ML20137K702
| ML20137K702 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 11/27/1985 |
| From: | Koester G KANSAS GAS & ELECTRIC CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| TASK-2.K.3.05, TASK-TM GL-85-12, KMLNRC-85-261, NUDOCS 8512030265 | |
| Download: ML20137K702 (10) | |
Text
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KANSA3 GAS AND ELECTRIC COMPANY THE! ELECTE COMPANY l
@LENN L NOESTER vect p.essorse?. asuckta.
Novenber 27, 1985 Mr. Harold R. Denton, Director Office of NLiclear Reactor Regulation U.S. Nuclear Regulatory Comission Washington, D.C.
20555 KMLNRC 85-261 RE:
Docket No. SIN 50-482 SUBJ: Response to Generic Ietter 85-12
Dear Mr. Denton:
The subject Generic Ietter provided confirmation of the acceptability of information provided by the Westinghouse Owners Group (h0G) with regard to 'IMI Action Item II.K.3.5, "Autonatic Trip of Reactor Coolant Punps".
Additionally Generic Ietter 85-12 requested that Kansas Gas and Electric Conpany submit plant specific information deliniated in Section IV of the Staff's Safety Evaluation Parts A, B, and C.
The enclosure to this letter provides the requested information.
Parts A.1, A.3, and B have been addressed for the SNUPPS plants.
Parts A.2 and C have been addressed as site specific for Wolf Creek Generating Station.
If you have any questions concerning this matter, please contact me or Mr. Otto Maynard of my staff.
Yours very truly, N
th
/
Glenn L. Kocster Vice President - 111 clear GM:see Enclosure Attachment xc: IO'Connor (2) w/a A
JCumins w/a OT L
I gf Ess!Si!!Nr P
201 N. Market - Wictuta, Kansas - Mad Address: PO. Box 208 i Nctuta, Kansas 67201 - Telephone: Area Code (316) 261-6451
I Mr. H.R. Denton Enclosur3 to KMIRRC 85-261 page 1
Novenber 27, 1985 Information Related to Reactor Coolant Puno (HCP) Trip Criteria A.
Determination of RCP Trio Criteria 1.
NRC Reauest Identify the instrumentation to be used to determine the PCP trip setpoint, including the degree of redundancy of each parameter signal needed for the criterion chosen.
SNUPPS Resoonse to A.1 In reference 4,
SNUPPS notified the NRC that reactor coolant system (RCS) pressure has been chosen as the trip parameter for the SNUPPS plants.
There are three wide-range pressure indicators that are available to the operator, two that receive their signals from the nuclear incore instrumentation guide tubes at the seal table (PEPI 403 and 405), and one that receives its signal from the top of the reactor vessel (BBPI 406).
These redundant, Class 1E transmitters are located outside of the containment.
Each of the transmitters is associated with a
different Separation Group of SNUPPS plant instrumentation (Ref.
FSAR Section 7.1).
The design features of this pressure instrumentation (consistent with the function, location, and environmental conditions) have been reviewed and availability is adequately assured for accident mitigation.
2.
NRC Reauest Identify the instrumentation uncertainties for both normal and adverse containment conditions.
Describe the basis for the selection of the adverse containment parameters.
- Address, as awropriate, local conditions such as fluid jets or pipe whip which might influence the instrumentation reliability.
Wolf Creek Response to A.2 Instrumentation uncertainity for wide-range pressure indicators PI-403, PI-405, and PI-406 was determined using the statistical methodology, as described in reference 3, previously approved by the NRC.
An operator reading error of one-half the smallest instrument scale dimension was statistically includM.
The resulting error for normal environmental conditions is 1.86%,
which gives an instrument uncertainity of 55.8 psi.
m
Mr. H.R. Denton Enclosura to KMLNRC 85-261 page 2
Novenber 27, 1985 The RCS wide-range pressure transmitters in SNUPPS plants are supplied by Westinghouse and are located outside containment in areas where they will not be adversely affected by accident conditions inside containment other than radiation, or accident conditions outside containment other than a potential local auxiliary steam line break which could adversely affect one of the three transmitters.
However, inplementation of the RCP trip criteria is not required for an auxiliary steam line break event. The total integrated radiation dose for a six-month post-IOCA period in the vicinity of these transmitters does not exceed the Westinghouse threshold criteria for a harsh radiation environment.
Therefore, the instrument uncertainty for normal environmental conditions applies for all cases when the RCP trip criteria may be invoked.
3.
