ML17292A746
| ML17292A746 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 03/20/1997 |
| From: | Gwynn T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | Parrish J WASHINGTON PUBLIC POWER SUPPLY SYSTEM |
| References | |
| NUDOCS 9703260203 | |
| Download: ML17292A746 (23) | |
Text
CATEGORY REGULAT~ INFORMATION DISTRIBUTIO YSTEM (RIDS)
ACCESSION NBR:9703260203 DOC.DATE: 97/03/20 NOTARIZED: NO DOCKET FACIL:50-397 WPPSS Nuclear Project, Unit 2, Washington Public Powe 05000397 AUTH.NAME AUTHOR AFFILIATION GWYNN,T.P.
Region 4 (Post 820201)
RECIP.NAME RECIPIENT AFFILIATION PARRISH,J.V.
Washington Public Power Supply System
SUBJECT:
Documents 970320 discussion w/util re performance of response time measurements at facility 6 provides overview of NRC conclusion.
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UNITEO STATES NUCLEAR REGULATORY COMMISSION REGION IV 611 RYAN PLAZA ORIVE, SUITE 400 ARLINGTON, TEXAS 76011-8064 eR 20 $9t Mr. J. V. Parrish (Mail Drop 1023)
Chief Executive Officer Washington Public Power Supply System P.O. Box 968 Richland, Washington 99352-0968
SUBJECT:
RESPONSE TIME MEASUREMENT IN ACCORDANCE WITH TECHNICAL SPECIFICATION SURVEILLANCE REQUIREMENTS
Dear Mr. Parrish:
The following letter documents the discussion between you and your staff and the NRC on March 20, 1997, regarding the performance of response time measurements at Washington Public Power Supply System Nuclear Project 2 (WNP-2).
The NRC has reviewed your staff's implementation of Boiling Water Reactors Owner's Group Licensing Topical Report NEDO-32291, "System Analyses for Elimination of Selected
Response
Time Testing Requirements,"
and has concluded that for certain instruments WNP-2 Technical Specification Surveillance Requirements to demonstrate that response times are within specified limits are not being met.
The NRC's review includes onsite inspection activities from September 24-26, 1996, meetings between your staff and NRC personnel on September 16 and October 4, 1996, and continuing in-office review.
This letter provides an overview of the NRC's conclusion.
In January 1994, the Boiling Water Reactors Owner's Group submitted Licensing Topical Report NEDO-32291 for NRC review and approval.
On December 28, 1994, the NRC issued a safety evaluation approving the implementation of NEDO-32291.
The NRC concluded that selected Technical Specification instrument response time tests could be eliminated from Technical Specifications and that licensees must confirm the applicability of NEDO-32291 to their plant when submitting license amendment applications for the elimination of the selected response time tests.
In a separate action, on June 26, 1995, the NRC issued Amendment 139 to Facility Operating License NPF-21.
The amendment modified the Technical Specifications to relocate instrument response time limittables for the reactor protection system, isolation actuation, and emergency core cooling system.
The affected instrument response time limit tables were to be relocated to the Final Safety Analysis Report.
The NRC's safety evaluation for Amendment 139 stated that "the changes do not affect the TS action statements for inoperable instrumentation, nor do they affect the surveillance requirements to perform response time testing."
On April 3, 1996, your staff approved SAR/Technical Specification Basis Change 96-017 to relocate the instrument response time limittables from the Final Safety Analysis Report to the WNP-2 Licensee Controlled Specifications.
Change Number 96-017 also changed 9703260203 970320 PDR ADQCK 05000397 P
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Washington. Public Power Supply System the instrument response time limit tables to indicate that certain sensors and actuation instrumentation were eliminated from response time testing and to indicate that certain response time limits were "NA" instead of discreet values.
Your actions to implement SAR/Technical Specification Basis Change 96-017 were not consistent with the NRC's safety evaluation for Amendment 139 to Facility Operating License NPF-21, which indicated that surveillance requirements to perform response time testing were not affected.
Instrument testing was performed. prior to the June 1996 completion of WNP-2 Refueling Outage R11.
