JAFP-89-0214, Responds to NRC Insp Rept 50-330/88-29 Re Monitoring of Pneumatic Supply Pressure to Safety Relief Valves & Issue Concerning Deficiencies in Tech Specs for LPCI Sys. Corrective Actions:Operating Procedure Will Be Updated

Responds to NRC Insp Rept 50-330/88-29 Re Monitoring of Pneumatic Supply Pressure to Safety Relief Valves & Issue Concerning Deficiencies in Tech Specs for LPCI Sys. Corrective Actions:Operating Procedure Will Be Updated

ML20247C454
Person / Time
Site:	FitzPatrick
Issue date:	03/20/1989
From:	Fernandez W POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
JAFP-89-0214, JAFP-89-214, NUDOCS 8903300194
Download: ML20247C454 (13)

Text

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' James A. FitzPetrick Nuclear Pour Plint

. P.O. Box 41 Lycoming, New York 13093 315 342-3840

' #> NewYorkPower William Femandez 11

&- Authort .

sesident Manager March 20, 1989 JAFP-89-0214 United States Nuclear Regulatory Commission Mail Station PI-137 -

Washington, DC 20555 Attention: Document Control Desk

Subject:

James A. FitzPatrick Nuclear Power Plant Docket No. 50-333 Insnection No. 88-29

Dear Sirs:

l This letter is our 30-day response requested in Inspection Report No. 50-330/88-29. The first issue concerns the monitoring of 1 pneumatic supply pressure to safety relief valves. The second <

issue concerns deficiencies in Technical Specifications for the ,

Low Pressure Coolant Inj ection (LPCI) System.

A. Monitoring of the Drywell Pneumatic Supply Pressure (0 pen Items 87-19-01, 88-29-03, 88-29-04)

NRC Bulletin 80-2J required three conditions with respect to the pneumatic supply to the drywell:

1. Protection of the safety relief valve pneumatic supply from overpressure.

2. Annunciation of high and low pneumatic supply pressure conditions to the control room operator.' The sensor should be as close as practical to the safety relief valves and downstream of any check valve connecting two or more pneumatic sources.

3. Appropriate operating arocedures to guide the operator upon occurrence of higa or low pneumatic supply pressure.

The FitzPatrick plant response to this bulletin, dated March 19, 1981, committed to the following during the 1981 refueling outage:

1. Provide overpressure protection on the pneumatic supply line just prior to the containment penetration.

2. Provide high and low pressure alarms. 1 8903300194 s90320  ;

DR ADOCK 0500 1

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TO: USNRC March 20, 1989 FROM: W.-FERNANDEZ JAFP-89-0214-

SUBJECT:

INSPECTION NO. 88-29 Page '

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3. Implement operator guidance in the form of Annunciator Response Procedures for the high and low pneumatic pressure conditions.

1 During the 1981 refueling outage the overpressure protection j and alarms were installed as committed. These installations met the intent of the bulletin fully. However, the Authority's response incorrectly stated that Annunciator Response Procedures would be revised to provide guidance for the high and low pressure conditions. The Authority's intent was to upgrade an existing comauter alarm for low pneumatic supply pressure to provide both high and low j pressure alarms.

Annunciator Response Procedures are written only for panel  !

annunciator alarms. Specific procedures for prescribing .;

i operator actions for computer alarms are generally placed in ]

1 the special procedures section of operating procedures. No i changes to the special procedure section of the operating 1 l procedures were generated for either high or low pressure at j the time the modification was installed. In December of i 1984, while responding to an INPO SOER regarding inadvertent j actuation of safety relief valves on high pneumatic l pressure, procedures for responding to a high pressure alarm were added to the operating procedure for the Containment j Air Dilution System, which is the source of the pneumatic supply. With this action, the Authority became partially in compliance with the original bulletin with respect to providing operator guidance for pneumatic supply alarm conditions.

During the 1987 refueling outage, a redundant supply of . I nitrogen was added to the pneumatic supply ring header in the drywell. In doing so, the existing high and low pressure alarm sensor installation no longer complied with the bulletin direction for location downstream of check valves when redundant sources were available.

