling Pump
-D. H. River Cooling Pump
-N. S. River Cooling Pump
-D. H. and N. S. Pump Area Cooling Fan
-Screen House Area Cooling Fan
-Spray Pump. (Initiated in coincidence with a 2 out of 3 R. B.
30 psig Pressure Test Signal..)
-Motor Driven Emergency Feedwater Pump c.
Following successful transfer to the emergency diesel, the diesel generator breaker will be opened to simulate trip of the generator then reclosed to verify block load on the reclosure.
4.5.1.2 Egouence Test a.
At intervals not to exceed 3 months, a test shall be conducted to demonstrate that the emergency loading sequence is operable, this test shall be performed on either preferred power or emergency power, b.
The test will be considered satisfactory if the pumps and fans listed in 4.5.1.lb have been successfully started and the valves listed in 4.5.1.lb have completed their travel as evidenced by the control board component operating lights, and a second means of verification, such as:
the station computer, verification of pressure / flow, or control board l
indicating lights initiated by separate limit switch contacts.
4-39 Amendment No. 70, 7N 149
~
I
. z
)
r F
~.
c.
When the Decay Heat System is required to be operable, the l
correct position of DH-V-19A/B'shall be verified _ by observation within four-hours of each valve stroking operation or valve maintenance, which affects the_ position indicator.
l_
. 4.5.2.3 Core Floodina a.
Du' ring each refueling period, a system test shall be-
. conducted to demonstrate proper operation of the system.
During depressurization of the Reactor Coolant System, verification shall be made that the check'and isolation valves in the core cooling flooding tank discharge lines operate properly, b.
The test will be considered satisfactory if control board indication of core flooding tank level verifies that all valves have' opened.
4.5.2.4 Comoonent Tests
-t a.
At intervals not to exceed 3 months, the components required for emergency' core cooling will be tested.
b.
The tes _will be considered satisfactory if the pumps and fans have'been success.ully started and the valves have completed their travel as evidenced-'by the control l board component operating lights, and a second means of verification, such as:
tie station computer, verification of.
pressure / flow, or control board indicating lights initiated by separate limit switch contacts.
Bases The emergency core cooling systems (Reference 1) are the principal reactor safety features in the event of a loss of coolant accident. The removal of heat from the core provided by these-systems is designed to limit core damage.
~ The. low pressure._ injection pumps are tested singularly _for operability-by opening
-the borated water.-storage tank. outlet valves.and the bypass valves in the borated
- water storage tank fill line. This allows water-to be pumped from the borated
- water--storage tank-through each of the-injection lines and back to the-tank.
l The minimum acceptable HPI/LPI flow assures proper flow and flow split between injection: legs.
With the reactor shutdown, the valves in each core flooding line are checked for
[
- operability-by' reducing the reactor coolant system pressure until the indicated level in the core flood tanks verify that the-check and is'olation valves have opened.
Reference (1) UFSAR, Section 6.1
" Emergency Core Cooling System" p
l j
4-42 l
Amendment No. fq7, f#, WJ,157
l 4.5.3 REACTOR BUILDING COOLING AND ISOLATION SYSTEM Acolicability Applies to testing of the reactor building cooling and isolation systems.
Ob.iective To verify that the reactor building cooling systems are operable.
Soecification 4.5.3.1 System Tests a.
Reactor Buildina Snray System 1.
At each refueling interval and simultaneously with the test of the emergency loading sequence, a Reactor Building 30 psi high presstre test signal will start the spray pump.
Except for the spray pump suction valves, all engineered safeguards spray valves will be closed.
Water will be circulated from the borated water storage tank through the reactor building spray pumps and returned through the test line to the borated water storage tank.
The operation of the spray valves will be verified during the component test of the Reactor Building Cooling and Isolation System.
The test will be considered satisfactory if the spray pumps have been successfu'ly started as evidenced by the control board component operating lights, and either the station computer or pressure / flow indication.
2.
Compressed air will be introduced into the spray headers to verify the availability of the headers and spray nozzles at least every five years, I
b.
Re ctor Buildina Coolina and Isolation Systems 1.
During each refueling period, a system test shall be conducted to
. demonstrate proper operation of the system. A test signal will actuate the Reactor Building Emergency Cooling System valves to demonstrate operability of the coolers.
2.
The test will be considered satisfactory if the valves have l
completed their expected travel as evidenced by the control board component operating lights and a second means of verification, such as:
the station computer, local verification, verification of pressure / flow, or control board component operating lights initiated by separate limit switch contacts.
l l
l l
4-43
=
l e
4.5.3.2--Component Tetti a.
At intervals not to exceed three months, the' components required-for Reactor Building Cooling and Isolation will be' tested.
l b.
The test will be considered satisfactory if the valves have completed their expected travel as evidenced by the control board component operating lights and a second means.of verification, such i
as:
the station computer, local verification, verification of~
t pressure / flow, or control board ' component operating lights initiated by separate limit switch contacts.-
Bases
[
The Reactor: Building Cooling and Isolation Systems and Reactor Building Spray System are designed to remove the heat in the con' inment atmosphere to prevent the building pressure from exceeding the design pr.asure (References 1 and 2).
= The delivery ca; ability of-one Reactor Building S)
Pump at a time can_be i
tested by openi1g the valve in the line from the x ted water storage tank, l
opening the corresponding valve in the test line, n. starting the corresponding.
pump.-
With the p!mps shut down and the Borated Water Storage Tank outlet valve closed, the Reactor Building-spray injection valves can each L: opened and closed by operator action. With the Reactor Building spray inlet valves closed, low-
~
pressure air can be blown through the test connections of the Reactor Building spray nozzles to' demonstrate that the flow paths are open.
