ML20010B733
| ML20010B733 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 07/28/1981 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20010B729 | List: |
| References | |
| NUDOCS 8108170480 | |
| Download: ML20010B733 (32) | |
Text
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Io UNITED STATES g
E yg NUCLEAR REGULATORY COMMISSION g
, E WASHINGTON, D. C. 20555 y+.v.../
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT T10S. 49 AND 43 TO FACILITY LICENSE HOS. OPR'-42 AND DPR-60 NORTHERN STATES POWER COMPANY PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNIT NOS. 1 AND 2 DOCKET NOS. 50-282 AND 50-306
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Introduction:==
' Dy let'.er dated May 16, 1980 the Northern States Power Company (NSP) requested an amendment to Facility Operating License Nos. DPR-42 and DPR-60 for the Prairie Ishnd Nuclear Generating Plant, Unit Nos.1 and 2 (PINGP). The amendments requested changes to the lechnical Specifica-tions (TS) for the folloking ' subjects:
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(1) To implement the protection from degrEded grid vo'ltage condition requirements (Millstone fixes).
(2) Emergency charcoal filter system test. initiation signal.
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(3) Contiinment Fan Coolers Design Performance Verification.
(4) RHR System Flow Requirements.
(5) Diesel generator surveillance.
(6) Safety relhted shock supressors.
(7) Miscellaneous corrections and clarifications.
(8) Organizational changes.
(9)
Definition of Operability.
These matters are addressed as items 1 through 9 in the following report.
This safety evaluation also addressis tae clarifications and correction of typographical errors previously contained in License Paragraph 2.C.(3) on physical protection programs and corrects omissions and typograohical errors introduced as a result of previous changes in Amendments 46/40 and 47/41 to Sections 3.1 and 4.1.
These matters are addressed as items 10 and 11 in the following report.
8108170480 81072G9 PDR ADOCK 05000282' p
. By letter dated February 20, 1980 the licensee requested amendments to Facility License Nos. DPR-42 and DPR-60 for the PINGP which would expand the control rod misalignment LCO's. This matter is addressed'as item 12 in the following report.
By letter dated July 31, 1980 the licensee requested amendments to Facility License Nos. DPR-42 and DPR-60 for the PINGP which would provide additional LCO's and surveillance requirements for fire protection systems.
This matter is addressed as item 13 in the following report.-
Discussion:
Item 1 - As a result of an evert at Millstone Unit No. 2 the NRC requested all utilities to investigate the vulnerability of each facility to similar degraded grid voltage conditions (Reference 1). NSP responded to this letter on September 20, 1976 (Reference 2). Further criteria and staff positions pertaining to degraded grid voltage orotection were transmitted to NSP by NRC generic letter. dated June 3, 1977 (Reference 3).
In response to tnis, by letters dated May 4, 1978 and October 12, 1979, NSP proposed certain design modifications and changes to. the TS.
.A. detailed review and technical evaluation of these proposed modifications and changes was performed by the Lawrence Livermore Laboratory (LLL) under contract to the NRC, and with general supervision by NRC staff. This work was reported in LLL report UCID-IS654, " Technical Evaluation of the Pro::osed Design Modifications ard Technical Specification Changes en Grid Voltage Degradation for the Prairie Island Nuclear Generating Plant Units 1 and 2" dated March 1980 (attached).
Item 2 proposes changes to TS 4.4.B.1 and 4.4.B.2, Emergency Cha. coal Filter Systems. Technical Specification 4.4.5 requires testing of both the containment shiold building ventilation system and the auxiliary building special ventilation system on a quarterly basis and on a ence per operating cycle or once per 18 month basis. The proposed change to tne TS would delete the requirement for the quarterly testing of these two systems to be initiated by a :imulated safety injection signal and v.culd allow manual initiation of these systems for the quarterly tests, i
The once per operating cycle or 18 month test of these systems would i
l continue to be initiated by a simulated safety injection signal.
Item 3 would change Technical Specification 4.5.A.3, Containment Fan Coolers, to replace the test requirements now in the TS with a more representative requirement that wculd measure terminal terperatures of the fan coil unit and coolant flowrate through the unit for contain-ment atmospheric conditions normally. experienced during power operation.
