B13137, Requests Exemption from Certain Requirements of 10CFR50, App J,Sections III.C.1 & III.C.2 Re Type C Testing Requirements.Diagrams for Each Penetration Included in Attachment I
| ML20245E803 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 04/28/1989 |
| From: | Mroczka E CONNECTICUT YANKEE ATOMIC POWER CO., NORTHEAST UTILITIES |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| B13137, NUDOCS 8905020146 | |
| Download: ML20245E803 (58) | |
Text
{{#Wiki_filter:- . . _ _ _ . . _ - _ _ _ _ _ - _ _ _
. bl '
o NORTHEAST UTILITIES o.n.,.i Ome.. . s.io.n si,..i. e.,iin Conn.ci,cui l rYwb rs$t P.O. box 270 L [.um.[s, wc,u,[m.[.o.$,,,, HARTFORD. CONNECTICUT 06141-0270
' J com. (203) 665-5000 April 28, 1989 Docket No. 50-213 l B13137 ]
Re: 10CFR50, Appendix J ISAP Topic 1.03 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20S55 Gentlemen: Haddam Neck Plant 10CFR50. Anoendix J. Reouest for Exemotion Pursuant to 10CFR50.12, Connecticut Yankee Atomic Power Company (CYAPCO) hereby requests an exemption from certain requirements of 10CFR50, Appendix J. Specifically, CYAPC0 is seeking a permanent exemption from the requirements of Section III.C.1 and III.C.2 of Appendix J to 10CFR50. By letter dated March 12,1986,II) CYAPC0 requested a schedular exemption from the requirements of 10CFR50, Appendix f2for certain containment penetrations. In a letter dated September 29, 1987, the Staff granted temporary relief from the requirements of Appendix J, Section III.C. Transmitted herewith is CYAPC0's request for permanent exemptions relating to the Type C testing requirements of 10CFR50, Appendix J, Section III.C with respect to certain containment penetrations. The exemption requests are provided in Attachment I, together with the technical justification for each exemption. Diagrams for each penetration are also included in Attachment I and are provided for information only. The Commission's regulations, specifically 10CFR50.12(a), provide that exemp-tions may be granted from the requirements in 10CFR50 if "special circum-stances" are present and the exemptions are " authorized by law, will not present an undue risk to the public health and safety, and are consistent with the common defense and security." Under 10CFR50.12(a)(2)(ii), "special circumstances" warranting an exemption are present whenever, among other things, application of the regulation in particular circumstances "is not necessary to achieve the underlying purpose of the rule. . . . " (1) J. F. Opeka letter to C. I. Grimes, "Haddam Neck Pl ant, 10CFRb0, Appendix J Request for Exemption," dated March 12, 1986. (2) D. M. Crutchfield letter to E. J. Mroczka, " Exemption from Certain Requirements of 10CFR Part 50, Appendix J," dated September 29, 1987. 1
' \
8905020146 890428 PDR ADOCK0500g3 I P
4 ( l U.S. Nuclear Regulatory Commission B13137/Page 2 April 28, 1989 On the basis of the information provided herewith, CYAPC0 concludes that the exemptions for the subject penetrations are justified under the standards of 10CFR50.12. As demonstrated in Attachment I, the requested exemptions will not present an undue risk to the public health and safety since no significant increase in the risk associated with containment leakage will result from the granting of the exemptions. In fact, granting the exemptions will result in a reduction in occupational exposure. Many of the modifications that would be required to satisfy the specific requirements of Appendix J would result in signifi a nt worker exposure. Furthermore, "special circumstances" are present in that application of the regulation is not necessary to achieve the underlying purpose of Appendix J. The purpose of the leak rate testing requirements in Appendix J is to provide reasonable assurance that containment leakage during a design basis event will not exceed the leakage limits specified in binding licensing documents or assumed in the pertinent accident analyses. The Type C testing requirements of Section III.C of Appendix J are established to provide adequate assurance that the leakage rate from containment penetrations is within acceptable limits. As shown in Attachment I, the requested exemptions from Type C testing requirements, for the selected penetrations, will satisfy the underly-ing purpose of the rule since containment leakage will be held within the i allowable limits. Given the system configurations and other factors presented in the exemption request, there is adequate assurance that the penetrations in
' question will remain sufficiently leak tight. This will be confirmed by the periodic integrated leak rate test (ILRT), which will ensure that the overall containment leakage rate is sufficiently limited.
Appendix J to 10CFR50 states that containment isolation valves (CIVs) must be leakage tested at design basis accident (DBA) pressure. In general, test pressure must be applied in the same direction as it would be during a DBA. This type of local leak rate test (LLRT) is called a direct pressure test. In certain circumstances, Appendix J permits test pressure to be applied to a valve in a direction opposite to that associated with a DBA. This type of testing is called reverse direction testing. By letter dated Augu.n 31, 1987,(3) CYAPC0 and Northeast Nuclear Energy Company (NNEC0) committed to provide the Staff with the results of CYAPC0's and NNEC0's review of CIV reverse direction testing at the Haddam Neck Plant and Millstone Unit Nos. 1, 2, and 3. This review was undertaken to satisfy a Staff request to assure that potential containment boundaries are not inappro-priately omitted from 10CFR50, Appendix J, LLRTs in which reverse direction testing is utilized. (3) E. J. Mroczka letter to W. T. Russell, "Haddam Neck Plant, Millstone Nuclear Power Station, Unit Nos. 1, 2, and 3, 10CFR50, Appendix J," dated August 31, 1987.
4 U.S. Nuclear Regulatory Commission B13137/Page 3 April 28, 1989 By letter dated May 2,1988,I4) CYAPCO and NNEC0 provided the Staff with the results of CYAPCO's and NNECO's assessment of the reverse direction testing programs at the Haddam Neck Plant and Millstone Unit No.1, respectively. The assessment for the Haddam Neck Plant concluded that omission of packings and seals on CIVs inside containment due to reverse direction testing is not an omission of testing of a containment boundary. For CIVs outside containment, the leak tightness of all potential containment isolation boundaries is I checked during their LLRTs. It should be noted that the description of Valve PU-V-242 for Penetration P-33 has been revised and corrected from our previous May 2, 1988, submittal. 1 In summary, CYAPC0 has concluded that exemptions for those areas discussed in Attachment I are warranted under the standards of 10CFR50.12. It should also be noted that through ISAP Topic 1.03, CYAPC0 has evaluated a number of plant modificationstoachigecompliancewithAppendixJrequirements. In a letter , dated March 2, 1989, CYAPC0 provided the Staff with an update to the ISAP l topic reviews. The modifications made during previous outages represent CYAPCO's prudent steps to improve the containment integrity of the Haddam Neck Plant and demonstrate CYAPC0's good f aith afforts to satisfy the requirements of Appendix J. Very truly yours, CONNECTICUT YANKEE ATOMIC POWER COMPANY 44 % W E. J. Mroczk'a // Seniof Vice Presf' dent cc: W. T. Russell, Region I Administrator A. B. Wang, NRC Project Manager, Haddam Neck Plant J. T. Shedlosky, Senior Resident Inspector, Haddam Neck Plant (4) E. J. Mroczka letter to W. T. Russell, "Haddam Neck Plant, Millstone Nuclear Power Station, Unit No.1, 10CFR50, Appendix J," dated May 2, 1988. (5) E. J. Mroczka letter to the U.S. Nuclear Regulatory Commission, "Haddam Neck Plant, Integrated Safety Assessment Program (ISAP)," dated March 2, 1989. j
l [ ?. l c. 1 ~ l Docket No. 50-213 B13137 Attachment I Haddam Neck Plant 1 Requests for Exemption From 10CFR50, Appendix J Requirements April 1989 _ _ _ _ _ _ _ . _ l
1 y
+ l U.S. Nuclear Regulatory Commission )
813137/ Attachment I/Page 1 April 28, 1989 Plant: Hadh Neck Plant
Title:
Exemption from 10CFR50, Appendix J, Section III.C.I. 1 Penetration: P-4 Pressurizer Relief Tank Vent ; Description CYAPC0 requests permanent exemption from the requirements of 10CFR50 Appen-dix J, Section III.C.1 that reverse direction Type C testing provide "equiva-lent or more conservative results" than direct pressure testing. CYAPC0 presently Type C leak tests valves WG-TV-1845, WG-V-984A, WG-A0V-558 in this pressurizer relief tank (PRT) vent penetration. WG-V-984A and WG-A0V-558 are direct pressure tested with nitrogen at 47 psig. WG-TV-1845 is reverse direction tested simultaneously with these valves. It can be shown that reverse direction testing of WG-TV-1845 is conservative with respect to disc / seat leakage. However, reverse direction testing does not ' include a flanged mounting joint, a stem packing, and a body-to-bonnet joint on the containment side of this valve. These features are potential containment isolation boundaries in this penetration. Justification Bases Reverse direction testing of WG-TV-1845 tends to push this valve's disc out of its seat. Since this valve is a diaphragm actuated, single seated globe valve, installed for air-to-open, flow-to-close use, this type of reverse direction test is conservative with res.ect to disc / seat leakage (i.e., containment accident pressure tends to close the valve more tightly than reverse direction test pressure). The body-to-bonnet joint, stem packing, and gasketed flange joint on the l containment side of WG-TV-1845 are Appendix J leak tested during periodic Type A integrated leak rate testing (ILRT). This is because the PRT is vented to containment during such testing. Under some beyond design basis accident l conditions, the PRT pressure could increase above the predicted maximum containment pressure. This is because the PRT rupture disk will not fail until the PRT pressure exceeds containment pressure by 75 psi. However, events that could be radiologically significant and result in pressurizing the PRT (e.g. stuck open PORV or code safety valve) would not result in failed fuel until long after the rupture disk had failed. For these events, Type A testing will bound the pressure seen by the valve when radiological conditions inside containment warrant low leakage. The containment side of WG-TV-1845 is normally exposed to PRT nitrogen blanket pressure (3 psig) during power operation. The potential containment bounda-ries missed by reverse direction local leak rate testing (LLRT) continuously demonstrate their leak tightness by retaining this pressure. Loss of PRT pressure triggers an alarm via transmitter #406.
._-_-_ - -- - _ - l
)
U.S. Nuclear Regulatory Commission B13137/ Attachment I/Page 2 April 28, 1989 The public risk implications of this Appendix J exemption have not been specifically addressed in the probabilistic risk assessment (PRA) performed I for penetration P-4 to date. However, the following statements and projec-tions can be made based on existing PRA. The risk of public exposure due to core melt accident leakage through this penetration has been assessed to be approximately .006 man-rem. If the combination of reverse direction testing i and continuous monitoring at 3 psig is judged to be a noncompliance with Appendix J, the assumptions used in the PRA would change, and the risk of 2 public exposure in a core melt accident would increase to approximately .06 l man-rem. If this were the case, modifications to eliminate reverse direction ; testing could attribute an approximate .054 man-rem public risk reduction l (i.e., .06 man-rem .006 man-rem = .054 man-rem). The relatively small public exposure consequences and public exposure savings associated with modifications in Haddam Neck's Appendix J program for Penetra-tion P-4 must be compared to the cost in dollars and the significant exposure associated with making plant changes necessary to support these modifications. It has been estimated that it will require 2.5 man-rem of craft personnel exposure to accomplish modifications which will permit direct pressure testing of WG TV-1845. These plant changes will cost approximately $90,000. Conclusion The leak tightness of all potential containment boundaries in penetration P-4 is demonstrated by the combination of the tests listed above. The continuous demonstration of leak tightness by certain boundaries retaining PRT pressure exceeds the requirements of Appendix J. Since continuous pressurization is at 3 psig rather than Pa = 40 psig as required for periodic ILRTs and LLRTs, it can not be stated that this demonstration is conservative with respect to Appendix J. However, CYAPC0 believes that the requested permanent exemption from Appendix J is justified on these bases.
4 HCP Attaclunent I page 2a
,h, *"
h HCP %
.-- J W WG-SOV-1845 S S WG-50V-558 7
A/S A/S C WEST OUTSIDE - - P.A.B. y PIPE TRENCH T
- PIPE TRENCH d d" BLONDOWN ROOM o
'"--* I d ~
o o o "P.A.B. PIPE TRENCH Q MVH-V-559
\
d o xx j 47 WG-V-136 C WG-V-113 l l 63 WG-TV-1845l j" d h_ j j' (L) l
> c] D' l> l 2 l huh
! (FC) j" FC) lWG-V-984 A [ (LC ) WG- A0V-558 l d Ih" j" SS-V-985A
._7, {"- RC-152R-71 dC., (
f, % ,
% - j"-RC-152R-71 SS-V-989A
[ PRESSURIZER RELIEF TANK
) P&ID - 16103-26007 SH. 3
( TK-8 j SYS. - REACTOR COOLANT (PRESSURIZER RELIEF TANK VENT) NOTE - FOR PENETRATION DETAILS SEE DWG. #16103-59018 SH. 23
& 16103-22027 SH.1&3
""~~~
I lXXXXXXl - IN ICATES VALVE I C*A*D* l TO DE TESTED l NOTt, MANUAL REVISIONS TO THIS AS-BUILT l DOCUMENT ARE PROH]B]TED. REFER TD g g GED 3.01 FOR REVISION DURING l CONSTRUCTION MAWAL REVIS10NS. ,,j g
@ "" NORTHEAST CONNECTICUT YANKEEUTILITIESATOMIC POWER CO. SERV
" CONTAINMENT ISOLATION VALVE TEST DIAGRAM
*- AS BUILT PER " PENETRATION
- 4 N/A 2 2,65 DCR-CY-S-494-89 k BY RJC C*D. RVK APP. WRK APP- WRK 8 9 011 1 '7 AS BUILT - PER "g g +g # fan 1-25-86 can 3-18-86 oan 3-20-86 Dan 3-20-86 es DSR-CY-5-336-86 o 4 PF Scd NONE 0E
- 5 ru a au emoc er om. w. w. ru 85-011 16103-20310 SH. 6 WRO
P% i U.S. Nuclear Regulatory Commission B13137/ Attachment 1/Page 3 April 28, 1989 Plant: Haddam Neck Plant
Title:
Exemption from 10CFR50, Appendix J Penetration: P-8 Reactor Coolant System Charging Description
. CYAPC0 requests a permanent exemption for Penetration P-8 from 10CFR50, Appendix J, Type C testing. P-8 is the charging system penetration. CYAPC0 presently does not perform a local leak rate test (LLRT) on check valve CH-CV-399.