Corrective Action The operating procedure for the Containment Air Dilution System will be updated by April 1, 1989 to include. operator

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response to a low pneumatic supply pressure alarm.

1 The following actions have been 4nitiated and will be completed prior to startup following the 1990 refuel outage:

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I o The pneumatic supply system will be modified to comply with the specific bulletin criteria for pressure sensor location or justification will be provided to the NRC for an alternative alarm scheme.

TO: USNRC March 20, 1989 FROM: W. FERNANDEZ JAFP-89-0214

SUBJECT:

INSPECTION No. 88-29 Page o Panel annunciation will be provided for the high and low alarms or justification provided for an alternative scheme.

o Appropriate procedures will be developed and implemented when the resulting alarms are installed.

New methods for tracking NRC commitments were implemented several years ago. In addition, new modification procedures are being presently written which specifically address review of applicable NRC bulletins during the design process. These actions should preclude this type of error in the future.

• l B. Technical Specification Revisions Inspection Concern The inspection report expressed conce-- with the timeliness of our actions to revise the Technical _y cifications to reflect the removal of the LPCI loop selection logic,from the plant in 1977. This was identified as Open Item 88-29-01.

It further requested a list of Technical Specifications known to NYPA that are incorrect and/or do not apply to the plant in its current configuration.

Response - List of Discrepancies l

Attachment A to this letter provides a list of known Technical Specification discrepancies, which reflect differences from the current plant configuration, or which otherwise require clarification, together with commitment dates for submission of amendments to address these discrepancies. All proposed amendments to the Technical Specifications necessary to correct these discrepancies will be submitted by August 31, 1989. In accordance with the inspection report request, we have not included the list of items related to Technical Specifications which are already formally docketed with the NRC.

To prepare this response concerning Technical Specifications, selected and experienced individuals representing appropriate corporate and plant departments were assigned, on a priority basis, to identify known problems and concerns. Editorial changes, clarifications, and improvements, other than those identified in Attachment A, will be included in our larger and ongoing program for improvement of the Technical Specifications.

TO: USNRC March 20, 1989 FROM: W. FERNANDEZ JAFP-89-0214

SUBJECT:

INSPECTION NO. 88-29 Page' 9 Response - LPCI Loop Select Logic - Discussion '

The inspection report addresses the fact that the Technical Specifications currently describe the LPCI loop select logic which was physically removed from the plant in 1977. (The inspection report states on Page 8 that the LPCI mode modification was completed in 1976. Also, the correct TS  !

LCO citation is 3.5.A.3 instead of "3.5.3" stated on Page 8). It goes on to express concern ". . . that this type of neglect could lead to a disregard for Technical Specifications." j i

I l The Authority agrees that resolution of this issue is long l overdue and fully recognizes its responsibility for prompt I and accurate submittals. However, the Authority disagrees that the lack of timeliness in this one instance implies ,

either neglect or a situation which would lead to a i disregard for the Technical Specifications. On the _

contrary, Technical Specification changes for the LPCI loop select modification were first submitted to the NRC (then' AEC) on July 10, 1975 - well before the modifications were completed.

Following the initial submittal to the AEC in early 1975, I there was an ongoing active engineering analysis and j consideration of this change by the Authority, General Electric, and Stone and Webster to resolva AEC/NRC concerns.

Much of this was directly tied to ongoing generic NRC I 1

concerns with BWR ECCS parameters and, in particular, LPCI run-out flow rates and also minimum acceptable flow rates for adequate heat removgl. It was-during this time period ,

that the AEC was reorganized into the NRC and the l J. A. FitzPatrick plant operation was transferred from the Niagara Mohawk Power Corporation to the Power Authority of the State of New York.

This active effort culminated in January 1981 when a second submission was made to the now NRC of a request to change the Technical Specifications to revise LPCI flow rates.

There was no response from the NRC to this request during the period from January 1981 through December 1985.

In December 1985 the Authority was informally told that the safety evaluation for the requested change was not adequate.

