ML20151P192

Proposed Tech Specs,Increasing Present +/- 1% Tolerance on Safety Mode Lift Setpoint for SRVs to +/- 3%

ML20151P192
Person / Time
Site:	Perry
Issue date:	07/13/1998
From:	CENTERIOR ENERGY
To:
Shared Package
ML20138K420	List: ML20151P189 ML20151P192
References
NUDOCS 9807160322
Download: ML20151P192 (13)

Text

._ _

i PY-CEI/NRR-2298L t

4 i

1 i

TECHNICAL SPECIFICATION PAGE REVISIONS 9807160322 980713 P

't PDR ADOCK 05000440 g P

PDR

,2

_. _. _ _ _ _. _ _... _ _ _ _ _ _ _ _ ~ _ _ _ _ _ _ _. _ _. _ _ _. _ _.

S/RVs 3.4.4 3.4 REACTOR COOLANT SYSTEM (RCS) i 3.4.4 Safety / Relief Valves (S/RVs)

LC0 3.4.4 The safety function of seven S/RVs shall be OPERABLE, l

SID

)

The relief function of six additional S/RVs shall be 3

OPERABLE.

.i APPLICABILITY:

MODES 1, 2, and 3.

i ACTIONS l

CONDITION REQUIRED ACTION COMPLETION TIME I

i

}

A.

One or more required A.1 Be in MODE 3.

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> l'

S/RVs inoperable.

NiQ A.2 Be in MODE 4.

36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> i

i SURVEILLANCE REQUIREMENTS j

SURVEILLANCE FREQUENCY SR 3.4.4.1 Verify the safety function lift setpoints In accordance i

of the required S/RVs are as follows:

with the l

Inservice Number of Setpoint Testing Program S/RVs insia)

IMN 8

1165 i E 4- ~

6 1180 i & &

E4 1

5 1190 1 & 9-k57)

(continued)

PERRY - UNIT 1 3.4-10 Amendment No. 69

a a

__4u.

..p._

as a

..__42nA4_4, o_

t Au.,,

A,..Ad._4 wi-M.

4-e

-m4.&....

m w

an,m..AJ_p4 S_

4, nujf m..h rm 6

s_m 4

w.,J,M-em._edt2 9

PY-CEI/NRR-2298L TECHNICAL SPECIFICATION BASES REVISIONS

SIRVs B 3.4.4 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.4 Safety / Relief Valves (S/RVs)

!DiEORI6ATION ONL

,,sss BACKGROUND The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Ref.1) requires the Reactor Pressure Vessel be protected from overpressure during upset conditions by self actuated safety valves. As part of the nuclear pressure relief system. the size and number of safety / relief valves (S/RVs) are selected such that peak pressure in the nuclear system will not exceed the ASME Code limits for the reactor coolant pressure boundary (RCPB).

The S/RVs are located on the main steam lines between the reactor vessel and the first isolation valve within the drywell.

Each S/RV discharges steam through a discharge line and quencher to a location below the minimum water level in the suppression pool. With one or more S/RVs stuck open, operators will close the S/RV. thus minimizing the increase in suppression pool water temperature.

The S/RVs can actuate by either of two modes: the safety mode or the relief mode.

In the safety mode (or spring mode

~,

of operation), the direct action of the steam pressure in the main steam lines will act against a spring loaded disk that will pop open when the valve inlet pressure exceeds the spring force.

In the relief mode (or power actuated mode of operation), a pneumatic operator and mechanical linkage assembly are used to open the valve by overcoming the s ring force, even with the valve inlet pressure equal to O g.

The pneumatic operator is arranged so that its malfun ion will not prevent the valve disk from lifting if steam inlet pressure reaches the spr lift set pressures.

In the relief mode, valves may opened manually or automatically at the selected preset pressure. Six of the S/RVs providing the relief function also provide the low-low set relief function s acified in LC0 3.6.1.6. " Low-Low Set (LLS)

Valves."

Eight of the S/RVs that provide the relief function are art of the Automatic ressurization System specified in.C0 3.5.1. "ECCS-Operat ng."

The instrumentation associated with the relieY valve function and low-low set function is discussed in the Bases for LCO 3.3.6.4. " Relief and Low-Low Set (LLS) Instrumentation."

i-(continued)

PERRY - UNIT 1 B 3.4-18 Revision No.1

S/RVs

~

INFORMATION ONLY BASES BACKGROUND and instrumentation for the ADS function is discussed in (continued)

LCO 3.3.5.1. " Emergency Core Cooling System (ECCS)

Instrumentation."