The equipment, piping, valves and instrumentation of the Reactor Building Cooling System are arranged so that they can be visually inspected. The cooling units and associated piping are located outside the secondary concrete shield.
Personnel can enter the Reactor' Building during power cperations to inspect and maintain this equipment.
The Reactor Building fans are normally operating periodically, constituting the
- test-that these fans are operable.
' Reference
- (1) UFSAR, Section 6.2
" Reactor Building Spray _ System" (2) UFSAR, Section 6.3
" Reactor Building Emergency Cooling System" 4-44 Amendment No. pp, H P, 157
-e
+-+ww
+,n-r m-w w-rww ww - w.wwr rww w, wry w w.ws-sv w~
,-t-i e-,-r-g-w, w
+ " * * * '
T"
- T""*
4 d.
The battery will be subjected to a load te'st at a frequency not to exceed refueling periods. The battery voltage as a function of time will be monitored to establish that the battery performs as expected during this load test.
4.6.3 Pressurizer Heaters a.
The following tests shall be conducted at least once each refueling:
(1)
Pressurizer heater groups 8 and 9 shall be transferred from the normal power bus to the emergency power bus and energized, Upon completion of this test, the heaters shall be returned to their normal power bus.
(2) Demonstrate that the pressurizer heaters breaker on the emergency bus cannot be closed until the safeguards signal is bypassed and can be closed following bypass.
(3) Verify that following input of the Engineered Safeguards Signal, l
the circuit breakers,-supplying power to the manually transferred loads for pressurizer heater groups 8 and 9. have been tripped.
Bases The tests specified are designed to demonstrate that one diesel generator will provide power for operation of safeguards equipment.
They also assure that the emergency generator control system and the control systems for the safeguards equipment will function automatically-in the event of a loss of normal a-c l
station service power or upon the receipt of an engineered safeguards Actuation Signal. The automatic tripping of manually transferred loads, on an Engineered Safaguards Actuation Signal, protects the diesel gener: tors from a potential overload condition. The testing frequency specified.s intended to identify end I
permit correction of any mechanical or electrical deficiency before it can result in a system failure.
The fuel oil supply, starting circuits, and controls are continuously monitored and any faults are alarmed and indicated. An abnormal condition in these systems would be signaled without having to place the diesel generators on test.
Precipitous failure of the station battery is extremely unlikel'.
The surveillance specified is that which has been demonstrated over the years to provide an indication of a cell becoming unserviceable long before it fails.
The PORV has a remotely operated block valve to provide a positive shutoff capability should the relief. valve become inoperable.
The electrical power for both the relief valve and the block valve is supplied from an ESF power source to l
ensure the ability to seal this possible RCS leakage path.
The requirement that a minimum of 107 kw of pressurizer heaters and their associated controls be capable of being supplied electrict.1 power from an emergency bus provides assurance that these heaters can be energized during a loss of offsite power condition to maintain natural circulation.
4-47 Amendment No. 78, 157
- - _ - ~ -. _ -..
4.15' MAIN STEAM SYSTEM INSERVICE INSPECTION i
Aeolicability This technical specification applies to the inservice inspection of four welds-in
- the Main Steam System identified as MS 0001, MS-0002, MS 0003, and MS 0004L of the THI-l Inservice Inspection Program.
Ob.iective The objec'tive of the Inservice Inspection Program is'to provide assurance of the continuing l integrity of that portion of the Main Steam System in which a postulated-failure would produce pressures in excess of the compartment wall and/or slab capacities.
'Soecification
~
4.15.1.
The four weld joints identified above shall be 100 percent inspected in accordance with the ASME Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant components, defined in the TMI-1 Inservice Inspection Program.
Inspections are to be performed at a frequency of once every 3-1/2 years (or during the nearest refueling outage).
Prior to initial plant operation, a preoperational inspection of the identified weld joints will be performed and any data acquired will-be recorded to form a baseline on which to compare results of subsequent inspections.
t Bases Calculations (Reference 1) postulated that bl.aks in the main steam lines at the l
containment penetrations in small compartments No. 2 and No. 5 could produce pressures in excess of wall and/or slab capacities.
Inspections are conducted at an inspection frequency of 31/2 year intervals following initial plant startup.
These inspections have revealed that no degradation of'the welds has occurred during the inspection cycles up to and including the 9R outage inspection. Consequently, as further degradation is not expected to occur, justification to extend the inspection frequency to once every ten (10) years is being developed. The conclusions of the technical benefit review will be submitted to the NRC for evaluation in a Technical Specification change request.
Reference (1) UFSAR,-Appendix 14A, Section 7.2.1 4-58 Amendment No. 149 4
t
10 9
8 Y
/
u.
7
/
=y REJECT 7
==
6
/
q, x-/
mC gS f g
C#
l
/
q r
8 CONTli4UE g
-TESTING
/
N
/
I@ /
' /,
5
/
2
,+
L*g F
s ACCEPT E
1 r
5 5g 0
10 20 30 40 50 60 70 80 90 1::
5 N
CUMULATIVE NO. OF SUUBBERS OF TYPE TESTED FIGURE 4.17-1 SNUSEER FUf1CT10NAL TEST - S2iPLE PLAN 2 4-67 l
Amendment flo.106 w
,-x:
--m, - -
p
- i c.
- THIS PAGE INTENTIONALLY LEFT BLANK l
l l
l (Pages 4-69 through 4-76A deleted) 4-68 Amendment Nos. 32, 101, 106, 146
___