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. Item 4 would revise TS 4.5.B.3.h.2 by deleting the words "with one pump in operation." The addition of these words was somewhat misleading since there are two RHR pumps, each of 1800 gpm or greater capacity, each discharging through its train's single reactor vessel injection line.
The change has no effect on safety or environmental impact. The provision of this clarification will enhance compliance with the surveillance requirement. Since this change is administrative in nature, it need not be evaluated further.
Item 5 would revise TS:4.6.A.2.b.3 ~to include a trip of the diesel generator on occurrence af a ground fault, as well as engine overspeed and generator differential current in the list of trips that are not automatically by-passed on receipt of a safety injection signal.
The proposed change
' clarifies the TS to be consistent with the previously existing diesel generator design.
The change has no effect on safety or environmental impact. Since this change is administrative-in nature it need not be evaluated further.
Item 6 would change TS Table T.S. 3.12-1 " Safety Related Shock Suppressors (Snubbers)" by adding additional shock suppressors to the list of suppressors categorized as safety related.
Item 7, subitems a through f, proposes changes which are administrative and clerical in nature.
These include correction of typographical errors (b and c) and clarifications regarding radiation control procedures (e) and the plant functional organization (f).
Items (a) and (b) have been accounted for in previous amendments to the license.
These are administrative changes and need not be evaluated further.
Item 3-The proposed change in the corporate organizational structure i
would replace the six management positions from the Flant Manager to the President with five positions, adds a manage:..nt position for quality assurance activities to the organization chart, adds a separate position for the management of training activities and adds positions for the offsite nuclear support group.
l The proposed change in the on-site operating organization includes changing the Quality Engineer's position to Superintendant for Quality Engineering now reporting directly to the Plant Manager and other changes in titles for given positions which are otherwise unchanged.
Item 9 would change the TS to clarify the meaning of the term OPERABLE.
By letter dated May 16, 1980 the licensee responded to the staff's letter of l,
April 10, 1980 on this subject by proposing changes in the definition of OPERABLE contained in the PINGP TS.
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. Item 10- In Amendment Nos 45 and 39 to Facility License Nos DPR-42 and DPR-60 issued on February 25, 1981 we consolidated and updated information contained in the license on Physical Protection programs.
The' licensee has pointed out that the titles of the program documents referenced in Amendments 45 and 39 are not consistent with the titles of the program docaments generated by the licensee and has also identified several typographical errors and omissions in Paragraph 2.C.(3).
Therefore Paragraph 2.C.(3) of License Nos. DPR-42 and DPR-60 are restated to correct these items.
This change has no effect on safety nor any environmental impact.
Since this change is administrative in nature it need not be evaluated further.
Item 11 - In Amendment Nos 46 and 40 issued on March 2,1981 we issued TS in respo6se to the Three Mile Island Unit 2 accident assessment.
In Amendment Nos 47 and 41 issued April 1, 1981 we issued TS in response to the decay heat removal capability and depth of water over the reactor ve:7.el flange during refueling operations.
The TS which were added by the earlier amendments were,.in some instances, inadvertently deleted by the issaance of the lathr amendment.
The pages involved are TS3.1-2, TS3.1-3, TS3.1-3A and 4.1-2A.
These pages have been corrected.to be consistent with both amendments 46/40 and 47/41 issued earlier. Since this change is administrative in nature it need not be evaluated further.
Item 12 - By letter to the licensee dated October 29, 1979 the staff stated that it had recently completed a review of the LER's and TS requirements related to Centrol Rod Position Indication Systems (RPI) at Westinghouse PWR's and had determined that a wide variation existed in the number of LER's reviewed and the TS requirenents and had therefore decided to clarify the regulatory requirements. By letter dated February 20, 1980 the licensee responded to the staff's earlier letter with a proposed change to the.
PINGP TS.