' Justification Bases The piping system inside containment from the containment wall to the reactor t coolant system (RCS) is seismically supported. A check valve (CH-CV-399) exists on the inside of containment. >
During a period . integrated leak rate test (ILRT), this check valve is exposed to containment pressure through the undrained RCS in the accident direction and with the back side of the valve depressurized. The periodic ILRT provides . identification of any significant leakage through the isolation valve. System leakage to the environment is inspected as part of the technical specification limit of three liters / hour maximum leakage (10CFR100 offsite doselimit). A temporary water seal will exist due to this system being filled with water at the start of a core melt accident. Half of the thyroid dose due to con-tainment leakage occurs during the first 15 minutes following the onset of a postulated accident. Either the residual heat removal or charging system will be in operation for more than 30 days following a core melt accident. No containment atmospheric leakage through this penetration could occur while the charging system is in operation. The operation and physical configuration of the charging system provides a natural deterrent to containment leakage through this system. The PRA analysis for this penetration calculated the public risk to be approx-imately .06 man-rem in its present configuration with no LLRT being periodi-cally performed. The man-rem expenditure required to complete the modification to P-8 is I approximately 3.0 man-rem. The cost to modify this penetration is approxi- l mately $112,000. j
i
'. s l N U.S. Nuclear Regulatory Commission
-B13137/ Attachment I/Page 4
-April 28, 1989 1
Conclusion l The consequences to public risk of not performing a Type C leak test on P-8 l l is negligible. The small decrease in public risk exposure .does not justify the man-rem exposure and cost in dollars required to make the modifications. Additionally, by letter dated September 29, 1987 50) the Staff granted a schedular exemption from the requirements of 10CFR50, Appendix J for penetra-tion P-8. In granting the schedular exemption, the Staff determined that the Type A test in conjunction with the technical specification leakage limits for the system provide reasonable assurance that leakage through this penetration is limited. Moreover, the Commission stated that "the operation and physical
~
configuration of the charging system will provide a. good deterrent to any containment leakage past the isolation valves." CYAPC0 believes that permanent exemption from the Appendix J Type C testing requirements are justified based on the above. j 1 1 (6) D. M. Crutchfield letter to E. J. Mroczka, " Exemption from Certain Requirements of 10CFR Part 50, Appendix J," dated September 29,1987.\
Attachment I page 4a S CH-SOV-110A
)> "
p
., ,1 # cs 4:*.,3 ,, "
I" A/S ;/' 4' _ CH-V-286C CH-V-286B
, N C*C]
CH-FCV-110A N-(RM) g. g TO REGENERATIVE
. CH-V-286D HEAT EXCHANGERS h e s CH-V-286E 3"-CH-2502R- 8 m 2l '
a o-S CH-SOV-110 v ^ 3"-CH-2502R-234% lCH-CV-399Fj
/
L.L. ,
;y - P -- 8
#" " - " ~ ^
FROM CHARGING I l " PUMPS ; y N % % > h (P-18-1 A & 18 ) CH-V-286 CH-FCV-110 k' l CH-V-398A 3"-CH-2502R-8 l - D. P.A.B. PIPE TRENCH i EAST OUTSIDE P&ID - 16103-26018 SH. 6 UNDER VALVE ROOM ~[ PIPE TRENCH SYS. - CHEMICAL & VOLUME CONTROL (REACTOR COOLING SYSTEM CHARGING) NOTE - FOR PENETRATION DETAILS r ---- C. A.D. I l NoiE, l SEE DWG. #16103-59018 SH. 27 MANUAL REVISIONS TO TH3S AS-BUILT l DOCUMENT GED 3.01 FOR REVISION ARE PROHIB]TED. REFER lXXXXXXl DURINc TO l - INDICATES VALVE A g
, C_ONSTRUCTIO_N MANUAL
_REgDNS._ j TO BE TESTED ,1-),
@ "8 NORTHEAST UTILITIES SERVI CONNECT] CUT YANKEE ATOMIC POWER CO.
CONTAINMENT ISOLATION VALVE TEST DIAGRAM AS BUILT PER PENETRATION # 8 N/A 2 '2,65DCR-CY-S-494-89
% or RJC C*IL RGW w. WRK w. WRK 8 5-011 1 3
7 AS BUILT - PER " g un t-22-86 un 3-18-86 un3-20-86 kn 3-20-86 es DSR-CY-5-336-86 ,e
- m NONE NS 5 w a un mm.s er os_ w. w. w 83-oit 16103-20310 SH.11 MRO
U.S. Nuclear Regulatory Commission B13137/ Attachment 1/Page 5 April 28, 1989 Plant: Haddam Neck Plant
Title:
Exemption from 10CFR50, Appendix J, Section III.C.7 Penetration: P11A Drain Header Sample. P11B Pressurizer Steam Sample P11C Pressurizer Liquid Sample P11D Loop 1 Hot Leg Sample Description CYAPC0 requests a permanent exemption from the following specific Type C testing requirements of Appendix J in these sampling system penetrations. o Appendix J Section III.C.2 states that local leak rate tests (LLRTs) shall be performed with air or nitrogen. o Appendix J Section III.C.2 also states that LLRTs shall be performed at Pa (Pa - the calculated peak containment pressure during a design basis accident (DBA); Pa - 40 psig for Connecticut Yankee). Justification Bases The containment isolation valves (CIVs) in. penetrations P-11A through P-110 are leak tested with reactor coolant at 2000 psig. These CIVs (SS-A0V-950, SS-A0V-955, SS-TV-960, and SS-TV-965) are Masonellon, diaphragm actuated, globe valves installed so that reactor coolant system pressure opposes the force of their closure springs. This installation configuration tends to maximize any potential for valve leakage during a direct pressurization LLRT. A valve that can pass such an LLRT will leak substantially less, if at all, at l Pa. Appendix J testing of the CIVs in these penetrations is conservative in this regard. The results of 2000 psig water LLRTs are converted to an estimated equivalent air leakage at 40 psig in P-11A through P-11D. The NRC does not concur with such projections, other than permitting their use on an interim basis and has requested that modifications be made to permit 40 psig air testing for the affected CIVs. CYAPC0 acknowledges that the accomplishment of 40 psig air ' tests would yield an improvement in the accuracy of Appendix J testing for the affected CIVs. However, CYAPC0 questions the cost of the modificadons necessary for air testing, in dollars and man-rem in view of the small improvement in accuracy and public safety achieved by such modifications. A probabilistic risk assessment (PRA) of penetrations P-11A through P-11D was performed. This PRA conservatively determined that the present risk of public exposure due to core melt accident leakage through these penetrations is approximately 4 man-rem. This risk assessment did not assume any risk reduc-tion could be achieved through improvements in LLRT methods. Therefore, it
U.S. Nuclear Regulatory Commission B13137/ Attachment I/Page 6 April 28, 1989 can only be stated that Appendix J modifications would reduce the potential for public exposure due to leakage through the subject penetrations by a fraction of 4 man-rem. CYAPC0 considers the relatively small potential public exposure savings to be insignificant compared with the man-rem exposure and costs associated with making the required modifications to these penetrations. It has been esti-mated that approximately 12 man-rem of craft personnel exposure would be accumulated in making the required modifications. These changes will have an estimated cost of $388,000. Conclusion CYAPC0 believes that a permanent exemption from the Appendix J Type C testing requirements identified above are justified on these bases. l
d' .: Attachment I page 6a f
~
... c <f __ .
g ! w : e t d N- lE
li A D E
=
z --..
! 1 53 u (i!
.l
)
. ~:
8 !gs s!
! !s! ii j a g s
) Eg es 3 m e! :v- 1
. g*W -
vr- h i.si L- ' " l 5* ,lp-
._, v,
, ,, l. g -~ 5. g' 'r j ( fl
*-l 8 i
8 h*k (d ,
"0":"
N*p l8 j. 1 ---- {' N'y E E's Es
$53 - {8 /
4L -
- "l 4 e : e, ,! l c . ,-,
n.
.:p M.,e =
li ! EM l 3' ks: f l M$ a I'l ! g8 !e@d.. Yg I l (j $ 13 @,'d h' **r ' ll l. !! !S i., gl $_. i k! d d
. % ,E e
A- 6 gg ag a EQ p Y g" A fi k 8 kW ll a tg 4 le. ___ i g
!! 3 Z o
e i ,I L -
\ -
]z g' a g o
. ~ ) ,
g "' u
; A 5 *** ;
6 L wic 2 d d g* h - 4 Ik in Ct:: O'
'
- A LA.
( ~ gg! ,! ! =g Z g ; y u u a u ; u. ; u -
.1 is vS - m. gi
- b es; r3 ls ts e pe ,
!!!n 9
Irt;i e Iri; s-
. @-ict - u
- s%j < l -
!!k-
-2 !!!;v k t
l
, 1 i 1 1 i, l1i 1i la ,
M,llus.hr n ;
. ni e n I "
+
i *
- n'6 l s i A j *nl.
'n*.
h e
# n ni
, * ,, ! ng, A v 1,,A ir
' A r-- ir ,, , ,
L D_. J U u y [ 3 I s i-nF a "
~ ,
\
, T. ,A ni N I
~ I ~
j*.5li = l l ,,E1i>
. !!Et ai < h5 n$
EU. i
.. g5 4,
- __ u_.________ _ _ _
..o...,.
+
Attachment I page 6b
<> HCP TO N B SS-SOV-955 S
-#b gE
@"-SL-2505R-23#
>- a s .
i TO VOLUME I-- s MCB , CONTR L T NK E ATER SS-V-958 h - - - - - - - l- -)@( HEAT EXCHANGER lSS-TV-955 gy , OUTLET SS-V-957 -' "- S 2 M -10
', 4 M M < -D[< MN tuH ,
S- jk Z l"-SL-2505R-10 Ik 6A SS-V-95BC E SS-V-700 h p_.11B y SS-v-992
@"-SL-2505R-4B Y s
,, , l"-SL-2505R-46 FROM TO FROM PRESSURIZER PRESSURIZER PRESSURIZER LIQUID SAMPLE LIQUlD SAMPLE HEAT EXCHANGER HEAT EXCHANGER (E-9-18 ) (E- 9-1B )
OUTLET INLET d -
- P&lD - 16103-26009 SH.1 SYS. - PRIMARY SAMPLE (PRESSURIZER STEAM SAMPLE)
NDTE - FOR PENETRATION DETAILS SEE
~~~ DWG. #16103-59018 SH. 24 &
C.A.D. I 16103-22027 SH.1 I~OTE g N
~
[
] NT E BITE REFER l lXXXXXXl - INDICATES VALVE g cED 3.01 FOR REVISION DURINo l TO BE TESTED
, , CONSTRUCT 10N MANUAL REVISIONS. J , NORTHEAST UTILITIES SERVICE CO.
@ FEt COPNECT] CUT YANKEE ATOMIC POWER CO.
CONTAINMENT ISOLATION VALVE TEST DIAGRAM
*- AS BUILT PER PENETRATION # 11B N/A 2 "
2,g- DCR-CY-S-494-89 k er RJC C*D- RGM W. WRK AFP. WRK AS BUILT - PER g f an 1-28-86 DSR-CY-S-336-86 $e'g 85-011 1 7 Dait 3-18-86 dan 3-20-es oArt 3-20-es as 4 4
- m NONE 05
- 5 w m an avisum er a. w. w. w 85-o1: 16103-20310 SH.16 WRO r_--_-_-------------- --
Attachment I page 6c
~.
<$) HCP SS-SOV-960 S
] h MCB E
^gE
__! __ __ __ __ __ 3 l l j"-SL-2505R-12 # l c l"-SL-2505R-12 l 3.. TO PRESSUR1ZER LIQUlD EAT SAMPLE CHANGER
', l[
4 SS-V-961 (( 4[ fE_g_ FC PR-V-959 e g lSS-TV-960 l"-RC-2501R-58 g P-11C. C PRESSURIZER E-8-1 b
"5 :
g ' P&ID - 16103-26009 SH.1 O SYS. - PRIMARY SAMPLE (PRESSURIZER LIQUID SAMPLE) NOTE - FOR PENETRATION DETAILS SEE DWG. #16103-59018 SH. 24
& 16103-22027 SH.1 I~~~~C.A.D. lXXXXXXl - INDICATES VALVE I MANUAL REVISIONS l DOCUMENT ARE PROH3BITED. REFER TO TO TH]S AS-BU]LT l TO BE TESTED g GED 3.01 FOR REVISION DURING l
' A g CONSTRUCTION MANUAL REVISIONS. J Q
@ "" NORTHEAST CDPf4ECTICUT YANKEEUTILITIES ATOMIC POWER CO. SERVIC
" CONTAINMENT ISOLATION VALVE TEST DIAGRAM
+- AS BUILT PER " PENETRATION # 11C N/A 2 2$ DCR-CY-S-494-89 h n RJC C*D. RGM W. WRK *PP. WRK 85-011 1 3
7 AS BUILT - PER W es DSR-CY-S-336-86 ,o 4
'g 4g mt 1-28-86
- scu NONE part 3-18-86 04:t 3-20-86 m e-cart 3-20 5 ru io. mu avmm er os. w. w. ru 85-oin 16103-20310 SH.17 MRC
Attachment I page 6d h HCP MCB ! 3 _ _ MCB k1 SS-SOV-965 A/S gE E g__l_________3 ," FROM 27}"-RC-2501R-2 l l LOOP NO.1 HOT LEG j"-SL-2505R-14 i' j'-SL-2505R-14 i cs 3../ SS-V-964 TO HOT LEG SAMPLE
~_
f HEAT EXCHANGER ,4( l[ l[ 4 4 (
, SS-V-966 DE AY I
j - lSS-Tv-965} j'-RC-2501R-47 C . I P-11D , i d = j C
}--- = ! < [ P&ID - 16103-28009 SH.1 l %.:=
as SYS. - PRIMARY SAMPLE g (LOOP NO.1 HOT LEG SAMPLE) ! C I8 )4 NOTE - FOR PENETRATION DETAILS SEE ,
,,, DWG. #16103-59018 SH. 24 & i
[ 16103-22027 SH.1 )
,. - ._ _ _ ._ 5 I
NOTE: C.A.D. l lXXXXXXl - INDICATES VALVE I MANUAL REVISIONS TO TH]S AS-BU3LT TO BE TESTED I l DOCUMENT ARE PROHIBITED. REFER TO I $ g GED 3.01 FOR REVISION DURING l " d u S9".S.!R,gC}{M_WANU_AL,,,R{VIS,{m, s._ ,_) Jc
@
- NORTHEAST UTILITIES SERVIC CONNECTICUT YANKEE ATOMIC POWER CO.