The Authority believed that the safety evaluation was acceptable when it was submitted in 1981, but five years later when the NRC reviewed the application, it did not meet the new criteria. Accordingly, NYPA withdraw the application to amend the specification in December 1985.

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a e f I TO: USNRC. March 20, 1989 FROM: W. FERNANDEZ JAFP-89-0214

SUBJECT:

INSPECTION NO. 88-29 Page Following this action, the Authority entered this item into its tracking system. It was, however, assigned a lower priority than other more pressing issues of concern to both  !

the Authority and the NRC.

Consequently, no further action was taken until August 1988 when a second person at the. plant site was assigned to address regulatory concerns. During late 1988 and early 1989, continued effort has been directed toward determining the best value to use for minimum acceptable pump flow in s the LPCI mode. This concern is addressed further in  !

Attachment A as Item 6.

Corrective Actions l

The Authority concurs that, regardless of the reasons, the three-year delay in resubmitting the request for amendment to this specification was not desirable. Accordingly, it initiated, beginning in July 1988, the following corrective q actions:

The JAF staff has actively supported and participated in the generic BWR Owners' Group Technical Specification improvement program for several years. Recognizing that this would result in a complete rewrite of the Technical  :

Specifications in a new format, the staff has hesitated to initiate large numbers of revisions to the existing l specifications. Nevertheless, the staff has, in recognition of the delay in resubmitting the LPCI Technical Specification change request, consciously increased the priority assigned to concerns with the existing Technibal Specifications. A second on-site licensin specialist position has been created. Two highly-qua ified 'j individuals, each having been previously licensed by the NRC {

as a senior operator and each with more than 12 years l experience at this plant site, have been assigned to these positions. At the White Plains office two additional contract engineers have been retained specifically to research and prepare Technical Specification changes as they are being identified.

I More significantly, the Authority has initiated a ,

three-phase program identifying short-term initiatives, l major technical initiatives, and long-term plans related to l improvement of the Technical Specifications.

This program includes as its first task a detailed page-by-page examination of the Technical Specifications and written identification of all discrepancies with a goal of completion by July 31, 1989. This will be followed by creation of solutions to these concerns and subsequent submission of amendment requests to the NRC.

TO: USNRC March 20, 1989 FROM: W. FERNANDEZ JAFP-89-0214 SUBJECT.: INSPECTION NO. 88-29 Page  !

While the increased programs just described are recent in  !

origin, we want to emphasize that there were and are other existing programs which have been in place for more than five years. These programs have contributed to correction l

of both the Technical Specifications and the plant procedures related to these specifications.

A site Quality Assurance review program related to Technical Specifications was initiated more'than five years ago. A principal focus of this program was the creation of a matrix to relate each Technical Specification Section 4 surveillance requirement with an associated plant ,

surveillance test. Simultaneously, these plant surveillance 4 tests are reviewed to determine their adequacy to accomplish 6 the Technical Specification surveillance requirements. This effort has been and will be used as an input into the

Technical Specification improvement program. j The routine quality assurance audit and surveillance program also identifies problems related to Technical Specifications. For example, the quality assurance program had already identified problems related to the LPCI loop select logic technical specification and issued a non-conformance report in June 1988, six months prior to the findings of NRC Inspection 88-29. i Accordingly, we are confident that these new initiatives, ,

when combined with the existing Quality Assurance and '

operator training programs, will improve the quality and l

accuracy of the plant Technical Specifications. The  ;

I Authority will continue to participate in the BWR Owners' l l Group Technical Specification improvement program. i C. Other Comments Concerning the Inspection Report l 1. Surveillance Observations 4.c, Pages 8 and 9 (0 pen Item 88-29-02) ,

l The inspection report expresses concern that plant i monthly operability tests for ECCS pumps do not contain l

detailed acceptance criteria for pump flow and I pressure. The Authority has reviewed the monthly testing requirements of some other nuclear plants. We have determined that our current monthly testing for operability exceeds the testing that other plants are specifically directed to perform by their Technical l Specifications. In the absence of specific NRC

, guidance on the exact extent of testing required, we I have confirmed that we are currently exceeding the l

testing done by other plants in the industry.