NRC Reauest In addressing the selection of the criterion, consideration to uncertainties associated with the EOG supplied analyses values must be provided.
These uncertainities include both uncertainties in the couputer program results and uncertainties resulting from plant-specific features not representative of the generic l
data group.
If a licensee determines that the EOG alternative criteria are marginal for preventing unneeded RCP trip, it is reconnended that a more discriminating plant-specific procedure be developed.
For exanple, use of the lac-required inadequate-core-cooling instrumentation may be useful to indicate the need for RCP trip.
Licensees j
should take credit for all equipnent (instrumentation) available to the operators for which the licensee has sufficient confidence that it will be operable during the expected conditions.
SNUPPS Response to A.3 The IGTRAN Conputer code was used to perform the alternate RCP trip criteria analyses.
Both Steam Generator 'nibe Rupture (SGTR) and non-IOCA events were l
sinulated in these analyses.
Results from the SGPR l
analyses were used to obtain all of the trip parameters.
IGTRAN is a Westinghouse licensed code used for FSAR SGPR and non-IOCA analyses. The code has been validated against the January, 1982 SGTR event at the Ginna plant. Results of this validation show that IGTRAN can accurately predict RCS pressure, RCS tenperatures and secondary pressures, especially in the first ten minutes of the transient. This l
l is the critical time period when mininum pressure and subcooling is experienced.
l l
Mr. H.R. Denton Enclosura to KMLNRC 85-261 page 3
Novenber 27, 1985 The major causes of uncertainties and conservatism in the conputer program results, assuming no changes in the initial plant conditions (i.e.,
full power, pressurizer
The following are considered to have the most inpact on the determination of the RCP trip criteria:
1.
Break flow 2.
SI flow 3.
Decay heat 4.
Auxiliary feedwater flow The following sections provide an evaluarion of the uncertainties associated with each of these iteus.
To conservatively sinulate a double ended tube rupture in safety analyses, the break flow model used in IGTRAN includes a substantial amount of conservatism (i.e, predicts higher break flow than actually expected).
Westinghouse has performed analyses and developed a more realistic break flow model that has been validated against the Ginna SGER tube rupture data.
The break flow model used in the liOG analyses has been shown to be approximately 30% conservative when the effect of the higher predicted break flow is conpared to the more realistic model.
The consequence of the higher predicted break flow is a lower than expected predicted mininum pressure.
Tl:e SI flow inputs used were derived from best estimate calculations, assuming all SI trains operating.
An evaluation of the calculational methodology shows that these inputs have a maxinum uncertainty of 10%.
The decay heat model used in the liOG analyses was based on the 1971 APE 5.1 standard.
When conpared with the more recent 1979 APE 5.1 decay heat inputs, the values used in the liOG analyses are higher by about 5%.
To determine the effect of the uncertainity due to the decay heat model, a sensitivity study vas conducted for the SGIR.
The results of this study show that a 20% decrease in decay heat resulted in only 1 % decrease in BCS pressure for the first 10 minutes of the transient.
Since RCS tenperature is controlled by the steam dunp, it is not affected by the decay heat model uncertainty.
The AFW flow rate input used in the liOG analyses are best estimate values, assuming that all auxiliary feed punps are running, mininum punp start delay, and no throttling. To evaluate the uncertainties with AFW flow rate, a
l
Mr. H.R. Denton Enclosur3 to KMLNRC 85-261 page 4
Novenber 27, 1985 sensitivity study was performed. Results from the two loop plant study show that a 64% increase in AEW flow resulted in only an 8% decrease in mininum RCS pressure.
Results from the 3 loop phnt study show that 27% increase in AEW flow resulted in only a 3% decre:ase in mininum RCS pressure.
The effects of all these uncertainties in the models and input parameters were evaluated and it was concluded that the contributions from the break flow conservatism and the SI uncertainty dominate.
The calculated overall uncertainty in the EOG analysis for the SNUPPS plants, as a result of these considerations, is -150 psig to +150 psig for the mininum RCS pressure RCP trip setpoint. Due to the minimal effects from the decay heat model and AEW input uncertainties on the RCS pressure uncertainty, the calculational uncertainity result includes only the effects of the uncertainties doe to the break flow model and SI flow inpits.
There are no uncertainties resulting from plant specific features not representative of the generic data group. PCP trip on RCS pressure has been selected as the appropriate trip paraneter.
PCS pressure provides anple margin (in excess of 300 psig) to the trip setpoint for the non-IOCA accidents that were evaluated in the EDG analyses.
B.
Potential Reactor Coolant Puno Problems 1.