CONTAINMENT ISOLATION VALVE , TEST DIAGRAM , gf, 2 yg_ AS BUILT PER " PENETRATION # 11D 3 89 DCR-CY-S-494-89 AS BUILT - PER h si RJC om RGW M. WRK APP. WRK L k 8 5-011 1 7 ss DSR-CY-S-336-06 W g #
"o 4 +g A
'an 1-28-86 an 3-18-86 mn 3-20-86 out 3-20-86 _ i 5 ru m an avras er xAtt NONE
- u. w. w. ru 85-011 osc. m 16103-20310 SH.18 f
unD i
p .. l l-c . l l U.S. Nuclear Regulatory Commission q B13137/ Attachment 1/Page 7 j April 28, 1989 ' Plant: Haddam Neck Plant I l
Title:
Exemption from 10CFR50, Appendix J, Section III.C.1 I i Penetration: P-14 Vapor Seal Head Tank Description CYAPC0 requests a permanent exemption from Section III.C.1 of 10CFR50 Appendix J in the Vapor Seal Hea( tank penetration. Specifically, CYAPC0 requests an exemption from the requirement that reverse direction Type C testing provide
" equivalent or more conservative results" than direct pressure testing, Containment isolation valve (CIV) DH-TV-1843 in P-14 is presently reverse direction local leak rate tested (LLRT) with nitrogen at 42-47 psig.
Justification Bases CIV DH-TV-1843 is a diaphragm actuated, double seated, globe valve with socket weld ends. Its direction of installation permits inclusion of the valve's body-to-bonnet joint and stem packing in reverse direction LLRT. Since this valve has a balanced port design (it is really a double plug flow control valve), the direction of LLRT or core melt accident pressurization should have no effect on disc / seat leak tightness. The inboard side of the subject CIV is routed to the vapor seal head tank. The head tank and the vapor seal lines are closed inside the containment. This is because the head tank is not vented, water from this tank is separated from the reactor coolant system by the No. I and No. 2 reactor coolant pump shaft seals, and water from this tank is separated from containment atmosphere by the No. 3 reactor coolant pump shaft seal. CYAPC0 acknowledges that modification of the plant to accomplish a direct pressure LLRT in P-14 would yield slightly more assurance of the subject CIV's performance during a core melt accident. However, CYAPC0 questions the cost of this slight increase in assurance in terms of dollars and craft personnel exposure required to make the associated plant modifications. A probabilistic risk assessment (PRA) of penetration P-14 has not been per-formed for this specific penetration. However, the required PRA would be similar to that performed for containment leakage through penetration P-78, the pressurizer relief tank vent. P-14 and P-78 connect upstream of CIV DH-TV-1844. The containment valves in these penetrations (DH-TV-1844, DH-TV-1843, and DH-TV-554) are local leak rate tested simultaneously. There-fore, it can be stated that the probabilistic public risk associated with reverse direction testing in P-14 is approximately the same as that calculated for reverse direction testing P-78. This estimated public probabilistic risk
)
l
- l. *o.
l, U.S. Nuclear Regulatory Commission B13137/ Attachment I/Page 8 April 28, 1989 is .04 man-rem. Modifications to accomplish Appendix J direct pressure testing could reduce this estimated public risk to approximately .004 man-rem. The relatively small potential public exposure saving must be compared to the exposure to craft personnel and cost in dollars in making the necessary plant modifications. CYAPC0 has estimated that 2.0 man-rem of craft personnel exposure and approximately $88,000 will be required to make the necessary plant modifications. Conclusion CYAPC0 does not believe that the slight potential decrease in public exposure risk achievable by modifying P-14 for compliance with Appendix J is justifi-i able on these bases. I
hs Attachment I page 8a HCP A,8 $ k - HCP A,8 i
,g s
> MCB C 7 TO PRIMARY l '
DRA]N TANK l-- - '@ TK-11-1A S 3
% DH-SOV-554 4DH-SOV-1844 A/S 4
2"-DRL-154-6 C
- I t
'l3 FC
#3 FC l H
2 zw & 2 ( FROM PRESSURIZER i , ,, , , > RELIEF TANK lDH-TV-1844l DH-TV-554 I TK-8 DH- -317 j 2"-DRL-152-6 [_l P.A.B. ! D" PIPE TRENCH = EAST OUTSIDE UNDER VALVE PIPE TRENCH bb Ik ROOM $.'T l . . , HCP A,8 j Lt i n DH-SOV-1843 h l REACTOR Q d A/S / P-14 COOLANT cb . PUMP
' ' FC l
{ 4, HEAD
,,,1 P N TANK r , m % L *'
lDH-TV-1843l l 1["-CRL-152-16 ( % DH-V-320 DH-V-315 (LC ) P&ID - 16103-26031 SH.2 (0UTSIDE) y 16103-26007 SH.1 (INSIDE) D. SYS, - MISC. SYS. REACTOR CONTAINMENT ! 7 (VAPER SEAL HEAD TANK DRAIN) ! l_ C.A.D. l NOTE - FOR PENETRATION DETAILS Il DOCUMENT MANUAL REVISIONS TO TH15 AS-BUILT l SEE DWG. *16103-59018 SH. 27 ARE PROH181TED. REFER TO g GED 3.01 FOR REVISION DURING I lXXXXXXl - INDICATES VALVE m {)i' CW u __ , dry,T,1,0,N,WA,N,U,AL,,, rey 5,,1,0),,,.S]
, TO BE TESTED Ja I
@ "ENORTHEAST CONNECTICUT YANKEEUTILITIES ATOMIC POWER CO. SERVI!
" CONTAINMENT ISOLATION VALVE d TEST DIAGRAM N/A 2 Ys AS BUILT PER Rrw PENETRATION # I4 89 DCR-CY-S-494-89 sr RJC C*D. RGM W. WRK NT. WRK 8 5-011 1 7 AS BUILT - PER j d'4 @ out 1-16-86 cart 3-18-86 oatt 3-20-86 part 3-20-e6 es DSR-CY-5-336-86 o +4
- Lua NONE ** i 5 w__ m an asum er ca. w. w. w e5-011_ 16103-20310 SH. 22 l ,
l p
- 1. U.S. Nuclear Regulatory Commission B13137/ Attachment I/Page 9 April 28, 1989 Plant: Haddam Neck Plant
Title:
Exemption from 10CFR50, Appendix J, Section III.C.2 Penetration: P-24 Safety Injection Recirculation , Description CYAPC0 requests a permanent exemption from the requirement of 10CFR50, Appen- ' dix J, Section III.C.2, to Type C leak test with air or nitrogen as the test medium for . valves SI-V-863A through D. These valves are the containment isolation valves (CIVs) of the high pressure safety injection (HPSI) recircu- , lation system. CYAPC0 presently Type C leak tests valves SI-V-863A through D with water as the test medium. Justification Bases The NRC does not concur with using a correlation of measured water leakage to air leakage. CYAPC0 acknowledges that the accomplishment of 40 psig air tests would yield an improvement in the accuracy of Appendix J testing for the affected CIVs. However, CYAPC0 questions the cost of the modifications necessary for air testing, in dollars and man-rem, in view of the small improvement in accuracy and public safety achieved by such modifications. During a periodic integrated leak rate test (ILRT), manual valves SI-V-863A through D are exposed to containment pressure throgh the reactor coolant system in the accident direction and with the back sics of the valves depres-surized. The periodic ILRT provides reasonable assurance that leakage through any penetration is limited. The HPSI system operates up to 8 hours after a large break loss-of-coolant accident (LOCA). This is long after the containment pressure peaks. Also, half of the thyroid dose due to containment leakage occurs during the first 15 minutes following the onset of a postulated accident. Any leakage of containment atmosphere through these lines to the environment would be through the refueling water storage tank (RWST) which would scrub iodine out of the leaking air. There exists one locked closed manually operated 3/4" globe valve in each of the four recirculation lines. These valves are Type C leak tested with water, i These manual valves will continue to be tested with water to identify any substantially leaking CIVs. The Staff has concluded that the Type C test with water and the use of the water to air conversion will provide reasonable assurance of the leakage integrity of these valves on an interim basis. l
U.S. Nuclear Regulatory Commission B13137/ Attachment I/Page 10 l April 28, 1989 System leakage to the environment is inspected for as part of the technical
~
specification limit of three liters / hour maximum leakage (10CFR100 offsite 4 doselimit). ) 1 The PRA analysis for this penetration calculates the public risk to be approx-imately 0.9 man-rem in its present configuration with the CIVs being leak tested with water.
'The man-rem expenditure required to complete the necessary modification is 1.0. The estimated cost is $100,000 to make the necessary plant modifica-tions.
Conclusion The increased environmental consequence due to Appendix J Type C leak testing penetration P-24 with water in lieu of air or nitrogen is negligible. The small decrease in public risk exposure does not support the cost in dollars and man-rem required to make the modifications. Therefore, CYAPC0 requests a permanent exemption from the . applicable requirements of Appendix J as der ibed above.
Attachment I page 10a
, n lXXXXXXl - INDICATES VALVE SAFETY SI-V-863B INJECTION CUBICLE P-24A (LC ) g=-SI-2501R-18 SI-V-863A +
l SI-V-863Dl-
$LC) {"-SI-2501R-17 g,,
N (LC ) ' ',
- v J
l2 SI-V-863C {"-SI-2501R-20 (\ i k" PI O) (LC ) \ - {"-SI-2501R-19
+ [l"-SI-2501R-20 '
l 656
- "-SI-2502R-25
~ - -866C
/ Q SI-CV-862D h
{ } SI-V-876A M 2" FROM i ( N vs
'I M N
1 IO tLOOP # 4
, (LO) ja CHARGING SI-MOV-861D COLD LEG
- FI FE PUMPS r 27f"-RC-2501R-14
- 651 651 '
(LO) h") {"-SI-2501R-18 {"-SI-2501R-17 Q # SI-V-876B ' 2" (C ) (D) O SI-V-868 a
= (
(A) (B )
/ (LC) 3"-SI-2501R-20 j"-SI-2501R-19
_ {"-SI-2502R-26 SECTION 1-1 g SI~[~869 2 LOOKING NORTH
> r , [ "-SI-2502R-25 AT THE P. A.B. I (LC ) '
I' SI-V-865 3" (LC ) P&ID - 16103-26010 SYS. - SAFETY INJECTION (RECIRCULATION) NOTE - FOR PENETRATION DETAILS Ir ,-,,--- C. A.D. l SEE DWG. #16103-59018 SH. 50
!EAIEEEpRoEed$'$lr'ER* "' I & 16103-22027 SH.1&2 '
g GED 3.01 FOR REVISION DURING l
, CONSTRUCTION MANUAL REVISIONS._ j gg NORTHEAST UTILITIES SERVICE CO. )
@ R8 CONNECTICUT YANKEE ATOMIC POWER CO.
I
" CONTAINMENT ISOLATION VALVE .
TEST DIAGRAM ) 4- AS BUILT PER PENETRATION # 24A N/A 2 285 DCR-CY-S-494-89 " b BY RJC DML RVK w. WRK M. WRK i 0EU~@ 8 5-011 1 3 7 es DSR-CY-S-336-88 "r o f
# h
+# h oAn 1-27-86 M1 NONE part 3-18-8s omit 3-20-86 oE n DAR 3-20-86 1
I 5 w m a um er u. w. w. w 85-011 16103-20310 SH. 35 l [ waO ]
l I Attachment I page 10b j SAFETY R lXXXXXXl - INDICATES VALVE l SI-V-863B TO BE TESTED INJECTION ] CUBICLE l (LC ) g=-SI-2501R-18 SI-V-863A p q(LC) g=-SI-2501R-17 P-24B l SI-V-863D + f_ (LC ) 1 (LC ) l SI-V-863C F- 3"-SI-2501R-20 ,,
!" :: D G i PI (LO) 3"-SI-2501R-19
- 656 %' -l"-SI-2502R-25 SI-V-866C 1
[]"-SI-2501R-19
~
n + - l- '2" SI-CV-862C (LO) }" f FROM ) r LOOP # 3 SI-MOV-861C p FI 651 FE 651 r3 L CHARGING PUMPS COLD LEG. f< 27{"-RC-2501R-9 (LO) b l
% {"-SI-2501R-18 {"-SI-2501R-17 :
SI-V-876B ' 2" (C ) (D ) { SI-V-868
'e .
(
~
J (A) (B) y) ~ ~ ~ ~ l j"-SI-2502R-26 l SI-V-869 SECTION 1-1
', y' g]"-SI-2502R-25 LOOKING NORTH i
(LC ) , j AT THE P.A.B.
