ML20197J119
| ML20197J119 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 10/03/1984 |
| From: | Knight J Office of Nuclear Reactor Regulation |
| To: | Novak T Office of Nuclear Reactor Regulation |
| References | |
| CON-WNP-0774, CON-WNP-774 TAC-55799, NUDOCS 8410180530 | |
| Download: ML20197J119 (10) | |
Text
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OCT 3 WM Docket No. 50-397 PEMORANDUM FOR:
Thomas M. Novak, Assistant Director for Licensing Division of Licensing FROM:
James P. Knight, Assistant Director Components and Structures Engineering Division of Engineering
SUBJECT:
SER FOR WNP-2 ON CONTAINMENT ISOLATION DEPENDABILITY BY DEMONSTRATION OF CONTAINMENT PURGE AND VENT VALVE OPERABILITY (TAC N0. 55799)
Plant Name: Washington Nuclear Project 2 Docket No.:
50-397 TAC No.:
55799 Licensing Stage:
OR Responsible Branch:
Licensing Branch No. 2 Responsible Project Manager:
R. Auluck Review Status:
Complete The enclosed Safety Evaluation Report (SER) was prepared under the direction of DE:C&SE, Equipment Oualification Branch.
The SER addresses containment isolation dependability by demonstration of containment purge and vent valve operability.
We completed our review of the information concerning operability of the containment purge and vent valves at WNP-2.
The technical review was performed by Brookhaven National Laboratory.
The staff finds the information submitted demonstrated the ability of the General Signal Corporation 24-inch and 30-inch valves, when these valves are blocked by mechanical means to an opening angle of 70 or less, to close against the buildup of containment pressure in the event of a DBA/LOCA.
Appropriate Technical Specifications should be issued to reflect the limitation of opening angle for the 24-inch and 30-inch valves.
e410180530 841003 ADOCK 05000
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GCT 3 1984 Thomas M. Novak This TAC number should be closed as no further review effort by C&SE is planned.
Original sEsnea syg James P. Knight, Assistant Director Components & Structures Engineering Division of Engineering
Contact:
R. Wright, NRR Ext. 28209 cc:
A. Schwencer R. Auluck V. Noonan G. Bagchi R. Wright E. Reeves
- 8. Miller, BNL
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WASHINGTON NUCLEAR PROJECT 2 DOCKET NO. 50-397 DEMONSTRATION OF CONTAINMENT PURGE AND VENT VALVE OPERABILITY l
1.0 Requi reinent Demonstration of operability of the containment purge and vent valves, par-ticularly the ability of these valves to close during a design basis accident, is necessary to assure containment isolation. This demonstration of operabil-ity is required by BTP-CSP 6-4 and SRP 3.10 for containment purge and vent valves which are not sealed closed during operating conditions 1, 2, 3, and 4.
2.0 Description of Purge and Vent Valves The valves identified as the containment isolation valves in the purge and vent systen are as follows:
Valve Size Valve Number (Inches)
Use Location CSP-V1 30 Vent Supply Outside Containment-Drywell C SP-V 2 30 Vent Supply Outside Containment-Drywell CSP-V3 24 Vent Supply Outside Containment-Wetwell CSP-V4 24 Vent Supply Outside Containment-Wetwell CEP-V-1A 30 Vent Exhaust Outside Containment-Drywell CEP-V-2A 30 Vent Exhaust Outside Containment-Drywell CEP-V-3A 24 Vent Exhaust Outside Containment-Wetwell CEP-V-4A 24 Vent Exhaust Outside Containment-Wetwell The containment purge and vent valves are butterfly valves manufactured by BIF, a unit of General Signal Corporation and are listed as BIF Model Number A-236765 (24-inch valves) and A-206763 (30-inch valves).
Miller Air Products Corporation Model A-83 cylinders (air open-spring close) are used for valve actuation.
The 24-inch valves use 8-inch cylinders, and the 30-inch valves use 10-inch cylinders.
The valve opening angles are presently limited to 70* (90*-full open).