NRC Reauest Assure that containment isolation, including inadvertent isolation, will not cause problens if it occurs for non-IOCA transients and accidents.
a.
Demonstrate that, if water services needed for RCP operations are terminated, they can be restored fast enough once a non-IOCA situation is confirmed to prevent seal damage or failure.
b.
Confirm that containment isolation with continued punp operation will not lead to seal or punp damage or failure.
SPRlPPS Response to B.I.a The automatic (Hi-3) containment isolation signal isolates conponent cooling water to and from each RCP motor and thermal barrier. Automatic (Hi-1) or manual containment isolation isolates RCP seal water return to the Chemical and Volume Control System (CVCS) but does no_t isolate seal water injection to the RCP.
Although the RCP seal system can operate for some time with seal water injection only,
Mr. H.R. Denton Enclosura to TM NRC 85-261 page 5
Novenber 27, 1985 the RCP motor bearings are more limiting and are qualified for 10 minutes operation without conponent cooling water with no resultant damage (reference 10).
Ten minutes provides adequate time for the operator to determine either that a non-IOCA accident has occurred and to restore conponent cooling water flow to the RCP or to determine that a IOCA has occurred and to trip the BCPs.
SNUPPS ResDonse to B.I.b Containment isolation (Hi-3) does not isolate seal water injection, but does isolate conponent cooling water to the reactor coolant punp and seal water return to the CVCS. The limiting conponents associated with the RCP under these conditions are the motor bearings which are qualified for 10 minutes operation without resultant damage.
As described in the response to B.1.a, 10 minutes provides adequate time to restore conponent cooling water flow or to trip the RCPs as appropriate.
2.
NRC Reauest Identify the conponents required to trip the RCPs, including relays, power supplies, and breakers.
Assure that RCP trip, when determined to be necessary, will occur.
If necessary, as a result of the location of any critical conponent, include the effects of adverse containment conditions on RCP trip reliability. Describe the basis for the adverse containment prarameters selected.
SNUPPS Response to B.2 The components required to manually trip the RCPs from the control room are listed on Attachment A.
To trip one RCP, three active devices nust function:
the hand-switch on the MCB; the trip coil mechanism; and the breaker. The Control and 'ntrbine Buildings are not subject to harsh environmental conditions resulting from a LOCA in the containment.
There are no conponents required for the function of tripping the RCPs located inside containment.
The only equipnent inside containment associated with interrupting power to the RCP motors are the RCP motors, electrical power cable to the motors, the in-containment portion of the electrical penetration assenblies carrying power to the pmp motors, the differential current transformers ( on the power cables) and associated electrical cable and electrical penetration assenblies for the differential ralay current circuits.
It is unlikely that adverse e
Mr. H.it. Denton Enclosura to KMrm C 85-261 page 6
Novenber 27, 1985 environmental conditions could affect the above equipment prior to initiation of a manual tript however, degradation of either power cables or differential relay current circuits would most likely result in an RCP trip signal generated by the differential relay circuits. Once the BCP breakers are open, the RCPs receive no power and further degradation of the equipment inside containment will not result in the breakers reclosing.
Adverse _ containment parameters assumed in this evaluation are those post-IOCA conditions of tenperature, pressure,
- humidity, radiation, chemical spray, and potential submergence provided in the SNUPPS NUREG-0588 Submittal (reference 5).
C.
Operator trainina and procedures (RCP Trio) 1.
NIC Request Describe the operator training program for RCP trip.
Include the general philosophy regarding the need to trip punps versus the desire to keep pungs running.
Nblf Creek RenLouse to C.1 Reactor Coolant Pung operation and the philosophy regarding tripping of the reactor coolant pungs is discussed and reviewed in both Licensed Operator Training and Licensed Operator Requalification Training.
These training sessions present information concerning RCP construction, operation of the RCP's during normal, off-normal, and emergency conditions, RCP trip / restart
- criteria, and the effects of tripping RCP's versus the effects on not tripping RCP's during different operational conditions.
The Westinghouse Owners' Group (ROG)
Emergency Response Guidelir.es (ERG) Rev.
1, from which the NCGS Emergency Procedures (ENG) were generated, provide the basic philosophy regarding RCP trip / restart criteria studied during the Licensed Operator Training and Licensed Operator Requalification Program.
2.
NRC Request Identify those procedures which include RCP Trip related operations:
(a)
RCP Trip using EDG alternate criteria (b)
RCP Restart (c)
Decay heat removal by natural circulation (d)
Primary system void removal l
~
Mr. H.k. Denton Enclosura to KMLNRC 85-261 page 7
Novenber 27, 1985 (e)
Use of Steam Generators with and withou RCPs operating (f)
PCP Trip for other reasons.