,' SI-V-865 (1 3
(LC ) P&lD - 16103-26010 ; SYS. - SAFETY INJECTION (RECIRCULATION) NOTE - FOR PENETRATION DETAILS ["$uAt. REvlsiONs TO Tms I I As-sun.T- - ~~ {-Q ~ SEE DWG. #16103-59018 SH. 50
& 16103-22027 SH.1&2 ,
l DOCUMENT ARE PROH381TED. REFER TO lI g GED 3.01 FOR REVIs10N DURING
, CONSTRUCTION MANUAL REV]SIONS. ,j ya i
@ " NORTHEAST CONNECT] CUT YANKEF STOMIC UTILITIES POWER CO.SER)
" CONTAINMENT ISOLATION VALVE f TEST DIAGRAM 'l 4- AS BUILT PER PENETRATION # 248 i N/A 2 2s5 e DCR-CY-S-494-89 k 6 BY RJC C*R RVK M. WRK M. WRK il l
85-011 1 7 AS BulLT - PER "g 4 g oAn 1-27-86 oAn 3-18-86 DAR 3-20-86 ona 3-20-86 ss OSR-CY-S-336-86 o4 4
- scALL NONE Dm e I
$ w a om uses er os. w. e. w 85-011 16103-20310 SH. 36 {
___ ____ _______ _- - ____ _ __ S
~ '
p __ Attachment I page 10c 4 R SI-V-863B l (LC) SAFETY INJECTION r, u [$" h CUBICLE 3"-SI-2501R-18 f SI-V-863 A l (LC) L g=-SI-2501R-17 + SI-V-863D [ l O (LC ) L - SI-V-863C {" SI-2501R 1" [ l"-SI-2501R-18 PI (LO)9 (LC ) l"-SI-2501R-13 85S
~g -SI-2502R-25 '
'g SI-v 8sG[ SI-CV-862B _ } < N Lm i TO SI-V-876A = FROM i Vi rw tLOOP # 2 CHARGING SI-MOV-8618 COLD LEG { PUMPS ') ' i FI 651 FE 6 51 r L 27h"-RC-2501R-6 l"-SI-2501R-18 {"-SI-2501R-17 s (C ) (D ) C '$j=l , Q , SI-V-876B '2" SI-V-868 {"-SI- 1R-20 3"-SI- 1R-19 (LC) SECTION 1-1
}"-SI-2502R-26 LOOKING NORTH SI-V-869
]"-S1-2502R-25 (LC) P&ID - 16103-26010
_ ] [SI-V-865 (LC ) SYS. - SAFETY INJECTION (RECIRCULATION) NOTE - FOR PENETRATION DETAILS [~~~~{~~Q""~"~"~~~l SEE DWG. #16103-59018 SH. 50
& 16103-22027 SH.1&2 ITAIuat I RmstoNs To Tms As-suitT l T
ofpjEg IN l lXXXXXXl - INDICATES VALVE p b E "S.IRg][pN,MANU,,AL,_R{VIS,1,g,, ,,,g TO BE TESTED y,
@
- NORTHEAST UTILITIES SERVIC CONNECT] CUT YANKEE ATOMIC POWER CD.
CONTAINMENT ISOLATION VALVE TEST DIAGRAM
*- AS BUILT PER PENETRATION # 24C N/A 2 2 85 DCR-CY-S-494-89 AS BUILT - PER "
h er RJC om RVK w. WRK W. WRK 85-011 7 & A f art 1-27-86 cart 3-18-86 mil 3-20-B6 cait 3-20-86 DSR-CY-S-336-86 "a 4 C A Sud NONE 1 as **
$ ru a un mimons er on. w. w. tv 85-011 16103-20310 SH. 37,
_ _ _ _ ___ _ ______ embl
o . __ Attachment I page 10d SI-V-863B + LC) lSI-V-863AF 3"-SI-2501R-18 ,,
N 3- (LC )
,,t- - O
" J SAFETY INJECTION SI-V-863D ]" SI-2501R-17 k
\
CUBICLE Z " g ) (LC ) SI-V-863C 3"-SI-2501R-20 +
- PI 1"t (LO) 1 p(LC) ("-SI-2501R-19 /~ l"-SI- 2501R-17
", 66 ~~
SI-V 866C M b SI-CV-862A
~ O FROM i g
N i TO
- SI-V-876A VI rw iLOOP # 1 o
g '2" CHARGING SI-MOV-861 A COLD LEG
% (LO)fa_f PUMPS 27d"-RC-2501R-13 FI FE
; 651 651 rg l-
=
{"-SI-2501R-18 {"-SI-2501R-17 r" n (LO)h"l M (C ) CD) 1 4 SI-V-8768 ' 2" ' l 7 ' SI-V-868 2 3"-SI-2501R-20 3"-SI-2501R-19 (LC ) SECTION 1-1 {"-SI-2502R-26 LOOKING NORTH SI- V-869 A T THE P. A.B. j ;' l"-SI-2502R-25 (LC ) , s
, SI-V 65
{" ' C) P&ID - 16103-26010 SYS. - SAFETY INJECTION (RECIRCULATION) NOTE - FOR PENETRATION DETAILS SEE ir -- - - -C.A.D. I Dwo. 416103-59018 SH. 50 4 { l! ACED UAL REVISIONS TO TH]S AS-BUILT DOCUMENT 3.01 FOR REVIS]DN DURING ARE I PROR]BITED. & 416103-22027 REFER TO SH.1 l &2 m
- j luCONSTRUCTION MANUAL REVISIONS. lXXXXXXl - INDICATES VALVE
,_j TO BE TESTED Jo
@ '* NORTHEAST UTILITIES SERVICE CO, CONNECT] CUT YANKEE ATOMIC POWER CO.
CONTAINMENT ISOLATION VALVE 3 TEST DIAGRAM N/A 2 Y6 AS BUILT PER Rrw I e9 DCR-CY-S-494-89 3 nr RJC DFA RVK W. WRK W. WRK 8 5-011 AS BUILT - PER " fg an 1-27-86 mit 3-18-66 an 3-20-86 "o +p +g A 1 7 att3-20-86 86 DSR-CY-S-336-86 W NONE w- aa
$ w in an amam er a e. w. eu 85-o33 16103-20310 SH. 38 wRD
1 U.S. Nuclear Regulatory Commission B13137/ Attachment I/Page 11 April 28, 1989 j l Plant: Haddam Neck Plant l l
Title:
Exemption from 10CFR50, Appendix J, Section III.C.2 j P-28 Component Cooling Water from Reactor Coolant Pump 011 Penetration: Cooler Description i CYAPC0 requests a permanent exemption from the requirement of Appendix J, Section III.C.2 that local leak rate tests (LLRTs) be performed with air or nitrogen. CYAPC0 presently performs the LLRT for containment isolation valve (CIV) CC-CV-853 with 44 psig water. The water leakage collected in this direct pressurization test (i.e., P t applied in same direction as Pa) on this component cooling water penetration is converted to an equivalent air leakage at 40 psig. Justification Bases The component cooling water (CCW) system is a water filled closed loop system. It is not seismically designed. Its normal operating pressure is 70 psig. While it can not be assured that the piping loop through P-28 will remain water filled or operational during a core melt accident, it can be stated that leakage out of this penetration would not be possible during a core melt accident with the system intact and functional. If the loop inside contain-ment is postulated to rupture and drain during a core melt accident, it is possible that containment core melt accident atmosphere leakage through CC-CV-853 would be retained by the closed CCW system outside containment. The CCW surge tank will close upon sensing radiation if functional after the postulated core melt accident. This vent is the only potential leakage path outside of containment if the system remains intact. The NRC does not concur with the correlation of water leakage to air or nitrogen leakage, other than permitting such correlation on an interim basis. CYAPC0 acknowledges that the accomplishment of a 40 psig air or nitrogen leakage test would yield a more accurate assessment of how CC-CV-853 would perform during a core melt accident. However, CYAPC0 questions the cost of the modifications necessary for air testing, in dollars and man-rem, in view of the small improvement in accuracy and public safety achieved by such modifications. A probabilistic risk assessment (PRA) of penetration P-28 has been performed. This PRA conservatively determined that the present risk of public exposure due to core melt accident leakage through this penetration is approximately
.0002 man-rem. Modifications to accomplish Appendix J air testing would reduce this public risk to approximately .00002 man-rem.
1 '~. U.S. Nuclear Regulatory Commission B13137/ Attachment I/Page 12 i April 28, 1989 This relatively small potential.public exposure saving must be compared to the dosage and cost required to make the plant changes necessary for its accom-plishment. It has been estimated that 2.0 man-rem of craft personnel exposure will be accumulated in making the required plant changes. These changes will have an estimated cost of $127,000. Conclusion Based on the analysis above, CYAPC0 requests a permanent exemption from the. requirements of 10CFR50, Appendix J, Section III.C.2. 1 i
)
i l
)
1
5- I V R nt N 1 5 1
- E S Te IO r
oW sK D S a A T 5 -* 5 E I aLO 8 1 5 T 9 v
- 8 I
L a mS I 44 - V I v
- - C - . T 17 51 C C C A U mTN 1P W
H r 1 l r T C S i t E KO TT 8 3 '9 9 4 m a c A EmI T R t C CN T A 4
- gy*
T4 S3 1
=
r D M m N 1 T41 5 s O 5 3 1 ro 1 W a C c cW s K C K
- e u 5 - 5
- 8 *1 5 3 m "
,u
- 8 e 7 v- -
43 C V s 17 C C Nv e 51 C m 1P W R v v a KO-T T s I 4
- 33 14 9
s3 9 5 1 IIm' __ 3 1 5 3 1 T41 1 W
- __mu 8 w K s g 5 -
e *3 v 8 5 5 8 4 m_m% a 5 42 C
- V QqmF"_
" o 11 C C Nv ' c 51 C 1P W
R r 1
-{ _a~" _
KO TT 4
- r m-u 4 9 9 53 -
g 5 3 332 14 3 141 4 [e ema g L A w 3 sg A 5 5 8 =- g 5 5 3
- . 8 w 3 v - g 41 - V a
- - C - g 17 C C 51 C 1P W
R " KO T T 9 9 E 33' s3 D I H 14 1 i41 S C T N
'0 U E DR E
$.k T
M OR E jI T S E OT 4- EI P 1 1 4 R 1 3- WP 1 .
- l S N 1 8
S
/
t 1 4 A 1 0 WW5 O W 8 D'iP. K ) NA C 6 1 V
- 5 8
V
- WE 2 l !l i*I S R
- I OI '6 - 7 C .
l T C C 7Mp9 gP V C C
- IL C C
- L B~P yu HM E
L O S H 2 [ 7 A 5
- CO NO O
C L S I
~
3
- C 3 5 ER A8 F 8 8 R L T1
- TE I
- 8 V
V C
- EL P A V
5NI G O. E0 D9 5 C V C C 1P. R . L P. N - O3 ( H O C. M j B. E A. PL O P S OR 8 C I 0 T1 A6 1 2- 5 M R 0TO T # 1 5 1 I' 0 6N ER F E NG Y W 2N . EW T M NT E C' C K
*- 3- OP W .
PD ES ) 0MC 1 RE NY 1 6 C O (C . OE 0 r$ '6 $], 1 FS hG 0N CI j s_ L 3
, D . E OO ,. I S T TO &Y O C ,
PS N 3, l 7, MOW
U.S. Nuclear Regulatory Commission B13137/ Attachment I/Page 13 April 28, 1989 Plant: Haddam Neck Plant
Title:
Exemption from 10CFR50, Appendix J, Section III.C.1 Penetration: P-33 Refueling Cavity Purification Description CYAPC0. requests a permanent exemption from the requirement of 10CFR50 Appen-dix J, Section III.C.1 that reverse direction Type C testing provide "equiva-lent or more conservative results" than direct pressure testing. CYAPC0 presently Type C leak tests containment isolation valve (CIV; PU-V-242 in this refueling cavity purification penetration in the reverse direction. CYAPC0 feels that the reverse direction testing of the diaphragm to seat seal gives , equivalent results as accident direction testing. However, reverse direction testing may not include a body to diaphragm seal on the containment side of PU-V-242. An ' exemption to Type C testing is required for these containment-boundaries. Justification Bases The subject CIV is a manual diaphragm valve with welded ends. 'When the valve is closed, a seal is created between the diaphragm and the seat by means of a hand wheel which drives a shaft down on the diaphragm and onto the seat. If there are any cracks in the diaphragm or any dirt particles on the diaphragm, it' is possible that the diaphragm will not seat. In this case, the valve would fail either a direct pressure or a reverse direction pressure test. As stated above, the portion of the diaphragm on the containment side would not be pressurized and any body to diaphragm joint leaks would not be detected during a reverse direction test. CYAPC0 proposes to pressurize the containment side of the valve during the periodic integrated leak rate test (ILRT) (through the high suction line of - the refueling cavity) at Pa. The periodic ILRT will provide reasonable , assurance that gross leakage through this valve to the environment will not occur. PU-V-242 and PU-V-242A (which is accident direction / direct pressure tested) are locked closed during plant operation. They are only opened during refuel-ing operations when the refueling cavity is filled with water. CYAPC0 acknowledges that direct pressure testing could provide greater assur-ance that PU-V-242 will function acceptably as a leakage barrier during a core melt accident. However, CYAPC0 questions the cost of this slight increase in assurance when it is compared to the dollars and craft personnel exposure required to make the associated plant modifications. 1
1
.. j U.S. Nuclear Regulatory Commission B13137/.tttachment I/Page 14 April 28, 1989 l
A probabilistic risk assessment (PRA) of penetration P-33 has been performed. l The PRA determined that the present risk of public exposure due to core melt leakage is approximately 0.3 man-rem. Modifications to accomplish Appendix J air testing could reduce this estimated public risk to approximately .03 man-rem. CYAPC0 considers the relatively small potential public exposure saving to be insignificant compared with the man-rem exposure associated with making the required modification. It has been estimated that 5.5 man-rem of craft personnel exposure will be accumulated in making the required changes. These changes have an estimated cost of $120,000. Conclusion CYAPC0 does not believe that the slight decrease in potential public exposure risk by modifying P-33 for compliance with Appendix J is justifiable on these bases. Therefore, CYAPC0 requests a permanent exemption from the requirements of 10CFR50, Appendix J as described above. CYAPC0 also proposes to test PU-V-242 to periodic ILRT pressures as the best means of assuring this valve's leak tightness with the existing plant configuration.