During testing, your staff used the methodology described by NEDO-32291 to determine whether instrument degradation had occurred.
The methodology qualitatively verified that sensors and other portions of trip functional instrumentation responded to an input.
WNP-2 Technical Specifications 1.36, 1.19, and 1.12, (WNP-2 License Amendment 139, issued June 26, 1995) respectively, defined REACTOR PROTECTION SYSTEM RESPONSE TIME, ISOLATION SYSTEM RESPONSE TIME, -and EMERGENCY CORE COOLING
.SYSTEM (ECCS) RESPONSE TIME to be the time interval from when the monitored parameter exceeds its actuation setpoint at the channel sensor until appropriate equipment actuations are completed (e.g., deenergization of scram pilot valve solenoids, isolation valve travel to required positions, or required ECCS equipment response such as pump discharge pressures reaching required values).
The response time may be measured by any series of sequential, overlapping, or total steps such that the entire response time is measured.
WNP-2 Technical Specifications 4.3.1.3, 4.3.2.3, and 4.3.3.3 (WNP-2 License Amendment 139, issued June 26, 1995) required that the REACTOR PROTECTION SYSTEM RESPONSE TIME, ISOLATION SYSTEM RESPONSE TIME, and ECCS RESPONSE TIME, respectively, of each reactor trip, isolation, and ECCS trip function be demonstrated to be within its limit at least once per 18 months.
We conclude that during Refueling Outage R11 your staff did not demonstrate in accordance with Technical Specification requirements, as amended, that response time limits were met for all required functions in that sensor response was not included in the response time measurement for all required functions.
For other functions, response time limits were not demonstrated, as portions of the trip functional instrumentation were qualitatively verified to respond "to an input. Therefore, we conclude that response times had not been demonstrated to be within specified limits at least once per 18 months for all required functions.
We conclude that the following REACTOR PROTECTION SYSTEM trip functional units were not demonstrated to meet response time limits: Reactor Vessel Steam Dome Pressure-High and Reactor Vessel Water Level - Low, Level 3 (see Table 1). Although your staff
I" /
Washington Public Power Supply System measured the response time of these reactor protection system trip functional units except for sensors, your staff only qualitatively confirmed sensor response to an input, which was not the measurement required by the Technical Specifications.
We conclude that the following ISOLATION SYSTEM trip functions were not demonstrated to meet response time limits: Primary Containment Isolation Reactor Vessel Water Level-Low Low, Level 2, Drywell Pressure
- High, Main Steam Line Pressure
- Low, and Main Steam Line Flow - High; Secondary Containment System Isolation Reactor Building Vent Exhaust Plenum Radiation - High, Drywell Pressure
- High, and Reactor Vessel Water Level
- Low Low, Level 2; Reactor Water Cleanup System Isolation ~ Flow - High and Reactor Vessel Water Level -, Low Low, Level 2; Reactor Core Isolation Cooling System Isolation RCIC Steam Line Flow - High, RHR/RCIC Steam Line Flow - High, and RCIC Steam Supply Pressure
- Low; and RHR System Shutdown Cooling Mode Isolation Reactor Vessel Water Level - Low, Level 3 (see Table 2). Although your staff measured response time of the main steam isolation valve actuation function for the Primary Containment Isolation Reactor Vessel Water Level - Low Low, Level 2, Main Steam Line Pressure
- Low, and Main Steam Line Flow - High, trip functions except for sensors, your staff only qualitatively confirmed sensor response to an input, which was not the measurement required by the Technical Specifications.
For the remainder of.the isolation actuation system, including the non-main steam isolation valve actuation functions for the above three trip functions, "NA" was substituted for the response time limit in Table 2.
For these trip functions with "NA" as the response time limit, your staff only qualitatively confirmed instrumentation response to an input, which was not the measurement required by Technical Specifications.