Item 13 - In the Safety Evaluation Report accompanying the issuance of Amendment Nos. 39 and 33 on September 6,1979 the staff noted that following implementation of certain modifications of the fire protection system it would be necessary to further modify the TS to incorporate the LC0's and surveillance requirements for these modifications.
By letter dated July 31, 1980 the licensee submitted proposed changes to the TS for some of these modifications. The proposal also corrected certain typo-graphical errors.
Evaluation:
Dearaded Grid Voltace Conditions Item 1 - In response to the degraded grid voltage concern, the following design m difications and Technical Specification changes were proposed o
and implemented by NSP:
, a.
Installation of second level undervoltage relays, two on each of the four 4160v Class 1E buses with a drop out setting at approximately 90%
of nominal bus voltage and a six second time delay. These relays will be part of a modified two-out-of-four coincidence logic scheme. The same logic is used for the existing first level o f undervoitage protection.
1 The LLL initial evaluation was based on an.NSP preliminary setpoint I
proposal of 89% + 1%. Following further evsluation by NSP a setpoint of 90% + 2% was Kroposed in the TS submittal. This is a more conser-vative accounting for anticipated 'astrument drift and is acceptable, b.
Installation of circuitry to block the undervoltage trip load shedding feature on the 4160v Class 1E buses when the diesel generators are supplying these buses, and automatically reinstating this feature when the diesel generator breakers are tripped.
c.
Addition of trip setpoin limiting conditions for operation and l
surveillance requirements in the Technical Specifications as3ociated with the design, modi.fications cited above.
The criteria used by.LLL in its technical evaluation of the above proposed changes include GDC-17, " Electric Power Systems," of Appendix A to 10 CFR 50; IEEE Standard 279-1971, " Criteria for Protection Systems for Nuclear Power Generating Stations;" IEEE Standard 308-197&, " Class 1E Power Systems for I;uclear ' Power Generating Stations;" and the staff positions defined in
!;RC generic letter to i;SP dned June 3,1977.
We have reviewed the LLL Technical Evaluation Report and concur in its findings that (1) the proposed modifications will protect the Class 1E Equipment and systems from a sustained degraded voltage of the offsite power source, and (2) the proposed changes to the TS meet the criteria for periodic testing of protecticn systems and equipment.
Therefore, we conclude that NSP's proposed design modifications and changes to the TS are acceptable.
TS 4.4.B Emeroency Charcoal Filter Systems Item 2 - NSP proposes to change the quarterly surveillance test initiation signals for the containment shield building ventilation system and the auxiliary building special ventilation system from a simulated safety injection signal to a manual initiation signal.
The TS changes make the.
testing requirements for these two systems consistent with the quarterly testing requirements specified for other PINGP safety related systems.
The revised testing requirements have also been modified slightly by us with the licensee's concurrence such that they are also consistent with comparable testing requirements included in the Standard Technical Specifications. We have reviewed the proposed changes as modified and find that they are acceptable.
. Item 3 - TS 4.5. A.3, Containment Fan Coolers TS 4.5. A.3 now requires that containment cooling fan coil unit performance he determined to be within design specifications (i.e. performance require-ments consistent with the mode of operation following an accident) during each refueling.
The requested change would replace the requirement to determine that unit performance is within design specification with a requirement to demonstrate acceptable fan coil unit terminal coolant temperatures and coolant flowrate for containment atmospheric conditions normally experienced during power operations.
It is not possible to demonstrate fan coil unit heat removal capabilities consistent with the design specifications because of the much reduced thermal loading on the fan coil unit from the containment atmospheric conditions experienced during normal operating modes relative to the steam environment of elevated temperature and pressures associated with an accident. Sufficient assurance of the operability of the fan coil units will be provided by.the revised specification which requires the measurement of fan coil unit terminal temperatures and the measurement of fan coil ccolant flow rate for comparison against a specified valoe.
Based sn our review, the revised TS is consistent with the flRC staff Standarc fechnical Specifications and is acceptable.
Iten
- TS Table TS 3.12-1 " Safety Related Shock Suppressors (Snubbers)"
TS Table TS 3.12-1 now includes only shock suppressors for systems defined by tne FSAR analyses to be essential in the analysis of high energy line break events.