I 4 V C T 2 R OW E O
- S T 2 A s S t
E EE L I L R T I mO 2 4 C-L I T wI
} U TUT Rt A C P
o 1 T I s l S T A CE M t o W i L
- g e t s
o i Ele N I I w - l t T l o A t R DR v e - w R 3 0 0 7 u O C T T S "3 m 2 9 N" F - - yY tS t rM o l { v o v-o m ' C aD w w gI - m ' T C E '~ [
]
r 1 _g- ~ h3 r i a 0 gP JE s e 1 s, i
,.. - - 0 v-
] 2
- I m v- u qY
,H t
s *2 - o V 2 w !J A 3 s t
~g r -
C a""
- 6" l ~ -
w - R 0 4 3
- 1 4
s 2
~ m"mM* s a s s
2- , 4 - Vg, u fe -ev l 4 5 {m o ".- QtE Ia u 1 P 1 P L R
- T N
E s n = T-D g 7
,I W "
m*m"- r A 1N -
-2
" T O
N# C
" wE'T
===7 a "-
2 2 R . 5 s O Il 1 L i eT C P
- m, e A E
R 7 r 3 3
-[ i,( [
P 2 4 2 )C vL 3 G 4 P 2 i 2 1 vq-Q ,, R u t s P
1
'rH +1 C
y N O 7 LI T 2 O h' A 2
" N n 3 4
R T C N F A I S LS I H 4 R OI A8 2- lC t CR T1 u 5 U V (
- o s' 5 1 E P E0 D9 s 5 V 5 N 2 2
- h 2 U Y N -
l v O 'F vO 3 H
. L O I T
O 03 V V 1 I
- u A S T 1 2 P B A A 6 5 8 C 1
t3 1 & R 1 ' L
- t 0 G T
- P N 6 L N E .
0I 2 A I N G Y 3 1AI M C L EW N T - I E P D A-C2 1 v 1 3 0 M U R E
!1 r-i5 R0 6
1 6 1 E F H E C R O E F S U PP M D$ F - - - I OU TP D I S
. E T
& Y P S O
N I d[DC hI , u ii
U.S. Nuclear Regulatory Commission B13137/ Attachment I/Page 15 April 28, 1989 Plant: Haddam Neck Plant
Title:
Exemption from 10CFR50, Appendix J, Section III.C.2 Penetration: P-34 Component Cooling Water from Reactor Coolant Pump Thermal Barrier Description CYAPC0 requests a permanent exemption from. the requirements of 10CFR50, Appendix J, Section III.C.2 to Type C leak test, with air or nitrogen as the test medium, containment isolation valve (CIV) CC-FCV-608. Penetration P-34 is the component cooling water from the RCP thermal barrier. CYAPC0 presently Type C leak tests CC-FCV-608 with water. Justification Bases The component cooling water (CCW) system is a nonseismic closed system inside containment. The normal pressure within this system is in excess of peak loss-of-coolant accident (LOCA) pressure. The CCW system is also a nonseismic-closed .,ystem outside containment with the exception of the CCW surge tank vent which will automatically close on high radioactivity in the vent line. The CCW system is water filled at all times during plant operation and will likely remain water filled during the early stages of a LOCA. Half of the thyroid dose due to containment leakage occurs during the first 15 minutes following the onset of a postulated accident. The NRC does not concur with the correlation of measured water leakage to air leakage. CYAPC0 acknowledges that the accomplishment of 40 psig air tests would yield an improvement in the accuracy of Appendix J testing for the affected CIV. However, CYAPC0 questions the cost of the modifications neces-sary for air testing in dollars and man-rem in view of the small improvement in accuracy and public safety achieved by such modifications. During a periodic integrated leak rate test (ILRT) valve CC FCV-608 is exposed to containment pressure through a vent connection inside containment in the accident direction and with its backside depressurized. The periodic ILRT provides a reasonable assurance that leakage through any penetration is limited. The Type C test with water will continue to be performed which will accurately identify any significant leakage of valve CC-FCV-608. The PRA analysis for this penetration calculates the public risk to be approx-imately 0.7 man-rem in its present configuration with its CIV being leak tested with water.
U.S. Nuclear Regulatory Commission B13137/ Attachment I/Page 16 April 28, 1989 The man-rom expenditure required to complete the modification is approximately 1.0 man-rem. The estimated cost to modify P-34 is $102,000. By letter dated September 29, 1987(7) a schedular exemption for P-34 was. granted in which the Staff agreed that the Type C test performed with water will provide, in the interim, a reasonable assurance of the leakace integrity. of the valve. Conclusion The increased off-site thyroid dose due to Type C leak testing penetration P-34 with water in lieu of air or nitrogen is negligible. The small decrease in public risk exposure does not support the costs in man-rem exposure and dollars required to make the modifications. Therefore, CYAPC0 requests a permanent exemption from the requirements of Appendix J, Section III.C.2 as described above. 1 (7) D. M. Crutchfield letter to E. J. Mroczka " Exemption from certain Requirements of 10CFR Part 50, Appendix J," dated September 29, 1987.
V C R E tu S t
,r t a
1 4 S 8 E I tt 6 T 91 V 0 4 I IL v aS N C 8 - T 2 C - U iu V 5 1 k d M - T S T1 c C C C 5' i A c A "6 [
] y -
- EwI T
l e R o 9 ,- O C
, a 3
8 N* V C C gOn { E F 1 Ce Ie Fs -
) 8 W8 s f
A EE 1 3 8 V C C f J. iI yt h s2 8 0 6 M O - $
" ,, m
- g. " , m " .
m 5_ V J.e ' V
- O R
%, n*_ nL 5
0 C C f N W O
- c. , e C _
C , 3 _
, g C D W
O L B c ""
- t",,w t
, m" r
t m l I g enu E"*L 2 4 dC S N H - r ,y 4 8 T E U TR 3 - G E
- T P C SI L
KP l 2 31 NT 3I
-- 3 ' j/
0-5P 2
-W M A v
CD 7 AR 7 F
" 7 2 -
V
- )
C R 3 C E 32 - 6 N-7 3 RI 3 N R 01 A 5F 2
-W M v 3 5
0 B 7 2 CO 8 2 L . 7 A AR- 2 C SHS F 7 A M "2 V R R
- *3 E A8 4 C H T1 33 C G
T E0 N-7 . D95 3 0-1 SNP.
.IL N -
5P 2 M v H3C. S + - . O3 I 0 E
-M C T1 V CO AR 7
2 R 0 'C M A 6 1 L A D
- F 7 C *2 R V TE 2" - 0 TO T #
V D 0 6 EF NR E . S S C N G E TE 5 C 2- N . E W T A E 3 4 OW P D C 3 0 P . N-7 M C. R E DIB 3 1 6 OC O E NO 0-1 1 C( F S IT 5P 2 CO
-M MD v - ' -
l x x AR 7 D . E x F 2 l S T x "2 7
- &Y O x x
V PS N l C C I. O Q-
U.S. Nuclear Regulatory Commission B13137/ Attachment I/Page 17 April 28, 1989 l Plant: Haddam Neck Plant
Title:
Exemption from 10CFR50, Appendix J, Section III.C.2 l l Penetration: P-61 Component Cooling Water from Neutron Shield Tank Cooler Description CYAPC0 requests a permanent exemption from the requirement of Section III.C.2 of Appendix J which states that local leak rate tests (LLRTs) shall be per-formed with air or nitrogen. CYAPC0 presently tests containment isolation valve (CIV) CC-TV-1831 in this component cooling water penetration (for the neutron shield tank) with 44 psig water. The water leakage collected is converted to equivalent air leakage at 40 psig. Justification Bases The component cooling water (CCW) system is a closed loop inside and outside containment. However, it is not QA Category I or seismically designed. The pressure in this closed loop is greater than DBA pressure when the CCW system is operating. The only normally open release point in the system, the CCW surge tank vent, closes automatically when radioactivity is sensed in the vent line. Leakage out of containment through P-61 is not possible with the CCW system intact and functioning (i.e., lines at pressure). The NRC does not concur with the correlation of measured water leakage to air leakage, other than permitting such correlation on an interim basis. CYAPC0 acknowledges that the accomplishment of 40 psig air tests would yield an improvement in the accuracy of Appendix J testing for the affected CIV. However, CYAPC0 questions the cost of the modifications necessary for air testing in dollars and man-rem in view of the small improvement in accuracy and public safety achieved by such modifications. A probabilistic risk assessment (PRA) of penetration P-61 has been performed. This PRA conservatively determined that the present risk of public exposure due to core melt accident leakage through this penetration is 0.7 man-rem. This analysis of public risk is dominated by the assumptions of a relatively high probability of a breach of pipi.ng inside and outside containment, and failure of the CIV to close. Since this scenario does not count on the leak tightness of the CIV, there is no decrease in public risk due to modifying the plant to perform an air or nitrogen LLRT. If lines are not assumed to rupture and CC-TV-1831 is assumed to close, modifications to permit air testing would yield a small public risk reduction. This risk reduction would be smaller than the public risk of approximately 0.7 man-rem calculated by the PRA. This relatively small potential public risk saving must be compared to the man-rem exposure and cost required to make the plant changes necessary for its accomplishment. It has been estimated that 1.0 man-rem of craft personnel
J f: , .
~ U.S.' Nuclear Regulatory Commission
- B13137/ Attachment I/Page 18 April 28, 1989 exposure will be . accumulated in making the required plant. changes. These changes will have an estimated cost of $90,000.
Conclusion CYAPC0 ' requests a permanent exemption from the requirements of Appendix J, Section III.C.2 as described above. l. 4 l _ _ _ - _ _ _ _ _ _ _ _ - ___ _ _ _ _ _ _ _ _ _ _ - _ _ . _ _ . _ _ _ _ _ _ _ . - ________ _ _-_ ___________________ __a
Attachment I page 18a h N .
~
CC-SOV-20TV-1831
}
'7 D __ C 1 h - - % HCP A,B 1 i U1 h O
W I' 1{"-KW-151-2 \ 1 3"
,c I
CC-T V-1831 l
~
FROM NEUTRON b N C l FC ( M $ H ELD TANK COOLER CC-V-883
CC-V-887 1h" ! (L )
M- ; 5 O : L_J ; M WEST OUTSIDE : BLOWDOWN ROOM ,
,p PIPE TRENCH , ;
~%
i P&ID - 16103-28008 SH. 5 ; SYS. - COMPONENT COOLING (C.C.W. FROM NEUTRON SHIELD TANK COOLER) t NOTE - FOR PENETRATION DETAILS i SEE DWG. #16103-59018 SH. 27 l lXXXXXXl - INDWATES VALVE
"~~~~D.A=D=
I gg, g TO BE TESTED I MANUAL REVISIONS TO TH]S AS-Bu]LT g m lg DOCUMENT GED 3.01 FOR REVISION AREDURING PROH]BITED. REFER TO l s.C,0,,NyR_UCT10t1 MAN _UAL,RE_VIS,1,0NS.j da NORTHEAST UTILITIES SERVICE CD. j .
- CONNECTICUT YANKEE ATOMIC POWER CO.
" CONTAINMENT ISOLATION VALVE TEST DIAGRAM
*- AS BUILT PER PENETRATION # 61 N/A 22s5 e DCR-CY-S-494-89 h BT RJC C*D.
f an 1-23-88 o*it 3-18-86 RWC APP. WRK NY. WRK 3 AS BUIL1 - PER oan s-20-86 DAIL 3-2Q-66 DSR-CY-5-336-88 $o '+p +g A M NONE 85-021 1 7 ss M 4 ce c3 m cnE3 w n a a to 85-011 16103 _20310 SH. 59 l
U.S. Nuclear Regulatory Commission B13137/ Attachment I/Page 19 April 28,-1989 Plant: Haddam Neck Plant
Title:
Exemption'from 10CFR50, Appendix J, Section III.C.I. Penetration: P-62 Service Air to Containment 1 Description l CYAPC0 requests a permanent exemption from compliance with Section III.C.1 of 10CFR50, Appendix J, .in this service air system penetration. Specifically, CYAPC0 requests an exemption from the requirement that reverse direction Type C testing provide " equivalent or more conservative results" than direct. pressure testing. l By letter dated September 29, 1987,(8) the Staff has acknowledged the imple-mentation of a 15 foot-pound closure torque on containment isolation valve (CIV) SA-V-413 in P-62 as a means of assuring conservative reverse direction Type C testing. However, this justification did not encompass the reverse directiontestingconcerqsraisedbytheStaffduringaroutineinspectionat Millstone Unit No. 1.g Specifically, the . Staff was concerned about the possible omission of potential containment g daries due to reverse direction testing. By letter dated August 31, 1987, CYAPC0 provided the Staff with an action plan to address this concern. Specifically, "for valves where reverse direction testing is nonconservative, alternate resolutions will be developed (e.g., exemptions, plant modifications, or additional surveillance programs)." CYAPC0 hereby submits this exemption request to resolve this inspection report issue. (8) D. M. Crutchfield letter to E. J. Mroczka, " Exemption from Certain Requirements of 10CFR Part 50, Appendix J," dated September 29, 1987. (9) W. V. Johnston letter to E. J. Mroczka, " Inspection No. 50-245/87-18," dated September 23, 1987. During the inspection, a meeting was held with CYAPC0, NNEC0, and NRC personnel to discuss the LLRT of selected CIVs at the Haddam Neck Plant and Millstone liuclear Power Station Unit Nos.1, 2, and 3. At the conclusion of the meeting, CYAPC0 and NNEC0 agreed to review the Appendix J testing program at the Haddam Neck Plant and Millstone Nuclear Power Station Unit Nos.1, 2, and 3, and provide the Staff with an action plan to address the issue. (See the E. J. Mroczka letter to W. T. Russell, dated August 31,1987.) (10) E. J. Mroczka letter to W. T. Russell, "Haddam Neck Plant, Millstone Nuclear Power Station, Unit Nos.1, 2, and 3 - 10CFR50, Apperdix J," dated August 31, 1987.