3.0 Demonstration of Operability 3.1 Washington Public Power Supply System (WPPSS) has provided operability demonstration infonnation for the containment purge and vent system isolation valves used at their Washington Nuclear Project 2 (WNP 2) in the follcwing submittals :
Reference A WPPSS letter, December 19, 1983, G. C. Sorensen to A. Schwencer (NRC).
Reference B WPPSS letter, December 8,1983, G. C. Sorensen to A. Schwencer (NRC).
Reference C WPPSS letter, June 22, 1983, G. D. Bouchey to A. Schwencer (NRC).
l W(NRC).PPSS letter, February 24, 1983, G. D. Bouchey to A. Schwencer 1
Reference D i
Reference E General Signal (BIF) letter, November 16, 1982, R. Ricapito to WPPSS.
Reference F WPPSS letter, Novenber 22, 1982, R. A. Holmberg to J. Mcdonald (BIF).
Reference G NRC letter, September 16, 1982, A. Schwencer to R. L. Fergurson (Suppy System).
3.2 Detemination of dynamic torques during valve closure against the buildup of containment pressure during a LOCA is based on dynamic torque coefficients CT obtained from BIF tests perfonned using different types of disc geometry and disc and shaft orientation with respect to direction of flow. The test i
meditsn is water and no air testing was perfonned.
One of the test configura-tions included a directly connected short radius elbow upstream to study the i
i effect of flow non-unifonnity on dynamic torque.
Other configurations tested j
included valve shaft in plane with the elbow, valve shaft at 90' to the elbow, l
valve disc closing clockwise, valve disc closing counter clockwise, and disc flat upstream and disc hub upstream. From these tests, the most severe case was determined to be a vertical shaft orientation (i.e., perpendicular to the plane of the elbow) with the flatside of the disc downstream and with a clockwise rotation of the disc.
This orientation results in an approximately 30% increase in maximisn dynamic torque coefficient over the straight pipe inlet configuration.
Torque coefficients used to determine dynamic loads for WNP-2 purge and vent valves are based on this worst case configuration.
I Installation configuration details for the valves showing upstream fittings and their locations with respect to the valves are provided in Reference A.
Volisne I, Section 5.
The differential pressure AP across the valve is calculated fram the data on voltsnetric flow rate under LOCA conditions, and using the equation:
P12 _ p22 Q = 963 Cy where:
Q = Ga s fl ow in SCFH P1 = Valve upstream pressure (psia)
P2 = Valve downstream pressure (psia) l G
= Specific gravity l
T1 = Upstream temperature in
- Rankine 2
Cy = Valve coefficient = 29.9 D /gy D = Valve port diameter (in.)
Ky = Coefficient of flow ig
No-load closure time for the valves ranged from 1-1/2 to 4 seconds based on tests perfonned at BIF.
The maximum no load closure time of 4 seconds is used for the analysis with a one second instrumentation time delay for a total of 5 seconds from LOCA initiation-to-valve closure.
As an additional conservatism, the drywell dhessure and temperature rise during a LOCA is used for all valves.
Dynamic torques are calculated for both saturated steam and air as the flow m ed ia.
The calculations are summarized and shown below in Tables 1, 2, 3, and 4 (Reference B) for both the 24-inch and 30-inch valves and for steam and air fl ow.
The peak dynamic torques during closure and the seating and bearing friction torques at 0* are compared to the design torques used in the seismic analysis report and indicate positive margins.
SUMMARY
OF RESULTS Tabl e 1.
30-inch valve, airflow, (TNET = 22,174 in-lb)
Dyn anic Time Angle To rque
( s) deg.
in 4 b 1.0 90 (Full open) 11,020
- 1. 5 78.75 23,098 2.0 67.50 18,138 2.5 56.25 14,747 3.0 45.00 12,428
- 3. 5 33.75 10,780 4.0 2 2. 50 8,014 4.5 11.25 3,972 5.0 9.0 (Full closed) 0.0 *
- At full closed position, the dynamic torque is zero and the net torque is due to seating and bearing friction.