APPLICABLE The overpressure protection system must accommodate the SAFETY ANALYSES most severe pressure transient. Evaluations have determined that the most severe transient is the closure of all main steam isolation valves (MSIVs) followed by reactor scram on high neutron flux (i.e.. failure of the direct scram associated with NSIV position) (Ref. 2). For the purpose of the analyses, the 13 safe valves with the highest setpoints were assumed to operational. Therefore, by requiring six S/RVs to be OPEPABLE in the relief mode and seven in the safety mode, the accident analyses assumptions are adequately met. The analysis results demonstrate that the design S/RV capacity is capable of maintaining reactor pressure below the ASME Code limit of 110% of vessel design pressure (110% x 1250 psig = 1375 ps

).

This LCO helps to ensure that the acceptance limit of 75 psig is met during the design basis event.

Reference 3 discusses additional events that are expected to actuate the S/RVs. From an overpressure standpoint. the design basis events are bounded by the MSIV closure with flux scram event described above.

S/RVs satisfy Criterion 3 of the NRC Policy Statement.

LCO The safety function of seven S/RVs is required to be OPERABLE in the safety mode, and an additional six S/RVs (other than the seven S/RVs that satisfy the safety function) must be OPERABLE in the relief mode. The requirements of this LC0 are a)plicable only to the capability of the S/RVs to meclanically open to relieve excess pressure.

In Reference 2. an evaluation was performed to establish the parametric relationshh between the peak vessel pressure and the number of OPERAB E S/RVs.

)

The results show that with a minimum of seven S/RVs in the safety mode and six S/RVs in the relief mode OPERABLE. the ASME Code limit of 1375 psig is not exceeded.

The S/RV set)oints are established to ensure the ASME Code limit on peac reactor pressure.is satisfied. The ASME Code specifications require the lowest safety val'!e be set at or l

(continued)

PERRY - UNIT 1 B 3.4-19 Revision No. 1 l

l S/RVs B 3.4.4 1

BASES LCO below vessel design pressure (1250 psig) and the highest (continued) safety valve be set-so the total accumulated pressure does not exceed 110% of the design pressure for conditions. The transient evaluations in Reference 3 are based on these g

setleints, but also include the additional uncertainties of-iB of the nominal setpoint to account for potential setpoint drift to provide an added degree of conservatism.

Operation with fewer valves OPERABLE than specified, or with setpoints outside the ASME limits, could result in a more severe reactor response to a transient than predicted, possibly resulting in the ASME Code limit on reactor pressure being exceeded, i

APPLICABILITY In MODES 1. 2. and 3. the specified number of S/RVs must be OPERABLE since there may be considerable energy in the reactor core and the limiting design basis transients are assumed to occur. The S/RVs may be required to provide pressure relief to discharge energy from the core until such time that the Residual Heat Removal (RHR) System is capable i

of dissipating the heat.

I In MODE 4. decay heat is low enough for the RHR System to provide adequate cooling, and reactor pressure is low enough that the overpressure limit is unlikely to be approached by assumed operational transients or accidents.

In MODE 5. the reactor vessel head is unbolted or removed and the reactor is at atmospheric pressure. The S/RV function is not needed during these conditions.

ACTIONS A.1 and A.2 With less than the minimum number of required S/RVs OPERABLE, a transient may result in the violation of the ASME Code limit on reactor pressure.

If one or more recuired S/RVs are inoperable, the plant must be brought to a PODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and to MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

(continued)

PERRY - UNIT 1 B 3.4-20 Revision No. 1

S/RVs B 3.4.4

-BASES (continued)

SURVEILLANCE SR~ 3.4.4.1 i

REQUIREMENTS

'This Surveillance demonstrates that the required S/RVs will open at the pressures assumed in the safety analysis of Reference 2.

The demonstration of the S/RV safet lift settings must be performed during shutdown, y function since this is a bench test, and in accordance with the Inservice Testing Program. The lift setting pressure shall correspond to ambient conditions of the valves at nominal operating temperatures and pressures.

The Frequency was selected because this Surveillance must be performed during shutdown conditions and is based on the

-(

3 time between refuelings.,

I a.u-2tA SR 3.4.4.2 The required relief function S/RVs are required to actuate automatically upon ' receipt of specific initiation signals.

A system functional test is performed to verify that the mechanical portions i.e., solenoids of the automatic relief function operate as designed when initiated either by an actual or simulated initiation signal. Tte LOGIC SYSTEM-FUNCTIONAL TEST in SR 3.3.6.4.4 overlaps this SR to provide complete testing of the safety function.