The requested change would expand the list of suppressors which must be determined to be operable to assure safe operation to include all suppressors required to protect systems required to safely shutdown and maintain the reactor in a safe shutdown condition. We have concluded that the proposed change increases the protection provided to safety related systems from seismic events and is acceptable.
Item 8 - TS Section 6.0, Administrative Controls We have evaluated the proposed changes in the licensee's corporate organizational relationship to the on-site operating organization and conclude that the changes are likely to provide more visability for the management of quality assurance activities and for the management of training activities and reflect a strengthening of the offsite support groups for the licensee's nuclear plants and are therefore acceptable.
Ws have evaluated the proposed changes in the licensee's on-site group and conclude that the changet are likely to provide more visability for the management of quality assurance activities. The proposed changes in titles for the same positions involve insignificant changes in responsibilities and functions.
Therefore, we find these changes acceptable.
. Item 9 - TS 1.I, Operable The guidance provided in the staff's letter of April 10, 1980 is based on the consideration that (a) the Limiting Conditions for Operation (LC0's) specified for the plant safety related systems address, in most cases, only single outages of component, trains or subsystems, and (b) for any particular system the LCO does not address multiple outages of redundant components nor does it address the effects of outages of any support system. Therefore the flRC staff's April 10, 1980 letter requested the licensee to incorporate two general types of specifications to assure that no set of equipment outages would be allowed to persist that would result in the facility being in an unprotected condition.
. One type of general specification specifies the corrective measures to be taken for circumstances in excess of those addressed in the specific system specification. Addition of this specification was intended to provide for the situation wher in a specific system specification addresses e
the action required to dpal with one inoperable component, train or sub-system but does not address the action required wherein further co::Gonents, trains or subsystems. redundant to the first.one above. are also inoperable.
The licensee addressed this subject further in a letter dated May 14, 1980 wherein he stated that since the PINGP TS have specific action statements and plant operating modes requirements integrated with the LC0's for each safety system, it would not be practical to include further STS requirements in this regard in the pit'GP TS. We have reviewed the pit'GP TS and find that for TS which address redundant components, trains or subsystems the TS require that all redundant corponents. trains or sub-systems (CTS) be operable except for a single CTS which may be inoperable for a specified time. The PIl!GP TS also explicitly require that the operability of the redundant CTS be demonstrated prior to initiating corrective action on the inoperable CTS.
Based on these considerations we conclude that the PIfiGP TS do not require further modifications to meet the objectives of the staff's model STS in this regard and are acceptable.
The second type of general specification addresses the situation for which a system would be declared inoperable solely because its emergency power source is inoperable. We have reviewed the licensee's proposed addition of general specifications in this regard end have determined that when implemented in conjunction with the overall PINGP TS they meet the objectives of t..e Model STS and are acceptable.
l 8-Item 12 - Technical Specification 3.10.E, Rod Misalignment Limitations Westinghouse has performed safety analyses for control rod misalignment up to 15 inches or 24 steps (one step equals 5/8 inch). Since analyses of misalignments in excess of this amount have not been submitted, we have determined that a LC0 restricting continued operation with a misalignment in excess of 15 inches is appropriate. Because tM analog control rod position indication system has an uncertainty of 7.5 inches (12 steps),
when an indicated deviation of 12 steps exists, the actual misalignment may be 15 inches. This is beca.use one of the coils, spaced at 3.75 inches, may be failed without the oparator's knowledge. The Standard Technical Specifications were written to eliminate any confusion about this, and restrict deviations to 12 indicated steps. Surveillance requirements, on the indication accuracy of 12 steps, were also prepared to ensure that the 15 inch LCO is met.
Since there is no difference intended in require-ments issued for any Westinghouse reactor, plants with Technical Specifications written in different terms of misalignment should consider the 12 step instrument inaccuracy when monitoring rod position.
A' related problem is that the installed analog control rod position indicating system equipment may not, in some areas, be adequate to maintain the control rod misalignment specification requirement because of drift problems in the calibration curves.
This is evidenced by numerous LER's concerning rod position indication accuracy.