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TO: USNRC March 20, 1989 )

FROM: W. FERNANDEZ JAFP-89-0214 4

SUBJECT:

INSPECTION NO. 88-29 Page 3 Nevertheless, because of our interest in continually improving our program, we are incorporating the a constructive comments of the NRC into our ECCS 1 surveillance. I l

2. Maintenance Observation 5.b,.Pages 9 and 10 (86-11-01 Related)

The RHR keepfull pumps do not serve an active safety ,

function. They have not been placed in service due to 1 concerns involving the modification design. Making ]

them operational is not of high priority. The .

l environment of the crescent during normal operation or 4 plant shutdown is not considered to be adverse, as it is stated to be in the inspection report. The pumps are not expected to experience any deterioration due to idleness or the environment that could result'dn a safety hazard to the RHR system or any other i

surrounding equipment. If it is determined at a future date that these pumps are to be made operational, a thorough review of their condition and necessary preventive maintenance will be conducted.  ;

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3. Conformance with 10 CFR 50.62 Anticipated Transient Without Scram Rule 9.a, Page 15 (0 pen Item 88-29-08)

The currently designed and installed ARI system meets i the NRC requirements for testability of ARI while at The Authority's commitment was satisfied during power.

the 1988 refueling Outage.

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WILLIAM FERNA EZ WF:RTL:lar Attachment l cc: Superintendent of Power -

QA Superintendent l NRC Resident Inspector PORC Members NRCI 88-29 File Document Control Center l

WPO Records Management E. Wenzinger (USNRC, Region I)

INSPECTION 88-29 ATTACHMENT A 1 LIST OF KNOWN ERRORS IN TECHNICAL SPECIFICATIONS l

1. Pg. 45 Section: Table 4.1-1 Subj ect: Reactor Pressure Permissive-Pg. 47 Section: Table 4.1-2 Amendment 122 to the Technical Specifications reflected the deletion of the pressure switches used to bypass the main steam isolation valve (MSIV) reactor scram signal and.to bypass the MSIV isolation signal on low main condenser vacuum, when reactor pressure is below the setpoint, with the mode switch in the refuel- 3 or startup positions. The effect of the changes is to make the.

bypass dependent on the position of the mode switch alone and independent of the reactor pressure. The Authority's application i failed to delete two references to the " Reactor Pressure Permissive" from lists of required instrumentation in the above.

tables. The Authority will submit a proposed Technical Specification change to correct this discrepancy by July 19, 1989.

2. Page 59 Section: 3.2 Bases

Subject:

Drywell Sump Leak Rate Alarm The first paragraph in the right column, referring to the flow integrator for the containment drywell sumps states "The alarm unit in each integrator is set to annunciate before the values speciTIe3 in Section 3.6.D are exceeded." (underlining added). The values ,

of Section 3.6.D are maximum permissible leak rates in gallons per j m3' ate. In fact, we have no indication that there has ever been an '

alarm unit installed to annunciate when the leak rate limits are exceeded. The leak rate is observed by control room operators by l comparing the trend lines on strip chart recorders with a predrawn i

template with a line which represents the limiting flow rate and is mounted on the face of the recorders. In addition, in accordance with Technical Specifications, leak rate is calculated every four hours by pumping out the sumps, comparing sump pump-out integrators  ;

to determine the number of gallons of leakage aumped out of the ~

sump and then dividing the gallons pumped out ay the time between

! pump-out operations to obtain an average leak rate expressed in gallons per minute. The Authority will submit a aroposed Technical Specification change to correct this discrepancy by June 15, 1989.