I U.S. Nuclear Regulatory Commission B13137/ Attachment I/Page 20 , April 28, 1989 Justification Bases CIV SA-V-413 is a solid wedge gate valve with female thread connections. As such, it has inboard and outboard disc / seat interfaces. It is possible that this valve's threaded body-to-bonnet joint and its stem packing may not be pressurized during a local leak rate test (LLRT), regardless of direction of pressurization, or a core melt accident. If the inboard seat is not leak tight, these other pressure retaining boundaries may become potential contain-ment isolation boundaries. CIV SA-V-413 is not likely to be exposed to containment pressure during a periodic integrated leak rate test (ILRT) even though the service air system is vented to containment atmosphere by opening SA-V-430. This is because check valves SA-CV-417 and SA-CV-415 will isolate SA-V-413 if they are leak tight. SA-V-413 is opened and pressurized to 100 psig during refueling outages. The leak tightness of the body-to-bonnet joint and the stem packing are function-ally demonstrated at this time. Since this CIV is in an accessible section of the Primary Auxiliary Building, CYAPC0 proposes to soap bubble check all of its joints (including its threaded ends) during every refueling outage. A probabilistic risk assessment (PRA) analysis of P-62 has been performed. This PRA conservatively determined that the present risk of public exposure due to core melt accident leakage through this penetration is 0.2 man-rem. This potential public risk exposure is small when compared to the costs in money and man-rem required to modify this penetration to comply with Appendix J requirements. It has been estimated that 1.0 man-rem of craft personnel exposure and $126,000 would be expended in making modifications to permit accident direction testing of the subject CIV. ; Conclusion CYAPC0 requests a permanent exemption from the reverse direction testing requirement of Appendix J as described above based on the additional surveil-lance that will be performed to meet the intent of 10CFR50, Appendix J.
Attachment I pace 20a SA-V-430] [ P&ID - 16103-26051 SH. 3 {"-ASC-151-108 -- SYS. - SERVICE AIR S A- V-431 r (SERVICE AIR ( LJ 3 Q TO CONTAINMENT) r, i SA-V-700 NOTE - FOR PENETRATION DETAILS ("- ASC-151;4) 1 SEE DWG. #16103-59018 SH. 27 p g3_Q_'432
& 16103-22027 SH.1&2 j
SA-{-433A g lXXXXXXl - INDICATES VALVE i r, TO BE TESTED l7l. .P . SA-V-433 3, {" e 2"- ASC-151-40 l = t L2 4 r -I 1"- ASC- 1 SA-V-413 l P 7 t:--- .=s, SA-V-621 % SA-CV-417 u u L2 < f""
% ==
2"- ASC-151-109h } 'N IN "
^
)SA-CV-415 S A-V-414 L)
["" ~.a 4 1"- ASC-151 / ( r, I ,,s i SA-V-419 ' 2"-ASC-151-14 i rg ,==i = ( L2
> r, e -I w
(L ) 3 sLr A-V-411 S (L )
* '>=,
d m, ( L' SA-V-423 SA-V-428 i ri C ,=,=,=ss4 m SA-V-418
}"- ASC-151-36 e
( )7 {"- ASC-151-33 3 r j g SA-V-622 SA-V-429 L6 1"- ASC-151-34 SA-V-422 t r {,- ASC-151-20
# 8 b^- -420 ,
,' t, g2 ,,,,, 1"- ASC-151-42 .+ SA-V-624 i r, N '1"- ASC-151-35 s D
( gj 3 1"- ASC-151-37 " i ri ga 1"- ASC-151-38 " 2"- ASC-151-14 SA-V-421 M SA-V-427 j" b SA-V-424 SA-V-425
~
' {"- ASC-151-3 r ---- C. A.D.
i NOTE: I MANUAL REVISIONS TO TH]S AS-BU]LT l l DOCUMENT ARE PROHIBJTED. REFER TD g g GED 3.01 FOR REVISIDH DURING l
, CONSTRUCTION MANUAL REVISIONS. j g
@ " NORTHEAST UTILITIES SERVICE CD,
" CONNECT] CUT YANKEE ATOMIC POWER CO.
CONTAINMENT ISOLATION VALVE TEST DIAGRAM 4- AS BUILT PER PENETRATION # 62 N/A 2 26 35 DCR-CY-S-494-89 AS BUILT - PER h Bf RJC C*D. RVK w. WRK M. WRK 85-011 1 3 7 as DSR-CY-S-336-86 $e 4 4 hg scad NONE g g f an 1-25-86 un 3-18-8s un a-20-B6 un a-20-es
- 15. ao.
$ re m an rymms er or,. w. w. Fu 85-011 16103-20310 SH. 60
_ _ _ _ _ _ _ _ _ _ . - - c'rsbl
4 U.S. Nuclear Regulatory Commission B13137/ Attachment I/Page 21 April 28, 1989 Plant: Haddam Neck Plant l
Title:
Exemption from 10CFR50, Appendix J, Sections III.C.1 and III.C.2 Penetration: P-63 Neutron Shield Tank Fill Description CYAPC0 requests permanent exemption from the fo6Naing specific Type C testing requirements of Appendix J in the neutron shield tank fill penetration P-63. o Appendix J, Section III.C.1 states that reverse direction testing is acceptable if it provides " equivalent or more conservative results" than direct pressure testing. o Appendix J, Section III.C.2 states that local leak rate tests (LLRTs) shall be performed with air or nitroger.. o Appendix J, Section III.C.2 also states that LLRTs shall be performed at Pa (the peak containment pressure during a DBA; Pa - 40 psig). The containment isolation valve (CIV) in P-63 is leak tested with component cooling water at 70 psig. LLRT pressure is applied in a direction which is the reverse of DBA pressurization. The CIV is identified'as CC-V-884. It is a Grinnell Weir diaphragm valve with socket weld ends. Justification Bases The results of 70 psig water leakage testing of CIV CC-V-884 are converted to equivalent air leakage at 40 psig. The NRC does not concur with any methods of correlating such leakage, other than permitting such correlation on an interim basis. CYAPC0 acknowledges that modification of the plant to accom-plish a 40 psig direct pressure air or nitrogen LLRT would yield a more accurate assessment of the subject valve's performance during a core melt accident. However, CYAPC0 questions the cost of this slight increase in accuracy when it is compared to craft personnel exposure required to make the i associated plant modifications. A probabilistic risk assessment (PRA) of penetration P-63 has been performed. It concluded that the present risk of public exposure due to core melt leakage through this penetration is approximately .00007 man-rem. Modifying this penetration to permit compliance with Appendix J would reduce public risk to approximately .000007 man-rem. This small potential public risk saving must be compared to the man-rem exposure and cost required to make the plant changes necessary for its accom-plishment. It has been estimated that 1.5 man-rem of craft personnel exposure
4 i 1 i U.S. Nuclear Regulatory Commission B13137/ Attachment I/Page 22 April 28, 1989 will be accumulated in making the required plant changes. The estimated cost l of the modifications is $131,000. With regard to reverse direction LLRT, it must be noted that such testing does l not include a portion of the subject CIVs body-to-bonnet joint. This gasketed j joint can act as a containment boundary during a core melt accident. j However, the subject LLRT does include a separate step which opens the CIV and exposes its entire body-to-bonnet joint to 70 psig water pressurization. This leakage test accomplishes the objective of Appendix J even though it does not comply with Appendix J. With respect to disc / seat leakages, it can be stated that a Weir diaphragm valve has a single seat and symmetrical inter-nal s . It should not be any more leak tight when pressurized in one direction than the other. Conclusun CYAPC0 requests a permanent exemption from the requirements Appendix J, Sections III.C.1 and III.C.2 as discussed above.
l Attachment I page 22a CC-V-851 CC-V-882 1h'- AC-152N-18 1}"-KW-151-52 % 6,- AC-152N-29
-V-730 I FROM < m '
r r ( TO PEN. I CCW i r i seg "x 3" 3'- AC-152N-93 y 3"- AC-2503N-31> A "-KW-151-3 6 CC-V-728 CC-V-852 Ij"- AC-152N-17 h [IlI CC-V-884A i / TO E-58-1 lCC-V-884}f
'2
]
NEUTRON SHIELD , , C TANK COOLER pid"-NSL-154-5 CC-V-885C [(L ) t A P-63 V 4 . l
+,BLOWDOWNROOM C: 2xh 2xe ;
;. r" - - - - - - i
, 4h{ ) PIPE TRENCH t CC-Cv-886 ! 4 y 2" l j WEST OUTSIDE PIPE TRENCH
-!- ~
P.A.B. PIPE TRENCH TO TK-31-1 C L.,,L,, I NEUTRON SHIELD WATER TANK
/
P&ID - 16103-26008 SH. 5 , SYS. - COMPONENT COOLING SYSTEM (NEUTRON SHIELD TANK FILL) NOTE - FOR PENETRATION DETAILS SEE DWG. #16103-59018 SH. 27
& 16103-22027 SH.1&2 I, c, C.A.D. j l MANUAL REVISIONS TO TH]S AS-BUILT g lXXXXXXl - INDICATES VALVE TO BE TESTED m lg DOCUMENT cED 3.01 FOR REVISION AREDURING PROHlBITED. REFER To l
,C_DNyR_UCTIO_N MAN _UAL_RE_VISIO_NS._ _,j Ja
@
- NORTHEAST CONNECTICUT YANKEE UTILITIES ATOMIC POWER CO. SERV
" CONTAINMENT ISGLATION VALVE TEST DIAGRAM i 4- AS BUILT PER PENETRATION
- 63 !
N/A 2 2s5 3 e DCR-CY-S-494-89 AS BUILT - PER k si RJC f an 1-24-86 08a RVK un 3-18-86 AFP. WRK onn3-20-86 M. WRK onn 3-20-e6 8 5-011 7 +g hg
'5R-CY-5-336-88 $e'+t 1 N a-86 V41 NONE
$ w a un am er on. w. e. w as-oit 16103-20310 SH. 61 mOj j
U.S. Nuclear Regulatory Commission B13137/ Attachment 1/Page 23 April 28, 1989 Plant: Haddam Neck Plant
Title:
Exemption from 10CFR50, Appendix J, Section III.C.2 Penetration: P-66 Component Cooling Water to Drain Cooler Description The component cooling water (CCW) system is a closed loop inside and outside containment. However, it is not QA Category I or seismically designed. The pressure in this closed loop is greater than design basis accident (DBA) pressure when the CCW system is operating. The only normally open release point in the system, the CCW surge tank vent, closes automatically when radioactivity is sensed in the vent line. Leakage out of containment through P-66 is not possible with the CCW system intact and functioning (i.e., lines at pressure). CYAPC0 requests a permanent exemption from the requirement of Section III.C.2 of Appendix J which states that local leak rate tests (LLRTs) shall be per-formed with air or nitrogen. CYAPC0 presently tests containment isolation valve (CIV) CC-CV-731 in this component cooling water penetration for the drain cooler with 44 psig water. The water leakage collected is converted to equivalent air leakage at 40 psig. Justification Bases The NR'C does not concur with the correlation of measured water leakage to air leakage, other than permitting such correlation on an interim basis. CYAPC0 acknowledges that the accomplishment of 40 psig air tests would yield an improvement in the accuracy of Appendix J testing for the affected CIV. 1 CYAPC0 questions the cest of the modifications necessary for air testing in dollars and man-rem in Oew of the small improvement in accuracy and public safety achieved by such mod fications. i A probabilistic risk assessment (PRA) of penetration P-66 has been performed. This PRA conservatively determined that the present risk of public exposure due to core melt accident leakage through this penetration is approximately 0.2 man-rem. Modifications to accomplish Appendix J air testing could reduce this estimated public risk to approximately .02 man-rem. This relatively small potential public risk saving must be compared to the man-rem exposure and cost in dollars required to make the plant changes necessary for its accomplishment. It has been estimated that 1.5 man-rem of craft personnel exposure will be accumulated in making the required plant changes. These changes have an estimated cost of $110,000. j I l J
U.S. Nuclear Regulatory Commission I B13137/ Attachment 1/Page 24--
~ April 28, 1989 Conclusion
..CYAPC0 requests a. permanent. exemption from Appendix. J, Section III.C.2' for P-66 on these bases.
1
)
'4 Attachment I page 24a W
" 'A g 6"- AC-152N-29 l "d 6"x3" 7
CC-V-730 D p LJ f" L. F1
.Z CC-V-731 A
[ p_gg Q ' [ M CC-CV-731l g M g V g 3"- AC-152N-93 j E-M g,,- I O y CC-v-732
/
__ j O 6"x3" E DRAIN COOLER P&ID - 16103-26008 SH. 5 E-52-1A SYS. - COMPONENT COOLING c (C.C.W. TO DRAIN COOLER) NOTE - FOR PENETRATION DETAILS SEE DWG. #16103-59018 SH. 27 l
"~~~
l lXXXXXXl - INDICATES VALVE I N C"A"D* l TO BE TESTED
! oTE: MANUAL REVISIONS TO TH)S AS-BUILT l DOCUMENT ARE PR010B]TED. REFER TD g l m g GED 3.01 FOR REVISION DUR]NG ,
s.C_0N_STRUCTIO_N MAN _UAL_RE_V1Sl%_ j da I
@
- NORTHEAST UTILITIES SERV CONNECTICUT YANKEE ATOMIC POWER CO.