Wolf Creek Response to C.2 RCS pressure was selected as the appropriate "}iOG Alternate Criteria" for the SNUPPS plants as described by reference 4.
'Ihe actual value of RCS pressure at which the RCPs are to be tripped was calculated per the RCP Trip / Restart section of the generic issues portion of the liOG ERG Rev.
1 Background document utilizing NCGS specific parameters.
Per the ERGS and the NCGS EMGs, the following procedures utilize the aforementioned RCP trip criteria:
DIG E-0, Safety Injection, Step 22, Foldout Ib dig ES-01, Rediagnosis, Foldout Ib EMG ES-02, Reactor Trip Response, Foldout lb EMG ES-04, Natural Circulation Cooldown, Foldout Ib EMG ES-05, Natural Circulation Cooldown with Steam Void in Vessel (without RVLIS), Foldout lb EMG ES-06, Natural Circulation Cooldown with Steam Void in vessel (with RVLIS), Foldout lb EMG E-1, Loss of Reactor or Secondary Coolant, Step 1 EMG E-3, Steam Generator Tube Rupture, Step 1 EMG C-21, Uncontrolled Depressurization of All Steam Generators, Step 3b All RCP related steps within the NCGS Emergency Operating Procedures are written in accordance with the !!DG Emergency Response Guidelines Rev.
1 and background documents.
Thus the LOG ERGS Rev. I provide the guidance for the NCGS 1
EMGs in regard to all RCP operations.
1 1
I L
Mr. H.R. Denton Enclosura to KMLNRC 85-261 page 8
Novenber 27, 1985 Referenceq 1.
NRC Generic Ietter 85-12, dated 6/28/85 2.
SINRC 83-0021, dated 4/22/83, Response to NRC Generic Ietter No.83-10c.
3.
SIRRC 84-0050, dated 3/23/84, Response to NRC Questions on Setpoint Methodology for SNUPPS 4.
SLNRC 84-66, dated 4/13/84, Final Response to NRC Generic Ietter No.83-10c.
5.
SNUPPS Report of Independent Review of Environnental Qualification Programs to NUREG-0588, SLNRC 83-15, dated 3/10/83 as revised by SLNRC 83-30, dated 5/27/83 and SLNRC 84-13, dated 2/1/84.
6.
liOG Emergency Response Guidelines Executive Volune, dated 3/20/84.
7.
liOG Evaluation of Conputer Code Uncertanties, dated 8/16/85.
8.
The following Bechtel Drawing No.'s.
E-03PA02 Rev. 12 E-03PA05 Rev. 11 E--03BB01 Rev. 13 E-03PA14 Rev. 8 E-OlIE01 Rev. 13 M-02BB02(Q) Rev.18 M-02BB04(Q) Rev. 6 M-0G063-07 9.
6 6 PSAR Table 3.ll(B)-3 10.
SNUPPS FSAR Section 5.4.1.
11.
BLUE 1895, dated 9/5/95, Circuits Penetrating Containment Excluded from T/S Table 3.8-1, 12.
Instruction manual for 13.8 KV Switchgear, E-009-0223-05, E010001 Rev. 7.
Mr. H.R. Dentx>n Attachnent to KMLNRC 85-261 page 9
Novenber 27, 1985 Eauiament Remired to Manually Trio Reactor Coo 1Pnt Punos from the Control Room Eaultanent Iocation/ Room #
A.
Main Control Board RLO21 Control Building /3601 (Main Control Room) 1.
Hand Indicating Switch (one per punp)
B.
Switchgear (PA01, PA02)
'Ibrbine Building /
2033 ft. el. W 1.
Breakers (one per punp) 2.
Trip Coil (one per punp)
(energize-to-trip) 3.
Fuses (2) for breaker control power 4.
Breaker "a" contacts (two per breaker) 5.
Wiring / terminal blocks C.
125 VDC Distribution Panel
'Ibrbine Building /
(PK41, PK62) 2033 ft. el. W (control power to breakers) 1.
Switches (#3, #4) 2.
Wiring D.
125 VDC Switchgear (PK01,PK02)
'Ibrbine Building /
2033 ft. el. W l.
Fuses, wiring E.
125 VDC Batteries (IEll, PK12)
Turbine Building /
l (power to 125 VDC switchgear) 2033 ft. el. W i
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