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4 INSPECTION 88-29 ATTACHMENT A LIST OF KNOWN ERRORS IN TECHNICAL SPECIFICATIONS 1

3. Page 76a Section: Table 3.2-6

Subject:

Range of Control Room Suppression Chamber-Temperature Indicator The last item on the page, " Suppression Chamber Temperature" has a l specified temperature range, in the next column, for the' indicator of "50' - 250 F". For the purpose of correcting a human factors discrepancy and meeting Regulatory Guide 1.97 requirements, the location of the instrument was changed from the relay room (below the control room) to the main control room console. The instrument was also upgraded at the same time, but the new instrument has, in ,

fact, a range of 30* - 230*F. The Authority will submit a proposed j l Technical Specification change to correct this discrepancy by i

! June 15, 1989.

4. Page 76 Section: Table 3.2-6

Subject:

Number of Reactor 1 76c Notes 5 & 6 Pressure Channels On Page 76, the last item, " Reactor Pressure", in the fourth

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l column, specifies the number of channels provided by design as '

five, including both recorders and indicators. In fact, there are only three actual pressure sensors, each having an independent indicator in the control room and two recorders associated with i these pressure indicators (one narrow and one wide-range). These pressure recorders can receive (through a switch) input from either of two of the three pressure indicators. Notes 5 and 6 on Page 76c appear to limit this arrangement although the description is not I clear. Several other reactor pressure indicator exist but the ranges differ from those listed in this table of Technical { l Specifications. The Authority will submit a proposed Technical l

Specification change to correct this discrepancy by l August 31, 1989.

5. Page 78 Section: Table 4.2-1

Subject:

Reactor Low-Low-Low Water Level Item number 2 in the table is listed as " Reactor Low-Low' Water Level". This item should be listed as " Reactor Low-Low-Low Water Level". There was a " low-low" trip. However, the "l'ow-low" water level primary containment isolation system functions cere changed to " low-low-low" level by an approved plant modification. The Authority's application for a change to Technical Specifications of June 25, 1986, which became Amendment 103, omitted the changes to this specific table. The Authority will submit a proposed i

Technical Specification change to correct this discrepancy by July 19, 1989.

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INSPECTION 88-29 ATTACHMENT A LIST OF KNOWN ERRORS IN TECHNICAL SPECIFICATIONS

6. Page 114 Section: 3.5.A.3

Subject:

LPCI Mode of RHR Page 115 4.5.A.3 System These sections refer to the LPCI mode of RHR for which the loop select logic was removed in 1977. The surveillance requirement refers to testing three pumps at 23,000 gpm which is obsolete based on the elimination of the loop select mode of LPCI. This is the subject of NRC Open Item 88-29-01 and is discussed in detail on Page 8 of that inspection report.

During the first FitzPatrick refueling outage, the LPCI system was modified. This modification eliminated the loop selection logic and divided the LPCI discharge header into two halves, one for each reactor recirculation loop. Electrical " swing-busses" for LPCI valves were also eliminated.

The NRC authorized installation of these modifications in Operating License Amendment 8. Plant operation was subsequently autho,rized in Amendment 14 and operating with LPCI valve independent power supplies, in place of the " swing-bus" arrangement, was authorized in Amendment 30.

Technical Specification changes to reflect this modification, including actual page changes, were submitted to the NRC in July 1975, well before the modification was installed in 1977. A formal amendment request was later submitted on January 6, 1981 (JPN-81-003). As noted in the inspection report, the Authority subsequently withdrew this application in a December 24, 1985 letter (JPN-85-092) to rewrite the supporting safety evaluation.

The Authority will submit a proposed Technical Specification change to reflect this modification and close Inspection Item 88-29-01 by May 31, 1989.

7. Page 114 Section: 3.5.A.3.a

Subject:

RHR Containment 115a 4.5.B.2 Cooling Mode 116 3.5.B.3 LCO 3.5.B.3 requires ". . . all remaining active components of containment cooling mode. . . " to be operable. Surveillance 4.5.B.2 requires demonstrating operability of only the ". . .

remaining redundant active components. . ." This apparent discrepancy will he corrected as noted in NRC Inspection Report Open Item 84-01-01..

With one RHR pump inoperable, LCO 3.5.B.3 for containment cooling allows 30 days of operation compared to LCO 3.5.A.3.a for the LPCI l mode which allows 7 days of operation.

The Authority will submit a proposed Technical Specification change to correct these inconsistencies by May 31, 1989.