The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an un)lanned transient if the Surveillance were perfonned with tie reactor at power.

0)erating experience has shown these components usually pass tie SR when performed at the 18 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

This SR is modified by a Note that excludes valve actuation.

This prevents an RPV pressure blowdown.

SR 3.4.4.3 A manual actuation of each required S/RV is performed to verify that the valve is functioning properl blockage exists in the valve discharge line.y and that no This can be demonstrated by the response of the turbine control valves or bypass valves, by a change in the measured steam flow, by the S/RV discharge pipe pressure switch, or any other method suitable to verify steam flow (e.g., tailpipe temperature.)

(continued)

PERRY - UNIT 1 B 3.4-21 Revision No. 1

1

~

l l

Bases 3.4.4 INSERT B 3.4-21A The safety lift setpoints will still be set within a tolerance ofi 1%, but the setpoints will be tested to within i 3% to determine acceptance or failure of the as-found valve lift setpoint

.(Reference 4).

L i

i i

1 5

S/RVs B 3.4.4 BASES SURVEILLANCE SR 3.4.4.3 (continued)

REQUIREMENTS Adequate reactor steam pressure must be available to perform this test to avoid damaging the valva. Also, adequate steam flow must be passing through the main turbine or turbine bypass valves to continue to control reactor pressure when the S/RVs divert steam flow upon opening. Sufficient time is therefore allowed after the required pressure and flow are achieved to perform this test. Adequate pressure at-which this test is to be performed is the pressure recommended by the valve manufacturer. Plant startup is I

allowed prior to )erforming this test because valve OPERABILITY and t1e setpoints for overpressure protection are verified, per ASME requirements, prior to valve l

installation. Therefore, this SR is modified by a Note that states the Surveillance is not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure and flow are adequate to perform the test. The 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowed for

)

manual actuation after the required pressure and flow are i

reached is sufficient to achieve stable conditions for testing and provides a reasonable time to complete the SR.

SR 3.4.4.2 and the LOGIC SYSTEM FUNCTIONAL TEST performed in SR 3.3.6.4.4 overlap this surveillance to provide complete testing of the assumed safety function.

If the valve fails i

to actuate due only to the failure of the solenoid but is capable of opening on over)ressure, the safety function of the S/RV is considered OPEMBLE.

The 18 months on a STAGGERED TEST BASIS Frequency ensures that each solenoid for each S/RV is alternately tested. The 18 month Frequency was developed based on the S/RV tests required by the ASME Boiler and Pressure Vessel Code.

Section XI (Ref.1). Operating experience has shown that these components usually pass the Surveillance when

)erformed at the 18 month Frequency. Therefore, the

requency was concluded to be acceptable from a reliability standpoint.

REFERENCES 1.

ASME Boiler and Pressure Vessel Code. Sections III and XI.

2.

USAR, Chapter 15. Appendix 15B.

3.

USAR. Section 15.

i 4

\\

4. NR(_.Sa(t.fY UQlNon & NEDC ~30'53Po Murth 8'04l3-PERRY - UNIT 1 B 3.4-22 Revision No. 1 m.-

PY-CEI/NRR-2298L SAFETY REVIEW FOR PERRY NUCLEAR POWER PLANT SAFETY / RELIEF VALVE SETPOINT TOLERANCE RELAXATION / OUT-OF-SERVICE ANALYSES (NEDC-32307P) i PROPRIETARY i

O

\\

General Electric Company AFFIDAVIT I, George B. Stramback, being duly sworn, depose and state as follows:

(1) I am Project Manager, Regulatory Services, General Electric Company ("GE") and have been delegated the function of reviewing the information described in paragraph (2) which is sought to be withheld, and have been authorized to apply for 4'

its withholding.

(2) The information sought to be withheld is contained in the GE proprietary report NEDC-32307P, Safety Reviewfor Perry Nuclear Power Plant Safety / Relief Valves Setpoint Tolerance Relaxation / Out-of-service Analyses, Class III (GE Company Proprietary Information), dated May 1994. This information is delineated by bars i

marked in the margin adjacent to the specific material.