Note:
The design torque used in the sei smic analysis report No. TR-74-8 by McPherson Associates for this valve is 2 7,800 i n-l b.
r.
SUMMARY
OF RESULTS Tabl e 2.
30-inch valve, steam flow, (TNET = 22,174 in-lb) h Dynamic Time Angle To rque
( s) d eg,
in lb 1.0 90 (full open) 11,032 1.5 78.75 23,175 2.0 67.50 18,142 2.5 56.25 14,668 3.0 45.00 12,242 3.5 33.75 10,580 4.0 22.50 7,80 9 4.5 11.25 3,86 7 5.0 9.0 (full closed) 0.0 *
- At full closed position, the dynamic torque is zero and the net torque is due to seating and bearing friction.
SUMMARY
OF RESULTS Table 3.
24-inch valve, airflow, (TNET = 13,808 in lb)
Dynamic Time Angle Torque
( s) deg.
in lb 1.0 90 (Full open) 5,525 1.5 78.75 11,692 2.0 6 7.50 9,095 2.5 56,25 7,428 3.0 4 5.00 6,239 3.5 33.75 5,430 4.0 22.50 4,043 4.5 11.25 2,020 5.0 9.0 (Full closed) 0.0 *
- At full closed position, the dynamic torque is zero and the net torque is due to seating and bearing friction.
Note:
The design torque used in the sei smic analysis report No. TR-74-8 by McPherson Associates for this valve is 17,000 in lb.
T StMMARY OF REStA.TS
. Table 4 24-inch valve, steamflow, (TNET = 13,808 in-lb) i Dynamic Time Angle To rque
( s) d ag.
in/lb 1.0 90 (Full open) 5,425 1.5 78.75 11,394 2.0 6 7.50 8,921
- 2. 5 56.25 7, 21 3 3.0 45.00 6,109 3.5 33.75 5,202 4.0 22.50 3,842 4.5 11.25 1,902 5.0 9.0 (Full closed) 0.0 *
- At full closed position, the dynamic torque is zero and the net torque is due to seating and bearing friction.
3.3 Denonstration of actuator torque margin is based on the minimun spring force developed sich is equal to the spring preload.
24-inch valve (8-inch cylinder) 16,890 in-lb (preload) >13,808 in-lb (seating torque).
30-inch valve (10-inch cylinder) 32,422 in-lb (preload) >22,174 in-lb (seating torque).
3.4 Based upon previous MIC recommendations, the 30-inch and 24-inch diameter purge and supply valves are to be limited to a maximum 70* opening angle, in order to assure that correlation methods used with compressible fluid calcula-tions are valid by maintaining the Mach nunber at less than 0.3 for all condi-tions. The 70* angle restriction also precludes the possibility of flow induced opening torque, thereby assuring valve closure in less than the re-quired 5.0 seconds.
- 3. 5 The valve's pressure rating of 150 lbs is provided by vendor data sheets j
(Reference A, Attachment L, Section 6).
The analysis for status pressure loading is provided for the 24-inch valves in Reference A ( Attachnent G. Sec-tion 7.0, Sheet B61), and for the 30-inch valves in Attachnent F Section 7.4, l
Sheet 7.4.6.1 of the same reference.
l l
iL,.
e
r 3.6 WPPSS provides a structural analysis for the purge and vent valves and their operators in Reference B.
This consists of three seismic / hydrodynamic requalification reports for the 30-inch and 24-inch valves and the operators.
The requalifjcation certifies both the 24-inch and 30-inch valves are contin-gent upon ear, bolt modification and the addition of shear plates.
Acceptance criteria for the structural analysis are taken from Section III of the ASME Boiler and Pressure Vessel Code or the AISC Construction Manual, whichever is applicable.
Loads used in the analysis are the valve operating loads combined with the dynamic loads which would result fran seismic and hydrodynamic events as determined by the piping analysis for the plant.
An SRSS analysis was set up in a computer program for each valve assembly in its specific orientation.