In these cases, the uncertainty may be more than 12 steps.
The licensee was requested b" htter dated October 29, 1979 to review the TS for the PINGP to ens t the control rods are required to be maintained with ! 12 steps
-aced position and that the rod position indication system is accurate to within ! 12 steps.
By letter dated February 20, 1980, the licensee responded to the NRC request and provided proposed TS to address the staff's concerns.
The proposed change differs from the staff's model by allowing misalignments up to 15 inches (24 steps) when the rods are positioned less than 30 steps or greater than 215 steps.
Since the reactivity worths of cortrol rods at positions less than 30 steps or greater than 215 steps are sufficiently small that misalignments up to 15 inches (24 steps) will have no appreciable ef fect on in-core power distribution and since calibrations are performed for the normal operation region of between 20 and 210 steps, the staff finds this difference with the model TS to be acceptable.
Based on our review of the licensee's submittal, we find that the proposed changes are in conformance with the staff's request and are, therefore, acceptable.
g.
Item 13 - TS 3.14 and 4.16, Fire Detection and protection Systems With the issuance of Amendment Nos. 39 and 33 on Septembcr 6,1979 the staff stated in the accompanying Fire Protection Safety Evaluation Resort (FPSER) that following the implementation of certain modifications of the fire protection systems the Technical Specifications would be fur ther modified to incorporate the applicable LCO's and surveillance requirements for these modifications.
In a submittal dated July 31, 1920 the licensee proposed further modifications of this type. The licensee also proposed modifications to clarify the wording and to correct several typographical errors.
We find that the proposed specifications are in addition to fire protection system technical specifications previously established, will not adversely
- affect the effectiveness of the plant's fire protection program and are generally worded consistent with the Standard Technical Specifications.
On these bases we conclude that the proposed c'.inges are acceptaole.
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. Environmental Consideration We have determined that the amendments do not authorize a change in effluent types or total amounts nor an increase in power level and will not result !r. any significant environmental impact. Havinc made this determination, we have further concluoed that the amendments involve an action which is insignificant from the standpoint of environmental impact cnd, pursuant to 10 CFR SSI.5(d)(4), that an environmental impact statement or negative declaration and environ-mental impact appraisal need not be prepared in connection with the issuance of t.hese amendments.
Conclusion We have concluded, based on the considerations discussed above, that:
(1) because the amendments do not involve a significant increase in the probability or consequences.of accidents previously considered and.do not involve a significant decrease in a safety margin, the amendments do not involve a significant hazards consider.ation, (2) there is reasonable assurance that the health and safety of the public will noc be endangered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Commission's
' regulations and the issuance of these amendments will not be inimical to the ccamon defense and sccurity or to the health and safety of the public.
Date: July 28, 1931 Attachments to SER:
1.
References 4
2.
Technical Evaluation Report by Lawrence Livermore Laboratory, UCID-18654
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0 Attachment i e
hI' REFERENCES l
1.
URC description of' degraded grid voltage condision,-D..L'. Ziemann-to L. O. Mayer, August 13, 1976.
2.
NSP' response to degraded grid voltage condition, L. O. Mayer to D. L. Ziemann, September 20, 1976.
i 3.
NRC request on degraded grid voltage condition,0. K. Davis to L. O.
j Mayer, June 3, 1977.
t 4.
NSP response 6n degraded grid volta'ge condition, L. O. Mayer to Director, NRR, May 4, 1978.
i 5.
NRC request 6n degraded grid voltage. condition A. Schwencer to L. O. Mayer, July 30, 1979.
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6.
NSP response on' degraded grid voltage condition, L. O. Mayer to l
- Director, NRR, October 12, 1979.
l 7.
NSP submittal of proposed Te:hnical Specif' cation changes, L. O. Mayer i
to Director, NRR, May '.6, 1980.
I 8.
NSP submittal of Supplement No. 1 to prc;osed TS changes, L. O.
'ayer to Director,-NRR, October 8, 1980.
9.
NRC letter, D. G. Eisenhut, to all licensees dated April 10, 1950.
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