We conclude that the following EMERGENCY CORE COOLING SYSTEM trip functions were not demonstrated to meet response time limits: Low Pressure Core Spray System; Low Pressure Coolant Injection Mode of RHR System - Pumps A and B; Low Pressure Coolant Injection Mode of RHR System - Pump C; and High Pressure Core Spray System (see Table 3). Although your staff measured the response time of the ECCS trip functions except for actuation instrumentation, your staff only qualitatively confirmed actuation instrumentation response to an input, which was not the measurement required by Technical Specifications.
While we conclude that your technical approach to verification of instrument operability is generally consistent with an approach that we have found acceptable, we nevertheless find that your approach to implementation is not in accordance with your Technical Specifications.
Washington Public Power Supply System Based on your failure to demonstrate that response time limits are met ig accordance with the Technical Specifications as discussed above, please inform us immediately as to what actions you plan to take so that we may consider your plans in determining the appropriate enforcement action.
Sincerely, Thomas P. Gwynn, D ecto Division of Reactor roje s
Docket No.: 50-397 License No.:
Enclosure:
Tables 1, 2, and 3 CC:
Frederick S. Adair, Chairman Energy Facility Site Evaluation Council P.O. Box 43172 Olympia, Washington 98504-3172 Mr. Paul R. Bemis (Mail Drop PE20)
Vice President, Nuclear Operations Washington Public Power Supply System P.O. Box 968 Richland, Washington 99352-0968 Mr. Rodney L. Webring (Mail Drop PE08)
Vice President, Operations Support/PIO Washington Public Power Supply System P.O. Box 968 Richland, Washington 99352-0968 Mr. Greg O. Smith (Mail Drop 927M)
WNP-2 Plant General Manager Washington Public Power Supply System P.O. Box 968 Richland, Washington 99352-0968
I
Washington Public Power Supply System Mr. David A. Swank (Mail Drop PE20)
Manager, Regulatory Affairs Washington Public Power Supply System P.O. Box 968 Richland, Washington 99352-0968 Mr. Al E. Mouncer (Mail Drop 396)
Chief Counsel Washington Public Power Supply System P.O. Box 968 Richland, Washington 99352-0968 Ms. Lourdes C. Fernandez (Mail Drop PE20)
Manager, Licensing Washington Public Power Supply System P.O. Box 968 Richland, Washington 99352-0968 Mr. Malcolm H. Phillips, Jr., Esq.
Winston & Strawn 1400 L Street, N.W.
Washington, D.C. 20005-3502
Washington Public Power Supply System NR 20 l997 bcc to DCD (IE01) bcc distrib. by RIV:
Regional Administrator DRP Director Branch Chief (DRP/E, WCFO)
Senior Project Inspector (DRP/E, WCFO)
Branch Chief (DRP/TSS)
Leah Tremper (OC/LFDCB, MS: TWFN 9E10)
J. Lieberman (OE, MS: 7-H5)
OE:EA File (MS: 7-H5)
G. F. Sanborn, EO W, L. Brown, RC Resident Inspector DRS-PSB MIS System RIV File M. Hammond (PAO, WCFO)
WCFO File DOCUMENT NAME:
To receive copy of document, indicate ln box: "C" = Copy without enclosures "E" = Copy with enclosures "N" = No copy RIV:D:WCFO KEPerkins" 3/20/97 NRR/HICB JWermiel" 3/20/97 OE JLieberman" 3/20/97 OGC RHoefling" 3/20/97 ADPR RZimmerman" 3/20/97 (T-KEP)
RIV:D:DRP TPGwynn 3/ f97 "previously concurred OFFICIAL RECORD COPY
Washington Public Power Supply System NR 20 1997 bcc to DCD (IE01) bcc distrib. by RIV:
Regional Administrator DRP Director Branch Chief (DRP/E, WCFO)
Senior Project Inspector (DRP/E, WCFO)
Branch Chief (DRP/TSS)
Leah Tremper (OC/LFDCB, MS: TWFN 9E10)
J. Lieberman (OE, MS: 7-H5)
OE:EA File (MS: 7-H5)
G. F. Sanborn, EO W. L. Brown, RC Resident Inspector DRS-PSB MIS System RIV File M. Hammond (PAO, WCFO)
WCFO File DOCUMENT NAME'o receive copy of document, indicate ln box: "C" ~ Copy without enciosures "E" ~ Copy with endosures "N". ~ No copy RIV:D:WCFO NRR/HICB OE OGC ADPR KEPerkins" 3/20/97 JWermiel" 3/20/97 3/20/97 3/20/97 JLieberman" RHoef ling" RZimmerman" 3/20/97 (T-KEP)
RIV:D:DRP TPGwynn 3/ f97 "prewously concurred OFFICIAL RECORD COPY
1
gag gt3 OVGO
- .M-20-S1 THU 14
- 02 REG10 NGFO DiLKM FAX NO.