The SRSS is taken at the maximun stress level due to seismic g-loading.
Operating loads due to seating torque force and dead weight are combined with the seismic stress by absolute sum.
Based on the results of the structural analysis, the valves will renain func-tional through 40 years of postulated hydrodynamic events, five operating basis earthquakes, and one safe shutdown earthquake.
4.0 Evaluation 4.1 Determination of dynamic torques during valve closure against the buildup of LOCA induced containment pressure is predicted on 6yr.amic torque coef-ficients (C ) obtained from BIF tests. These tests, whith included upstream T
elbows, were perfonned using both steam line and flat plate discs. Test data were obtained for both upstream and downstream flatsue disc orientations.
The latter wre always caused higher dynamic torques.
Experimental test results also showed the most severe torque (30% increase) occurred with the valve shaft perpendicular to the elbow plane, disc flat side downstream, and a clockwise disc rotation.
Recognizing that the dynamic torque coefficients detennined from model tests were based upon an incompressible water flow media, it was found that if the Mach ntsnber is kept at 0.3 or below, air and water do correlate well in the determination of dynamic torque coefficients.
WPPSS has determined by analysis that the Mach nunber of the fluid flowing through both the inner and outer isolation valves would renain below Mach 0.3 following a postulated LOCA, if the valve opening was limited to 70* maximum (90' full open).
This 70' disc angle restriction also precludes the possibility of a flow-induced opening torque, which would occur at larger disc opening angles (73* to 90*).
WPPSS has committed itself to restricting these valve openings to 70', which will limit dynamic flows to 0.3 Mach. Therefore, it is concluded that the detennination of the dynamic torque coefficients is acceptable.
7 4.2 No load closure time for the valves ranged from 1-1/2 to 4 seconds bt. sed on tests perfonned at BIF.
The maximtsn no load closure time of 4 seconds is used for the analysis with a 1 second instrtsnentation time delay for a total of 5 seconds fran LOCA initiation to valve closure.
Based on this, WPPSS applies a ramp-rise approach, acceptable to the staff, in determining dynamic torque, that is, the drywell pressure and temperature rise during a LOCA is used for all valves.
4.3 Air operator spring closure torques were evaluated for both the full open valve (90') and the restricted open (70') dynamic torques.
Also presented were the torques for both the 24-inch and 30-inch valves at all angular disc po si tions.
The table below shows the dynamic closing torques and the avail-able actuator spring torques for the 24-inch and 30-inch valves at the 70* re-stricted opening.
Dynamic Torque Actuator Spring Torque Valve Size (Closing)
Available (C1osure)
Seating Torque (Inches)
(in.lb)
(in-lb)
(in-lb) 24 12,000 18,600 13,808 30 24,000 33,700 33,700 Conservatisms accounted for in developing the dynamic torques are:
Torque coefficients used to detennine dynamic loads are based on the worst case valve installation configuration, i.e., elbow shaft out of plane with the flat side of the disc downstream and a clockwise rotation of the disc.
The flow considered was that presented at a time of 1.0 seconds.
Thi s accounts for a 1 second delay time.
As the above table shows, the 24-inch valves largest torque of 12,000 in-lb occurs at the 70* opening, with a corresponding spring return actuation torque of 18,600 in lb.
For the 30-inch valve, the largest flow-indu::ed-torque of 24,000 in-lb occurs at the 70* opening, with a corresponding spring return actuation torque of 33,700 in lb.
Accounting for the conservatism mentioned previously, it is concluded that actuator operability has been denonstrated.
4.4 An analysis was made conparing design torques with seating and bearing torques to the maximisn dynamic air and steam flow torques for the 24-inch and 30-inch valves.
Valve opening angles in approximately 11' steps were explored (see table below).
In all cases, the design torque was the higher value.
Additional calculations have detennined that the resultant stresses due to valve seating torque plus mA loadings did not exceed the allowable stress levels for faulted conditions.
~-..