CONTAINMENT ISOLATION VALVE TEST DIAGRAM 4- AS BUILT PER PENETRATION # 66 N/A 2 2s-e9 DCR-CY-S-494-89 " h er RJC C*D- RWC w. WRK AS BUILT - PER S +p *g fz orn !-23-86 oan 3-10-86 onn3-20-86 dan 3-20-06
#% WRK 85-011 1 7 es DSR-CY-S-336-86 "e W scu NONE " "S
$ w no. om amns er . cm. w. w. w 85-011 16103-20310 SH. 64
_ _ _ _ _ _ . . _ MRLc0
U.S. Nuclear Regulatory Commission B13137/ Attachment I/Page 25 April 28, 1989 Plant: Haddam Neck Plant
Title:
Exemption from 10CFR50, Appendix J, Section III.C.2 Penetration: P-67 Cot ~::rt Cooling Water from Drain Cooler Description CYAPC0 requests a permanent exemption from the requirement of Section III.C.2 of Appendix J which states that local leak rate tests (LLRTs) shall be per-formed with air or nitrogen. CYAPC0 presently tests containment isolation valve (CIV) CC-FCV-611 in the component cooling water (CCW) penetration for the drain cooler with 44 psig water. The water leakage collected is converted to equivalent air leakage at 40 psig. Justification Bases The CCW system is a closed loop inside and outside containment. However, it is not QA Category I or seismically designed. The pressure in this closed ! loop is greater than design basis accident (DBA) pressure when the_ CCW system is operating. The only normally open release' point in the system, the CCW surge tank' vent, closes automatically when radioactivity is sensed in the vent line. Leakage out of containment though P-67 is not possible with the CCW system intact and functioning (i.e., lines at pressure). The NRC does not concur with the correlation of measured water leakage to air leakage, other than permitting such correlation on an interim basis. CYAPC0 acknowledges that the accomplishment of 40 psig air tests would yield an improvement in the accuracy of Appendix J testing iar the affected CIV. However, CYAPC0 questions the cost of the modifications necessary for air testing in dollars and man-rem in view of the small improvement in- accuracy and public safety achieved by such modifications. A probabilistic risk assessment (PRA) of penetration P-67 has been performed. This PRA conservatively determined that the present risk of public exposure due to core melt accident leakage through this penetration is approximately 0.7 man-rem. This analysis of public risk is dominated by the assumptions of of a relatively high probability of a breach in piping inside and outside containment, and failure of the CIV to close. Since this scenario does not count on the leak tightness of the CIV, there is no decrease in public risk due to modifying the plant to perform an air or nitrogen LLRT. If lines are not assumed to rupture and CC-FCV-611 is assumed to close, modifications to permit air testing would yield a small public risk reduction. This risk reduction would be smaller than the already small public risk of approximately 0.7 man-rem calculated by the PRA. This relatively small potential public risk reduction must be compared to the man-rem exposure and cost in dollars required to make the necessary plant Y___-_-_-________-_.
'g" U.S. Nuclear' Regulatory Commission B131?'/ Attachment I/Page.26 April 28, 1989 change modifications. It has been estimated that 1.0 man-rem of craft person-nel exposure 'will be accumulated in making the required plant changes. The estimated cost of the modifications is $108,000.
Conclusion
- CYAPC0 requests a permanent exemption from Appendix J, Section III.C.2 for P-67 on these bases.
4 ,. 1 I 4
Pap Ma HCP A,B CC-SOV-20T V-611 l S j _p j A/S % ' I c3 , CC-V-738 FE
-J
'f 611 ,Q TO REACTOR lCC-FCV-611 w
,W' q~m " " TIA 6 11 CONTAINMENT SUMP l lll l
BLOWDOWN ROOM
# WEST OUTSIDE ll CC-V-737 PIPE TRENCH g ja
, CC-V-735M
( 2 i E A J ' I
~
CC-RV-134 r CC-V-736 3"-AC-152N-79 J F1 \CC-V-736A 6 ( 3" W ' " 3"- AC-152N-38 I
# 3"- AC-152N-38 P-67
_ FROM DRAIN COOLER E-52-1 A
'\
4 P&ID - 16103-26008 SH. 5 , SYS. - COMPONENT COOLING (C.C.W. FROM DRAIN COOLER) NOTE - FOR PENETRATION DETA]LS SEE DWG. #16103-59018 SH. 27
~
{ NOTE:~ ~ "" ~ C "~A " f" - lXXXXXXl - - -l - INDICATES TO BE TESTED VALVE
! MANUAL REVISIONS TO THIS AS-BUILT 1 lg GED DOCUMENT ARE PROHIB]TED. REFER TO lg 3.01 FOR REVIS10N DURING CONSTRUCTION MANUAL REVISIONS. J n
@ "* CONNECTICUT NORTHEAST UTILITIES SERVIC YANKEE ATOMIC POWER CO.
CONTAINMENT ISOLATION VALVE 4-TEST DIAGRAM N/A 2 2 AS BUILT PER RTW PENETRATION # 67 85 DCR-CY-S-494-89 m RJC C*lk RWC W. WRK W. WRK 85-011 1 7 AS BUIL1 - PER " p g #F un 1-24-86 mt 3-18-86 mt3-20-es mt3-20-86 sc DSR-CY-S-336-8B "e 4 4 # satt NONE DE m
$ w m an amm a n. w. w. w 85-011- 18103-20310 SH. 65 WRC
1 g g . U.S. Nuclear Regulatory Commission B13137/ Attachment I/Page 27 April 28, 1989 Plant: Haddam Neck Plant
Title:
Exemption from 10CFR50, Appendix J, Section III.C.2 Penetration: P-68 Primary Water to Containment Description CYAPC0 requests a permanent exemption from the requirements of 10CFR50, , Appendix J, Section III.C.2, to Type C leak test with air or nitrogen, con- ' tainment isolation check valves PW-CV-139 and PW-CV-140. Penetration P-68 is the primary water supply to containment. CYAPC0. presently Type C leak tests PW-CV-139 and PW-CV-140 with water. Justification Bases The normal pressure within this system is normally maintained in excess of peak loss-of-coolant accident (LOCA) pressure. The system will likely remain water filled during the early stages of a LOCA. Half of the thyroid dose due to containment leakage occurs during the first 15 minutes following the onset of a postulated accident. The primary water supply system is not directly connected to the reactor coolant system. Any radioactive release through this system would have to occur by way of a breach of the piping system both inside and outside of containment. There exists two check valves in series, one inside containment and the other outside containment. Both valves will continue to be Type C leak tested with water. This test will identify any significant leakage through the check valves. The NRC Staff has concluded that the Type C test with water and the use of the water to air conversion will provide reasonable assurance of the leakage integrity of these valves on an interim basis. During a periodic integrated leak rate test (ILRT), check valves PW-CV-139 and 140 are exposed to containment pressure through a spigot connection inside containment in the accident direction and with the back sides of the valves depressurized. The periodic ILRT provides reasonable assurance that leakage through any penetration is limited. The NRC does not concur with the correlation of measured water leakage to air leakage. CYAPC0 acknowledges that the accomplishment of 40 psig air tests would yield an improvement in the accuracy of Appendix J testing for the affected containment isolation valves (CIVs). However, CYAPC0 quest;cris the cost of the modifications necessary for air testing in dollars and man-rem in view of the small improvement in accuracy and public safety achieved by such modifications. l I4
U.S. Nuclear Regulatory Commission B13137/ Attachment I/Page 28 April 28, 1989 The probabilistic risk assessment (PRA) analysis for this penetration calcu-lates the public risk to be approximately 0.5 man-rem in its present configu-ration with its CIVs being leak tested with water. The estimated cost to modify this penetration to comply with Appendix J is
$106,000. . The man-rem expenditure required to complete the modification is 0.5 man-rem.
I Conclusion The present public risk consequences due to Appendix J Type C leak testing penetration P-68 with water in lieu of air or nitrogen is negligible. The small decrease in public risk exposure to modify this penetration does not support the cost in dollars and man-rem required to make the modifications.
.Therefore, CYAPC0 requests a permanent exemption from the requirements of Appendix J, Section III.C.2 for P-68 on these bases.
I j 4
Attachment I page 28a L WEST OUTSIDE PIPE TRENCH
- = a , ,
_1 y p __ gg
=== 2"-ED-152-31
)I lPW-CV-140 l
~
3 s < h
) x:
1" PW-V-141 Rs
~
ga ) PW-CV-139 Q{ PW-V-557 I lPW-CV-600 II [* 4
.> PW-V-146 O
VM ( J PW-V-830
] [ pw.V-145 PW-V-601 o <
. ril LJ r, i
- PW-V-109
. Y-
~~~~~".
P&ID - 16103-26046 SH. 2 SYS. - PRIMARY WATER SUPPLY TO CONTAINMENT NOTE - FOR PENETRATION DETA]LS SEE DWG. #16103-59018 SH. 27 j
& 16103-22027 SH.1&2 lXXXXXXl - INDICATES VALVE I r------------g TO BE TESTED l
l NOTE: MANUAL REVISIONS TO THIS AS-B1DLT l DDCUMENT ARE GED 3.01 FOR REVISION PROH3BITED. REFER TO l DURING l CONSTRUCTION MANUAL REVISIONS. J Jg { l
]
@ '* NORTHEAST UTILITIES CONNECT] CUT YANKEE ATOMIC POWER CO.SERV I
" CONTAINMENT ISOLATION VALVE '
TEST DIAGRAM 4-AS BUILT PER PENETRATION
- 68 N/A 2 zg9- DCR-CY-S-494-89
" b et RJC DE RVK AW. WRK WP. WRK 85-011 7 AS BUILT - PER A A fi un 1-24-86 pait 3-10-86 can 3-20-86 ontt3-20-06 es DSR-CY-S-336-86 $o'F W & sud NONE 1
m aa
$ w m. un amm er a. w. w. w e5-011 16103-20310 SH. GG
."b
.U.S. Nuclear Regulatory Commission B13137/ Attachment 1/Page 29 April 28, 1989 Plant: Haddam Neck Plant
Title:
Exemption from 10CFR50, Appendix J, Section III.C.2 Penetration P-69 Reactor Coolant System Loop Fill Description s .CYAPC0 requests a permanent exemption from the requirement of 10CFR50, Appen-dix J, Section III.C.2, to Type C leak test with air or nitrogen containment
' isolation check valve FH-CV-296. Penetration P-69 is the reactor coolant system loop fill penetration.
CYAPC0 presently Type C leak' tests check valve FH-CV-296 with water as the test medium. Justification Base The loop fill system is connected directly to the charging system. After the termination .of the emergency core cooling system (ECCS) short term recircula-tion, following a large break loss-of-coolant accident (LOCA), either the charging pumps or the residual heat removal pumps will be used for long-term recirculation. If the . charging pumps are used, the loop fill header outside of containment would contain a pressure greater than the containment LOCA pressure for longer than 30 days following a LOCA and leakage of containment-atmosphere'could not occur while the charging pumps were running. A temporary water seal will exist due to this system being filled with water at the start of a core melt accident. Half of the thyroid dose due to con-tainment leakage occurs during the first 15 minutes following the onset of a postulated accident. The NRC does not concur with the correlation of measured water leakage to air leakage. CYAPC0 acknowledges that the accomplishment of 40 psig air tests would yield an improvement in the accuracy of Appendix J testing for the affected containment isolation valve (CIV). However, CYAPC0 questions the cost of the modifications necessary for air testing, in dollars and man-rem, in view of the small improvement in accuracy and public safety achieved by such modifications. During a periodic integrated leak rate test (ILRT), check valve FH-CV-296 is exposed to containment pressure in the accident direction with the back side of the valve depressurized. The periodic ILRT provides reasonable assurance that leakage through any penetration is limited. System leakage to the environment is inspected for as part of the technical specification limit of three liters per hour maximum leakage (10CFR100 offsite doselimit). 4
o. U.S. Nuclear Regulatory Commission B13137/ Attachment I/Page 30 April 28, 1989 FH-CV-296 will continue to be Type C leak tested with water. The water leak test will assure that significant leakage does not occur. The Staff has concluded that the Type C test with water and the use of the water to air conversion will provide reasonable assurance of the leakage integrity of this valve on an interim basis. The probabilistic risk assessment (PRA) analysis for this penetration conser-vatively calculates the public risk to be approximately 3 man-rem in its > present configuration with the CIVs being leak tested wifh water. The estimated cost to modify this penetration to comply with Appendix J is
$115,000. The man-rem expenditure required to complete the modification is approximately 4.0 man-rem.
Conclusion The present public risk consequence to continue Appendix J type leak testing P-69 with water in lieu of air or nitrogen is small. The small decrease in public risk exposure does not support the cost in dollars or man-rem required to make the modifications. Therefore, CYAPC0 requests a permanent exemption from Appendix J, Section III.C.2 for P-69.
l Attachment I page 30a ! o. 50 d ____ d $W g $e (-$ g > :i A e ___ e y- m r 7 e E l[ : E* g st a R a i a e a e a @ $ g*g 8 , A R$ R$ R$ E R$ RE$ .
!g $,s_Ek $
s c -s g gy s - s c ses g-y o^ g km :e j? kats)kmti $m:gi uf l l LE 55 t lji-
.& 6 i 4
l dj
%n 66%nQ\ i x n
e e n = -
%a % f E
a i s Y C g ____ ; s W ~@' I Y
*-'.1 6
i t -
- i-sel JJ cn o 2ij d $$ g e 6 l gl[gje d_ 6 k / dg
!d !?I? *e 7 $$ !4 , ,,!l l! *$d dg gg i pM m= i n wa is u - g
_.._.._.._.._.._.._: . I.___p,5lgl
, 3 g 3 M i A I $
'v' '
g . -s u b %
- n g zi k ha e t % :kl sg E! $I
- d g5
> N s ig a m:
c 4 a E N k 0 0
*gE
.N
- a E R g b 8 s
~ 6 g a.