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l l INSPECTION 88-29 ATTACHMENT A ,

LIST OF KNOWN ERRORS IN TECHNICAL SPECIFICATIONS  !

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8. Page 176 Section: 4.7.A.3

Subject:

Containment Atmosphere 177 Continuous Leak' Rate Monitor Page 176 contains only the section title, "3. Continuous Leak Rate Monitor". Page 177 on the top rights statcs, "When the primary containment is inerted, it shall be contir.uously monitored for gross leakage by review of the inerting system makeup requirements. l The monitoring system may be taken out of service for maintenance, I but shall be returned to service as soon as possible." (underline added).

l Leakage is in fact monitored by control room operators rev.ieving l the makeup requirements (frequency of ordering nitrogen to fill the I

tanks) and by recording flow integrator (totalizer) readings daily j on Surveillance Test 40D. Monitoring of drywell to suppression I chamber differential pressure also provides monitoring of drywell I integrity and suppression chamber to drywell vacuum breaker status.

A significant change in the amount of nitrogen makesp required to 4 maintain the differential pressure at the specified 1.7 psid is a direct indication of drywell leakage, vacuum breaker leakage, or low suppression pool level.

However, there is no " monitoring system" installed as is implied by l this specification. There is one flow integrator installed on one I

of the nitrogen makeup trains. There is none on the second train, but by procedure, if the second train is used, it is valved through the train with the integrator. The Authority will submit a proposed Technical Specification change to correct this discrepancy by July 19, 1989.

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INSPECTION 88-29 ATTACHMENT A LIST OF KNOWN ERRORS IN TECHNICAL SPECIFICATIONS

9. Pg. 198-209 Section: Table 3.7-1

Subject:

Process Piping  ;

Penetrating Primary Containment Table 3.7-1 specifies the primary containment isolation valves which are required to be operable. This change will update the table to reflect the as-built condition of the plant, plant modifications resulting from the NRC's TMI lessons learned program, correct typographical errors, and improve consistency with the NRC Standard Technical Specifications. The notes associated with the table are also being revised by this amendment application.

The Authority submitted an application to calete this table on I December 19, 1986 via JPN-86-060. This app]ication was based on the NRC's Technical Specification Improvement Project's final report which endorsed using the FSAR as an appropriate place for this type of information. A more up-to-date table, listing the containment isolation valves, was added to the FitzPatrick FSAR in July 1986. As of today, the NRC has not acted upon this application.

1 The Authority will submit a proposed Technical Specification change by May 31, 1989.

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10. Pg. 211-213 Section: Table 3.7-2

Subject:

Exceptions to Type C Leak Rate Tests Table 3.7-2 clarifies the 10 CFR 50 Appendix J Type C leak testing requirements which are imposed on the containment isolation valves.

This change will reflect changes proposed to Table 3.7-1 in Item 9 above.

The Authority will submit a revised proposed Technical Specification change to further update this table by May 31, 1989.

11. Page 242 Section: 4.11.E.3

Subject:

Intake Deicing Heaters Page 244 E. Bases Testing Frequency Page 242 requires checking the resistance to ground "Once per i operating cycle". Page 244, last paragraph, discusses an "Innual I check". In practice they have been and are checked on annual l

basis during the warm weather season when they are not energized.

The Authority will submit a aroposed Technical Specification change i to correct this discrepancy by July 19, 1989.

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INSPECTION 88-29 ATTACHMENT A LIST OF KNOWN ERRORS IN TECHNICAL SPECIFICATIONS

12. Page 244b Section 4.12.A.4

Subject:

Fire Pump Starting Pressures This section specifies that verification must be made ". . . that each pump starts (at 95 psig for the electric pump and 85 psig for the diesel-driven pump) . . ." Since it is generally not possible to guarantee that a pressure sensor' switch will trip at an exact p're s sure , these two specified pressures will be changed to insert.

greater than or equal to" for each specified pressure. The Authority will submit a proposed Technical Specification change to correct this discrepancy not later than May 31, 1989.

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