Design Fully Closed Seat-Maximtm Dynamic Torque Torque ing and Bearing (in lb)
Description (in -lb)
Torque (in lb)
At 78.75*
At 67.50" 24-inch valves:
C SP-V -3,
-4.
17,000 13,808 11,692 9,095 CEP-V-3A, -4A 17,000 13,808 11,394 8,921 30-inch valves :
CSP-V-1, -2 27,800 22,174 23,098 18,188 CEP-V-1A, -2A 27,800 22,174 23,175 18,142 WPPSS provides a structural analysis acceptable to the staff for the purge and vent valves and their operators denonstrating their structural integrity.
The submittal consists of seismic / hydrostatic requalification reviews for the valves and their operators.
- 5. 0 Summary 5.1 We have completed our review of the information submitted concerning operability of the 24-inch and 30-inch containment purge and vent valves for WNP-2. We find the information submitted demonstrated that these valves have the ability to close against the buildup of pressure in the event of a DB A/ LOC A from the restricted 70' position.
Sections 4.1, 4.2, 4.3, and 4.4 are the basis for this determination.
Dace;E
}?fg OCT t 2 W MEMORANDUM FOR:
Thomas M. Novak, Assistant Director for Licensing Division of Licensing FROM:
R. Wayne Houston, Assistant Director for Reactor Safety Division of Systems Integration
SUBJECT:
INSTRUMENT SETPOINT METHODOLOGY FOR WNP-2
References:
- 1) Memorandum from R.W. Houston to T. Novak dated August 30,1984, " Instrument Setpoint Methodology for WNP-2."
- 2) Memorandum from V. Stello to H.R. Denton dated September 24, 1984, " Request by Washington Public Power Supply System for Confirmation of CRGR Review."
Plant Name:
Washington Public Power Supply System Nuclear Project No. 2 Docket No.:
6 TAC No.:
55579 Licensing Stage:
0L Project Manager:
Raj Auluck Review Branch:
ICSB Review Status:
Complete for TAC 55579 The purpose of this memorandum is to provide the backgrcund information needed in responding to concerns raised by CRGR regarding the staff's review of the protection system instrumentation setpoints for WNP-2.
The Commis.sion's regulations relevant to the issue of protection system setpoints are General Design Criterion 20, 10 CFR Part 50.36 and Part 50.46.
Criterion 20, Protection System Functions, states that "the
Contact:
M. Virgilio. ICSB X29454 t' A Y $fN [0! bob
T. Novak.
protection system shall be designed (1) to initiate automatically the operation of appropriate systems including the reactivity control systems, to assure that specified acceptable fuel design limits are not exceeded as a result of anticipated operational occurrences and (2) to sense accident conditions and to initiate the operation of systems and components important to safety." Part 50.36 states " limiting safety system settings for nuclear reactors are settings for automatic protective devices related to those variables having significant safety functions. Where a limiting safety system setting is specified for a variable on which a safety limit has been placed, the setting shall be so chosen that automatic protective action will correct the abnormal situation before a safety limit is exceeded." Part 50.46 specifies the performance criteria for the emergency core cooling systems. These criteria include a maximum peak cladding temperature, a maximum cladding oxidation, a maximun total amount of hydrogen generated, and requirements that core geometry remain amenable to cooling for long term decay heat removal. Guidance on acceptable methods for complying with these regulations is contained in Regulatory Guide 1.105, " Instrumentation Setpoints" Revision 1, dated November 1976.
The staff has on an audit basis requested applicants to demonstrate compliance with the above cited regulatory requirements.
Documentation dating back to the mid-1970's (prior to CRGR) shows that the NRC staff requested the applicants for D.C. Cook - Unit 2, Salem - Unit 2, North Anna - Unit 1, Sequoyah - Unit 1, McGuire - Unit 1, and V.C. Summer -
Unit 2 (Westinghouse-designed reactors) reply to a series of detailed questions concerning the methodology for determining instrument setpoints.
In response to these requests,each applicant provided a proprietary report that included the instrument setpoint methodology, the assumptions used in establishing the methodology, the equations used to compute each set-point and the values input to the equations to account for each sensor and signal conditioning component error.