510 0356 P. 01 RPS Ifistl'ume(itag)fffl 1.3.1.>
TABLE 1 3.1.1-1 (Page 1 of I)
REACTOR PROTECTION SYSTEM RESPONSE TIME
~ ~
~ ~ ~
intermediate Range Honftors:
Heutron Flux - High b.
(noparatfve
" Average Peat Range Honftor(a~f a.
Neutron F1ux - Upscale.
Setdown b.
Floe fffased Sic<lated Thotsnal Pceer - Vpsoale c.
Fixed ffeutron Flux -'pscale d.
lnopcrat fve Reactor Vessel Stctun Gcee Pressure High Reactor Vessel Vater Level Low. Lovel 3 Hain Steant Lfne isolation Valve - Closure GKLETED Prfaery Containment Pressure - High Scram Ofscharge Volume Vater Level - High Level Transmttter h.
Float Switch Turbine Throttle Valve - Closure Turbine Governor Valve Fast Closure, Trip Of'1 Pressure - Lot Reactor Hodo Switch Shutdown Position Henual Scram RKSPOHSE TIHE (Seconds)
HA 6 l(>>
s 0.09 HA
. s O.SS<<l s l 06<c~
s 0.06 s 4.08 s 0.0S(@
(al
. ffeutron'detectors are exenot frees response time testing.
Response
tfnst shall be measured frets the detector output or frtxn the input of the first e1cctronfc conponent in the <<hannel.
(ft) lnoludfng simulated thermal power tine constant.
(e)
Sensor is eliminated from response the testing for these RPS circufts.
Response
time testing and conformance ta the adntnfstrative limits for the rcnetntng channel relay logfc are required.
(d)
Heanured frere start of turbine control valve fast closure.
- T% NC CONCLUDES THAT THESE TRIP, FUNCTIONAL UNITS MERE NOT DEMONSTRATED TO MEET NKSKNSK TINE LIMITS.
1.3.l.i-'I Revisioo 0
cacti cv oi shu at
~ vc.
I s VC rnh <<v. oiu 0 uooo ps@~on Actuation Instrumentation 1.'3.R. I TABLE 1.3.2.I-I (Page 1 of 2)
ISOLATION SYSTEH INSTRUHEHTATIOH RESPONSE TINE
~
~ ~
tWCTIOH L.
Primary Contaiceent !solstice a.
Reactor Vessel Vatcr Level Ij Lost, Level 3 y
2)
Lou Lpe, Level 2 b.'rwell Pressure - High c.
4.in Stean Line I) aa.am 2)
Pressure Less 3)
Flou>> High HA g I.o(")IHA-HA s I.O('4HA-4 5(b)/HA s d.
i%in Steam Line Tunnel Teeperature High Hain Stela Line Tunnel a Temperature - High Condenser Vacuue - Let q.
Hanua 1 Initiation Sedondary Contaiaaent Systte Iso)ation a.
Reactor Building Vent Exttsust planus Radiation - High b.
4ryue'll Pressure - High c,
Reactor Vessel Mater Level - Loa Lou. Level 2 d.
Hanua'I initiation Reactor Vater C)eanup Systea Isolation Flee - High b.
Heat Exchanger Area Tesperature High e.
Heat exchanger Area Ventilation a Ttep. - High Puep Area Tceperature - High lych RA(cj-lih Nh Nh (a)
(cont inued)
Isolation systse instnwentatfon response tilde specified for the Trip Function aotuating each valve group shall be added to isolation t>ec for valves in carh valve group to obtain I50LATIgH SVSKH RES7085$
TIHE for tach valve, Isolation systce lnsttaaentation response time for HSIVs on1y.