(3) In making this application for withholding of proprietary information of which it is the owner, GE relies upon the exemption from disclosure set forth in the Freedom of Information Act ("FOIA"),5 USC Sec. 552(b)(4), and the Trade Secrets Act,18 USC Sec.1905, and NRC regulations 10 CFR 9.17(a)(4), 2.790(a)(4), and 2.790(d)(1) for " trade secrets and commercial or financial information obtained from a person and privileged or confidential" (Exemption 4). ' Die material for which exemption from disclosure is here sought is all " confidential commercial information", and some portions also qualify under the narrower definition of " trade secret", within the meanings assigned to those terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass Enerav Proiect v. Nuclear Regulatory Commission. 975F2d871 (DC Cir.1992), and Public Citizen Health Research Groun

v. FDA,704F2dl280 (DC Cir.1983).

(4) Some examples of categories of information which fit into the definition of proprietary information are:

a.

Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention ofits use by General-Electric's competitors without license from General Electric constitutes a competitive economic advantage over other companies; b.

Information which, if used by a competitor, would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product; GBS-98-3-af Perry SRV setpoint tolerance. doc Aflidavit Page 1

~ - - - - - - - -.. -. -

c.

Information which reveals cost or price information, production capacities, budget levels, or commercial strategies of General Electric, its c'ustomers, or its suppliers; d.

Information which reveals aspects of past, present, or future General Electric customer-funded development plans and programs, of potential commercial value to General Electric; F

e.

Information which discloses patentable subject matter for which it may be desirable to obtain patent protection.

The information sought to be withheld is considered to be proprietary for the reasons set forth in both paragraphs (4)b. and (4)d., above.

(5) The information sought to be withheld is being submitted to NRC in confidence.

The information is of a sort customarily held in confidence by GE, and is in fact so held. The information sought to be withheld has, to the best of my knowledge and j

belief, consistently been held in confidence by GE, no public disclosure has been made, and it is not available in public sources. All disclosures to third parties including any required transmittals to NRC, have been made, or must be made, pursuant to regulatory provisiers or proprietary agreements which provide for maintenance of the.information in confidence. Its initial designation as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in paragraphs (6) and (7) following.

(6) Initial approval of proprietary treatment of a document is made by the manager of the originating component, the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge. Access to such documents within'GE is limited i a "need to know" basis.

(7) The procedure for approval of extemal release of such a document typically requires review by the staff manager, project manager, principal scientist or other equivalent authority, by the manager of the cognizant marketing function (or his delegate), and by the Legal Operation, for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside GE are limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary agreements.

(8) The information identified in paragraph (2), above, is classified as proprietary because it would provide other parties, including competitors, with a valuable comparative data base and interpretive information regarding transient computer

]

codes and transient analysis, and Safety Relief Valve (SRV) discharge piping anchor loads and quencher anchor loads test results. The computer code and SRV related j

GBS 98 3-af Perry SRV setpoint tolerance. doc Affidavit Page 2 i

.1

tests were conducted and expended by General Electric to develop this information in support of the BWR Owners' Group.

(9) Public disclosure of the information sought to be withheld is likely to cause substantial harm to GE's competitive position and foreclose or reduce the availability of profit-making opportunities. The information is part of GE's comprehensive BWR technology base, and its commercial value extends beyond the original development cost. The value of the technology base goes beyond the extensive physical database and analytical methodology and includes development of the expertise to determine and apply the appropriate evaluation process.

The research, development, engineering, and analytical costs comprise a substantial investment of time and money by GE.

The precise value of the expertise to devise an evaluation process and apply the correct analytical methodology is difficult to quantify, but it clearly is substantial.

GE's competitive advantage will be lost ifits competitors are able to use the results of the GE experience to normalize or verify their own process or if they are able to claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions.

The value of this information :o GE would be lost if the information were disclosed to the public. Making such intitmation available to competitors without their having been required to undertake a similar expenditure of resources would unfairly provide competitors with a windfall, and deprive GE of the opportunity to exercise its competitive advantage to seek an adequate return on its large investment in developing these very valuable analytical tools.

GBS-98 3-af Peny SRV setpoint tolerance. doc Affidavit Page 3

I STATE OF CALIFORNIA

)

)

ss:

l COUNTY OF SANTA CLARA

)

l l

George B. Stramback, being duly sworn, deposes and says.

l That he has read the foregoing affidavit and the matters stated therein are true and correct to the best of his knowledge, information, and belief.

Executed at San Jose, California, this 4 day of [/J 1998.

l

/

G/6Tge B.4tramback

~

General Electric Company k 7Fday of MM

,1998.

Subscribed and sworn before me this

'a NOTARY ORMA SnT. ees [

Notary Public, State of California I

SAN 3

wc l

l GBS-98-3-af Perry SRV setpoiat tolerance. doc Affidavit Page 4