In addition, the margin to the applicable safety analysis limit was provided for each reactor protection system instrumentation channel.
Safety evaluations written by the staff documented the details of the setpoint methodologies proposed by the individual utilities to confirm compliance with the applicable regu-lations and the bases for staff approval. As a result of the inter-actions between the applicants and the staff on this issue, mutual agreement was reached to establish more conservative than originally proposed setpoints for several trip functions.
With regard to plants with General Electric-designed reactors, the method-ology for determining instrument setpoints has not yet reached the same level of completion as that for plants with Westinghouse-designed reactors.
The staff is currently working together with a BWR Owners Group to resolve this long overdue concern.
T. Novak With respect to WNP-2, the staff identified a concern during the OL review regarding the values selected for protection system instrument setpoints. Section 7.1.2.5 of the WNP-2 FSAR includes the following statements. "The design of each safety-related system considers instru-ment drift, setability and repeatability in the selection of instrumenta-tion and controls in the determination of setpoints. Adequate margin between safety limits and instrument setpoints is provided to allow for instrument error. The safety limits, setpoints and margins are listed in Chapter 16, Technical Specifications. Refer also to Tables 7.2-1, 7.3-1-thru 5, 7.4-1 and 7.6-4, 7.6-6, 7.6-8 through 7.6-12.
Also refer to Chapter 16 Technical Specifications. The amount of instrument error is determined by test and experience.
The setpoint is selected based on these known errors."
During its review, the staff did not find that margins were included in Chapter 16. Technical Specifications, or in any of the other references.
Therefore, the staff was unable to confirm that the values selected for the key protection system setpoints were sufficiently conservative.
By letter dated September 7, 1983, from A. Schwencer (NRC) to D. W. Mazur (WPPSS), this staff concern was transmitted to the Washington Public Power Supply System (the applicant) in the form of a request for additional information, and a proposed license ccndition.
The staff requested that the applicant provide revised Technical Specification instrument'setpoint allowable values and proposed a license condition for providing additional information regarding the setpoint methodology, plant specific error allowances that were input to the setpoint methodology, and confirmation that the setpoints selected for WNP-2 ensure that the reactor core and reactor coolant system are prevented from exceeding the licensing safety limits.
By letter dated September 23, 1983, from G. C. Sorensen (WPPSS) to A.
Schwencer, the applicant responded to the request for information and proposed license condition.
In this response, the applicant expressed the opinion that the staff concern was a generic issue, common to all plants with a General Electric nuclear steam supply system (NSSS).
The applicant proposed that, in lieu of the plant specific scheduled proposed by the staff, they join other BWR applicants who had banded together in an effort to conserve resources in answering requests for information.
The applicant proposed that the schedule for answering the request for information be based on the generic schedule proposed by this BWR owners group.
l On December 20, 1983, WNP-2 was granted an operating license by the NRC.
l i
i
T. Novak By letter dated August 2, 1984, from G. C. Sorensen to A. Schwencer, the licensee for WNP-2 requested confirmation that CRGR had formally reviewed the NRC staff's request for additional information on the setpoint methodology issue, and confirmation that the request for information has received an Office of Management and Budget approval.
By memorandum dated Augyst 30, 1984 (Reference 1), we reconfirmed our desire to have the licensee for WNP-2 submit the requested information.
In addition, we provided a brief history of the NRC staff /BWR Owners Group interactions over this issue and evidence to support a conclusion that our request for additional information was not a new requirement.
We also noted that the licensee had stopped participating in the BWR Owners Group formed to address this issue, contrary to their September 23, 1983, proposal.
By memorandum dated September 24, 1984 (Reference 2), the R0GR provided an assessment of NRR staff's review practice in the area of instrument setpoints.
Two of the ROGR's findings documented in the September 24, 1984, memorandum merit a response.