( -
' e$ 8=8 5"
N 28 Eb $ in a sr lX ge e en 3
~
N]_ s _ s v 'l X 3 6 6 Xgas X a f. $ $d MW
-= @E er 8
8 5 m E g5 n gw go g- A - ( e us a' C 5 -A!3 _E 8 8 7 f Ian Idi Ink ia i i d h \'[ __Yd} [: \ - c
e 1 t 4, , U.S. Nuclear Regulatory Commission 813137/ Attachment I/Page 31
' April.28, 1989 Plant: Haddam' Neck Plant
Title:
Exemption from 10CFR50, Appendix J, Section III.C.1 Penetration: P-78 Relief Tank Drain Description CYAPC0 requests a permanent exemption from compliance with Section III.C.1 of Appendix J, in the pressurizer relief tank drain penetration P-78. Specifi-cally, CYAPC0 requests an exemption from the requirement that reverse direc-tion Type C testing provide " equivalent or more conservative results" than direct pressure testing. Containment Isolation Valve (CIV) DH-TV-554 in P-78 is presently reverse direction local leak rated tested (LLRT) with nitrogen at 42-47 psig. Justification Basel CIV DH-TV-554 is a diaphragm actuated, single seated, globe valve with socket weld ends. Its direction of installation permits inclusion of the valve's body-to-bonnet joint and stem packing in the reverse direction LLRT. The reverse direction LLRT tends to-force the valve's disc into its seat. Core melt accident pressure tends to force the valve's disc out of its seat. 1he containment side of this valve is continuously exposed to pressurizer relief tank (PRT) N2 pressure (3 psig). Leakage through the CIV would depres-surize the: PRT and trigger an alarm. Therefore, the leak tightness of the containment side of DH-TV-554 is demonstrated by retaining PRT pressure on a - continuous basis. Since the PRT is vented to containment during a periodic. integrated leak rate test (ILRT), the inboard side of this valve is also leakage checked during a periodic ILRT. This is a direct pressure test of the subject CIV. Under some beyond design basis accident conditions, the PRT pressure could increase above the predicted maximum containment pressure. This is because the PRT rupture disk will not fail until the PRT pressure exceeds containment pressure by 75 psi. However, events that could be radiologically significant and result in pressurizing the PRT (e.g. stuck open PORV or code safety valve) would not result in failed fuel until long after the rupture disk had failed. For these events, Type A testing will bound the pressure seen by the valve when radio-logical conditions inside containment warrant low leakage. CYAPC0 acknowledges that modification of the plant to accomplish a direct pressure LLRT in P-78 would yield a more accurate assessment of the subject CIV's performance during a core melt accident. However, CYAPC0 questions the cost of the slight improvement in accuracy anticipated in terms of dollars and craft personnel exposure required to make the associated plant modifications.
e
*.. j U.S. Nuclear Regulatory Commission B13137/ Attachment I/Page 32 April 28, 1989 A probabilistic risk assessment (PRA) of penetration P-78 has been performed.
It determined that the present risk of public exposure due to core melt accident leakage is now approximately .04 man-rem. . Modifications to accom-plish Appendix J direct pressure testing could reduce this estimated public risk to approximately .004 man-rem. This relatively small potential public exposure saving must be compared to the dosage and cost required to make the plant changes necessary for its accom-plishment. It has been estimated that 6.0 man-rem of craft personnel exposure and $95,000 will be required to make the required plant modifications. Conclusion CYAPC0 does not believe that the slight decrease in public exposure risk justifies modifying P-78 to comply with Appendix J. CYAPC0 also proposes to perform a reverse direction LLRT in P-78, the continu-ous retention of PRT pressure by DH-TV-554 and its periodic exposure to direct pressurization during a periodic ILRT as the best means of assuring the valves leak tightness with the existing plant configuration. I l 1
,h _ Attachment I page 32 gl HCP A,B N MCB(
- I TO PRIMARY l C -4 >4--HCP A,8 [ PRESSURIZER h DRAIN TANK I I RELIEF TANK 'l OS S P-78( TK-B /
q A/S b q 2"-DRL-154 -6 DH-SOV-1844 DH-SOV-554 D y ' g 2'-DRL-154-12 I
= FC FC
. 2> l y,
, M ,
{
"*""" J lDH-TV-1844l lDH-TV-554l i 'j O (LC)C [DH-V-563 DH-V-317 .
F ) (LC )
'J f % 2"-DRL-152-6 '
L_J P A B. l D-PIPE TRENCH _l_ EAST OUTSIDE REACTOR h UNDER VALVE T PIPE TRENCH CONTAINMENT ROOM l SUMP HCP A,B --e% i
}- . <C DH-SOV-1843 E
hl
, l
, A M A/Sl
& o 4
> FC l 1 ,4 y 2 '
l ( FROM REACTOR COOLANT 1 PUMP HEAD TANK DH-TV-1843 1)"-CRL-152-16 DH-V-315 P&ID - 16103-26031 SH. 2 (OUTSIDE) (LC ) 16103-26007 SH. 3 (INSIDE) LJ D. SYS. - REACTOR COOLANT (PRESSURIZER RELIEF TANK DRAIN)
- -~ ~~ Q~-g -~ ~~ ~ ~ ~~i NOTE - FOR PENETRATION DETAILS SEE DWG #16103-59018 SH. 27
["NuAt I En"IjE Revisions To Tms As-su1LT l I RE R N R g lXXXXXXl - INDICATES VALVE ' A l I
, CONSTRUCTION MANUAL REVIS10NS. ]
TO BE TESTED g,Q NORTHEAST UTILITIES SERVICE CO.
@ HR CONNECTICUT YANKEE ATOMIC POWER CO. l
" CONTAINMENT ISOLATION VALVE m TEST DIAGRAM 4- AS BUILT PER g PENETRATION # 78 N/A 2 2$ DCR-CY-S-494-89 " P DE. W. WRK M.
si RJC RGW WRK 85-011 1 3 7 AS BUILT - PER 8
'+g +g
- f aan 1-16-86 aan 3-18-86 can 3-20-86 Dan 3-20-06 es DSR-CY-5-336-86 "o m NONE N a-a w m au avam er os. e. e. w 85-011 16103-20310 SH. 78 [
W
I t c. U.S. Nuclear Regulatory Commission B13137/ Attachment I/Page 33 i April 28, 1989 Plant: Haddam Neck Plant
Title:
Exemption from 10CFR50, Appendix J, Section III.C.1 Penetration: P-80 Auxiliary Spray from Fire System Descripti_o_n CYAPC0 requests a permanent exemption from compliance with Section III.C.1 of 10CFR50 Appendix J in this fire water to containment spray penetration. Specifically, CYAPC0 requests an exemption from the requirement that reverse direction Type C testing provide " equivalent or more conservative results" than direct pressure testing. The containment isolation valve (CIV) in this penetration is RH-MOV-31. This valve is presently local leak rate tested (LLRT) with reverse direction water pressure to 80 psig (fi system minimum pressure). By letter dated September 29, 1987,grotectionthe Staff granted a permanent exemption for this penetration. This exemption evaluated reverse direction testing, with water rather than air at a pressure 7. eater than DBA pressure. This exemption did not consider the possibility tut the stem packing and body-to-bonnet joint on RH-MOV-31 may be omitted from a reverse direction LLRT. The necessity of addressing these potential containment isolation boundaries requires revisita-tion of the exemption request for P-80. Justification Bases The exemption request for RH-MOV-31 was granted based on the following:(12)
"The auxiliary spray (P-80) penetration is isolated by motor-operated valve RH-MOV-31 which is tested in the reverse direction.
Piping associated with this penetration is pressurized by the fire water system at a minimum pressure of 80 psig. The fire water system uses the Connecticut River as its source of water, and the system pumps have a capacity of 2900 gpm at 100 psig. The system is comprised of an electric driven pump with a backup diesel driven i pump; the status of the pumps is provided in the control room, and ! the pumps may be operated from the control room. The fire water system is maintained in accordance with the quality assurance (11) D. M. Crutchfield letter to E. J. Mroczka, " Exemption from Certain Requirements of 10CFR Part 50 Appendix J," dated September 29, 1987. (12) See the J. F. Opeka letter to C. I. Grimes, " Request for Exemption," dated March 12, 1986, at page 1-8. l
t c, ' U.S. Nuclear Regulatory Commission l B13137/ Attachment I/Page 34 April 28, 1989 requirements of the fire protection program. System pressure is normally controlled at 100 to 110 psig by a hydropneumatic pressure maintenance system. While the yard piping loop is a single line, sectionalizing valves are provided to permit partial line isolation without interruption of service to the entire system. Thus, a pressure at least double accident pressure (Pa) is reliably maintained in this piping, which would assure that the valves in this penetration are not exposed to containment atmosphere leakage. Although this penetration is not sealed by an installed seal system, the penetration is effectively sealed by the system itself, and water testing of this penetration at a pressure not less than 1.1 Pa is appropriate. Further, reverse direction testing of this penetration is appropriate since post-accident leakage would tend to occur into the containment. On this basis, CYAPC0 believes that a permanent exemption from the requirements of Appendix J to permit reverse direction testing of Penetration P-80 is justified; based upon demonstration of an effective 30-day water seal, no exemption is required to test this
>'netration with water."
The subject valve is a solid wedge gate valve with butt weld ends. Since this valve has inboard and outboard disc / seat interfaces, it is possible that its stem packing and gasketed body-to-bonnet joint may never be pressurized by an LLRT, regardless of the direction of pressurization, or a core melt accident. This is because the leak tightness of both disc / seat interfaces would prevent pressurization of the valve's bonnet. If the inboard disc / seat interface is not leak tight, the CIV's stem packing and body-to-bonnet joint are potential containment isolation boundaries that may be omitted from a reverse direction LLRT. These potential containment boundaries should be included in an inte-grated leak rate test (ILRT). The periodic ILRT procedure blanks and vents the outboard side of RH-MOV-31 to establish proper differential pressure across penetration P-80. However, the inboard side of the CIV may not be exposed to periodic ILRT pressurization if check valve RH-CV-35 inside containment is leak tight. CYAPC0 proposes to amend its periodic ILRT procedures to open a capped line inside containment between RH-M0V-31 AND RH-CV-35 to assure direct pressurization of the subject CIV in future periodic ILRTs. This point was not addressed in the previous exemption request for P-80. CYAPC0 contends that the new issues discussed above do not diminish the bases for exemption previously submitted to the NRC. In addition to these justifi-cations for exemption, CYAPC0 contends that the present Appendix J testing program for P-80 is defensible based on an assessment of probabilistic risk assessment. i A probabilistic risk assessment (PRA) of P-80 has been performed. It con- j cluded that the present risk of public exposure due to design basis accident l
t U.S. Nuclear Regulatory Commission B13137/ Attachment I/Page 35 April 28, 1989 (DBA) leakage through this penetration is approximately .0005 man-rem. If the ' LLRT for P-80 was in compliance with Appendix J, this potential public expo-sure could be reduced by a factor of approximately ten (i.e. to .00005 man-rem). This small potential public exposure saving must be compared to the dosage and cost required to make the plant changes necessary for compliance to Appen-dix J. It has been estimated that 2.5 man-rem of craft personnel exposure and $118,000 will be required to make the modifications necessary for Appen-dix J compliance. Conclusion CYAPC0 does not believe that the slight decrease in potential public exposure by modifying P-80 for a compliance with Appendix J is justifiable. Therefore, CYAPC0 requests a permanent exemption from Appendix J, Section III.C.1 for P-80. 1
IC r V c N R E S ma S E Et L TI a IL a r TI v U r T l
- T S
uN c uE A c aN
)
S = EwI T R o
*t c T 4
3 7 3 3 P 4 5
= O N a N se 0
5 5 0 5 5 W 5 ( P 0 5 5
= -
1 1 1 T 1 - S
=
- - - T -
H H G H
- H '
R R R 3 R
- N -
s s R I P 6 Y
= l
=
-- A R =
P S g T N E
]mt
~ ta o
- t sr V
N a a [sm YM LE PT I A ~ s . e o (t t v - T PS UY m' N O 3. om te ri1 m SS YY CA C wU A. ysehst C iest"= vs s i NR EP l V a uv [n 9 t CS 1 mae R % 2 mnd(I
- ET 0 6 , n MN EE 6
0 5
' 6 0
- U A ,
3 H 9 2 l3 uo M 5 5 5 3 R 0 9N 3 5 ! 2 3 I 1 H
- 0 v 1
81
- V-8 L 1A 8 M 2 R 1 R NW C 5T c R 0 MHR O 0 NI 1
1N ' 8
- H u *8 -
3 - V H O t T P P H - N RC
- R H R "8 T O
l 6 8 y e[R ] r T OMN CI
- 9 1
- 3 1
[ 5 3 AA E RN YS [ O E
- 5 1
l
] I-I R1 U32 4 3
3 1 C g 0 T M*- 6 6 - - -
'" v R O
N R
"8 3
V
- C A
yr iu0V o 0 u 6 1 4 H H C
'. '8 R R A I 4Il '[a -
- _- ' [
1 l
"8
_- 3 I [' N" R 6 H
- 8 1
5
- Y A M
)
[ 1 RE T
,,( H PS yr ) R SYS 7 @ T 3
1 E NN V 8' 1' D H I S C
~
EO VI H T N R J E R ]NT C D T 18 A E T E 1 NTT O S A P & IONO T OR PI A g;( g i-I ;* I E 1 8 C CP
,I ,l -
5- 2 E D N 0 T YE T RR U TE 6 O AI
'" O R
G A KS 2- R lF Pl R 3
' R
- [ X -
E D N LE AG UN DA 01 EXIM 6 IRUO AR U uIH 1 F(F 9 1 D 2 0 0SC rex FRE - - 1
- N 2 D .
V NL O - I S O I V P R T - & Y F G Cu P P S R Et F Yr EE DROS NIRT UFPS i Q:
;ul 7C O i}}