Ro diesel generator delays assumed.
Sensor is eliminated free response t>se testing for the Hsfv actuation logic circuits.
,Response ttee testing and conformance to the adelniatrative
)iantS for the reeaining Channel including trip un~t and relay logic are required.
Radiation detectors are erenpt free response tice testing.
Response
time shall be eeasured free the detector output or free the
>nput of tho first electronic component in the <<nannel.
- T% )NG CONCLUDES THAT THESE TRIP FUNCTIONS MFRE NOT DEt<ONSTRATEO TO NEET RfStONSK TIbtE LIMITS.
W
)I
AtfK CU 5(
InU ld ~ J(
SHREW l(Culuti IV WL'l U P. 04
)OX NU.
>10 760366 (td51%)jan Actuatjon Instr umentat~on 1.3.2.1
~
~ ~ ~
TABLE l.3.Z.I->
(Page 2 of 2)
IMLATIOH SYSTEH INSTRUHENTATION RESPONSE TIHK 5.
IdH Reactai. Vater Cleanup System Iso lation (continued),
e.
Pump Area Ventilation a Tenp.
High SLCS Initiation g.
Reactor Vessel Vater Level -,l.ow Law. Level 8 h.
RMCUCRCIC Line Routing Area Temperature - High i.
RMCQ Line Routing Area Temperature High Hanual Initiatian
'k.
8)oe<<tnt Flee - High Reactor Core Isolatian Cooling System Isolat!on a.
RCIC Steea l.ine Flaw High b.
RHR/RCIC Stean Line Flow - High c.
RCIC Steam Supply Pressure Law d.
RCIC Turbine exhaust diaphragm Pressure - High e.
RCIC Equipaent Ralso Temperature - liigh RCIC Kaaipnent Roam 4 Tcayerature High g.
RMCV/RCIC Steam Line Rauting Area Temperature " High h.
drywall Pressure - High Hanual Initiation
$% Syataa Shutdawn Caaling HOde Italatian a.
Reactor Uessel Vater Level - Law, l.evel 3 b.
Reactor gassed (RHR Cut in Pensissive)
Pressure " High c.
Equipment brea Temperature - High Equipment Area Ventilation a Temp. - High e.
Shutdcwn Ccaling Return Flow Rate - High RHR Heat Exchanger Area Tavierature - High g.
Hanua I tnitiat ion (Seconds) (af NA s Z.s(")
HA-RA HA ~
(a) isolation system instrument>t ion respanse time spccii'led far the Trip Function actuating each valv group shall be added to isolatsan time for valves in each valve group to obtain ISdLATIdH STSTBi RKSPdnSK TIHE for each valve.
(d) includes process response.
instr~ntatian.
logic and time delay ccnmonents.
e THE NK CONCLUDES THAT THESE TRIP FUNCTIONS MERE NOT OPAONSTRATED TO MEET kESKNSK TIME LIMITS.
0, h
P
l~ CV 0 I l(lV 1<'U ht.Mavis a v vtVi u rnn nv. Jtu
@to uooo r ~ UQ
- ~3 ECCS Instrumentation 1.3.5.1 TABLE 1.3.5.1-1 (Page 1 of 1)
EMERGENCY CORK COOLING SYSTEH RESPONSE TINE
~ 4 ~
Levv >reesore Core Spray System
~ pressoro Coolant ln5eotion Node Of RHR Systenl
~.
heaps A and B
Foep C
- l. ~~tin Oepressor<zatlon System Nyk Preesore Core Spray Systan L
Lees Of Peer'ESPONSE TNE (Seconds) s 43(a) s 4S<>>
s alta)
(e)
RC5 aotoatton
)nstromentatton ks eliminated froa response the testing.
+ THE NC CONCLUDES THAT THESE TRIP FUNCTIONS MERE NOT DEMONSTRATED TO MEET
%SPONSOR TINE LIMITS.
L Revision 0