The first "...it is clear that the staff's understanding and scrutiny of the setpoint methodology issue has increased over the years. The amount of detailed information required by the staff to satisfy itself that licensees' approach to establishing setpoints is conservative has also increased."; and the second, "I note here also that at least some of the information requested by the staff in this instance appears to indicate that current staff practice is as described in proposed regulatory positions included in draft Revision 2 to the Reg. Guide 1.105, which the CRGR recommended be deleted during its recent review of that Reg. Guide."
In response to the first of the R0GR's findings, we have on an audit basis since the mid-1970's requested certain applicants to demonstrate compliance with the regulations relevart to the issue of protection system setpoints.
In addition to those issues which are repeatedly reviewed on each OL application, we haie asked different questions of different applicants with similarly designed plants as part of our audit review. Setpoint methodology is an issue that we have not reviewed in detail on each OL application.
Most FSARs currently under review contain nothing more than a sentence confirming that the intent of R.G.1.105 has been satisfied in establishing the instrument setpoints. On some OL applications, we have requested additional information to examine the details of the setpoint methodology.
Where we have asked for this information over the past seven years, we have requested essentially the same level of detail, unless in response I<
to our initial request we find that the methods used are not conservative.
As an example, on December 23, 1977, the NRC licensed D.C. Cook, Unit 2.
As a condition of the license, the licensee was required to complete an ongoing effort to confirm the acceptability of the instrument setpoints.
The license condition read as follows:
T. Novak.
"Ir. diana and Michigan Power Company shall submit for Commission review within six months of the date of issuance of this operat-ing license, the following values for each Reactor Protection System and Engineered Safety Features instrumentation channel:
(a) the technical specification trip setpoint value; (b) the technical specification allowable value (the technical specification trip setpoint plus the instrument drift assumed in the accident analysis);
(c) the instrument drift assumed to occur during the interval between technical specification surveillance tests; (d) the components of the cumulative instrument bias; and (e) the maximum margin between the technical specification trip setpoint, the allowable value, and the trip value assumed in the accident analysis."
To resolve this issue, the licensee provided a proprietary report (developed by the NSSS vendor at a cost to the licensee of about
$40,000) which included the equations used to compute each setpoint, justification to support the assumptions of the methodology, the error allowances for each reactor trip and engineered safety features actuation channel component (associated with process measurement accuracy, primary element accuracy, calibration accuracy, pressure effects, temperature effects, drift, environmental allowance, and comparator setting accuracy), the safety analysis limit associated with each instrument channel setpoint and the margin when all errors are considered. A copy of this proprietary report and similar proprietary reports generated by the NSSS vendor or applicant for several Westinghouse plants and one Combustion Engineering plant are available in the ICSB's files.
By letter dated September 7,1983, the staff proposed a license condition to obtain the same level of detail concerning setpoints from WPPSS for WNP-2 as was requested in 1977 for D.C. Cook, Unit 2.
The proposed license condition reads as follows:
"Within six months following the issuance of this operating license a detailed summary discussion shall be provided on the methods used to establish the trip setpoints and allowable values for the protection system channels assumed to operate in the transient and accident analyses.
This discussion shall include the following:
T. Novak (1) the values assigned to each component of the combined channel error allowance (e.g., modeling uncertainties, analytical un-certainties, transient overshoot, response times, trip unit setting accuracy, sensor setting accuracy, test equipment accuracy, sensor drift, nominal and harsh environmental allowances, trip unit drift), the basis for these values, and the methods used to sum the individual errors. Where zero is assumed for an error, a justification that the error is negligible shall be provided.
(2) Confirmation that the setpoints selected for the initiation of protective actions ensure that the reactor core and reactor coolant system are prevented from exceeding the licensing safety limits for the transients and accidents analyzed."
It should be noted that, even though the wording of the proposed license condition has changed somewhat over the years, the staff has not requested WPPSS to provide more detailed information than that provided by other applicants / licensees over the past seven years.
We would also like to clarify the second issue addressed by the R0GR letter; that the staff is imposing new requirements that CRGR rejected during its review of Reg. Guide 1.105, Revision 2.
In the Minutes of CRGR Meeting Nu. 53, the CRGR acknowledged that the methods described in Revision 1 to Reg. Guide 1.105 have been incorporated into ISA Standard S67.04.
CRGR concurred with the staff's recommendation to endorse this standard via revision to Reg. Guide 1.105; however, CRGR disagreed with the staff on the need for modifying Regulatory Positions.
CRGR concluded that the proposed Regulatory Positions " imposed unnecessary paperwork requests," " detracted from the clarity of the standard" and were " tutorials rather than guidance." The proposed Regulatory Positions addressed setpoint methodology, periodic testing, instrument ranges, initial calibration techniques and reporting requirements.
Focusing on those proposed Regulatory Positions that address setpoint methodology, we find two issues common to the ongoing review of the BWR Owners Group setpoint methodology and the proposed Regulatory Positions. One is the definition of a safety limit, and the other a recommendation that when an instrument performance parameter (such as drift) is not readily available, engineering judgment should be used in establishing an estimate of the value. We believe that both of these issues fall within the category of tutorial and do not constitute new requirements. With regard to " unnecessary paperwork requirements,"
we do not believe that CRGR was referring to the applicants' documenta-tion supporting the setpoint values or submittal of information necessary for the staff to review an applicant's setpoint methodology.
Certain
T. Novak Regulatory Positions did propose additional paperwork requirements.
Draft 5 of the Reg. Guide included extensive recommendations regarding documentation in a three-page table.
These recommendations have not been an issue in our review of the BWR owners group setpoint methodology.
It should be noted that without amplifying Regulatory Positions,an en-dorsement of ISA S67.04 will recommend that an applicant have a docu-mented basis for each setpoint, including the parameters and assumptions upon which the setpoint selection was based.
Of all the BWRs currently under review or recently licensed, WNP-2 is somewhat of a special case. As a result of poor performance character-istics specifically related to drift, stability and sensitivity, most licensed BWRs, and all other BWRs currently under OL review or recently licensed have installed new reactor protection system instrumentation.
Pressure switches that actuate relays in the trip logics have been replaced with more reliable transmitters and trip units.
The new system is described in GE Topical Report NE00-21617," Analog Transmitter /
Trip Unit System for Engineered Safeguard Sensor Trip Input." WNP-2 has not been modified to include this new system. This was not an issue that prevented the staff from approving the design; however, it presents a unique aspect to the WNP-2 setpoint methodology.
Speci-fically, certain error allowances input to the equations used to calcu-late the setpoint and/or available margin, and the equations utilized should be different (and more conservative from the standpoint of safety) than those equations and values used for the other BWR Owners Group members. As a result of these design and analytical differences, the staff has certain special concerns regarding instrument setpoints for WNP-2.
In conclusion, we do not find that our request for additional information imposes a new requirement.
Staff practice in terms of information request-ed and acceptance criteria has been consistent in this area of review since the mid-1970's.
Our approval of WNP-2 was based in part on WPPSS's September 23, 1983, commitment to provide additional information on the basis for the setpoint values. Therefore, we recommend in response to the licensee's August 2,1984, letter, you request an action plan for answering the staff's setpoint methodology questions be provided, with a level of detail consistent with that provided by the BWR Owners Group.
Original signed by i
- n. Wayne Houston R. Wayne Houston, Assistant. Director For Reactor Safety Division of Systems Integration g, J cc:
See attached list ICSB/DSI ICSB/DSI' ICSB/DSI ADh6 MVirgilio:ct JCalvo FRosa RWHouston 10/ n /84 10/ 4 84 10/ i /84 10/
/84 1
i.
T. Novak cc:
R. Berr,ero J. Stefano R. Auluck A. Schwencer B. J. Youngblood J. Conran S. Stern Distribution:
Docket File ICSB Rdg.
l M. Virgilio (PF)(2)
WNP-2 Subject File J. Calvo F. Rosa ADRS Rdg.
J. Joyce R. Kendall R. Stevens J. Mauck i
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OFFICIAL RECORD COPY b
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