IR 05000255/1990020
| ML18057A838 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 03/29/1991 |
| From: | Lougheed V, John Monninger, Vanderniet C NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML18057A837 | List: |
| References | |
| 50-255-90-20, NUDOCS 9104090103 | |
| Download: ML18057A838 (34) | |
Text
U.S. NUCLEAR REGULATORY COMMISSION
REGION III
Report No. 50-255/90020 (DR_S)
Docket No. 50--255 Licensee:
Consumers Power Company l945 West Parnell Road *
Jackson, Mibhigan 49201 License No. DPR-20 Facility Name:
Palisades Nuclear Generating Plant *
Inspection At:
Palisades Site, Covert, Michigan Inspection Conducted:
September 4, 1990 throughMarch 28, Inspectors:.
. f J {~kl
- J VA./1.1 *
V ~ P. Lo~gheed ~
Dhte1
cJ-. v~
~.
- flhA1 *
J. D. Monninger.
Di!i.tfi Approved By: (j;i.i ~
- * *.
3{z_cif1t C. L. Vanderniet, Chief Ddter *
Operations Program Section Inspection Summary 1991 Inspection on September 4, 1990, through March 28, 1991 (Report No. 50-255/90020(DRS)
Area Inspected:
Special safety inspection by regional inspectors and staff members of the Off ice of Nuclear Reactor Regulation to review allegations concerning (1) potential post loss of coolant accident (LOCA) return to criticality due to insufficient boron in the safety injection tanks, 'and (2) potential post-LOCA hydrogen burn due to increased* amounts of aluminum inside the primary containmen Results:
No violations. were identifie Both allegations were partially_
substantiate Regarding the first issue, the NRC determined that a return to criticality could not be ruled out based on realistic assumptions; however, this return would not significantly impact the course of post LOCA recovery from a safety standpoin Regar~ing the second issue, the NRC determined that the maximum* hydrogen concentration values present in the FSAR were incorrectly low; however, the correct value would still be below flammability limit Two Unresolved Items were identified and are discussed in sections 2.a and 2.b of the repor ~AR4o901oa 910403 O
ADOCK 05000255 PDR
- .DETAILS Pers.ons Contacted Consumers Power Company
- K.* *J. *T. *W. *G. *G. R.* b. J.
Haas, Reactor Safety Development Manager Kuemm, Plant.Licensing Administrator Buczwinski, Reactor Engineering Superintendent Roberts,. Licensing Engineer..
Pratt, Senior Reactor Engineer Packard, General Reactor Engineer Gerling~ ~c*c*ident anQ.. T-ransient fu'lalysis Superv*isor Vandewalle, Director, Safety and Licensing
. -u. s. Nuclear Regulatory Commission
.*M. P. Phillips; Chief, Operations Branch
- C. L. Vanderniet, Chief, Operational Programs Section
- J. K. Heller, Senior Resident Inspector
+B. E. Holian, Project Manager, Office of Nuclear Reactor Regulation
- Denotes those attending the exit meeting on March 281 199 +Denotes those participating in tJ:ie.exit meeting by telephone on March 28, 19.9.
Allegation CRIII-90-A-0062)
On June 15, 1990, Region III received two allegations regarding the performance of the Palisades facility.
foliowing a large break loss of coolant accident (LOCA).
The first allegation involved the potential for a return to criticality during a LOCA due to insufficient.boron in the *
Safety Injection Tanks (SITs).
The second concern pertained to the amount of hydrogen generated during a LOCA, with an emphasis on the potential for a hydrogen bur On September 4, 1990, the inspectors met with the alleger and the alleger's attorne The purpose of this meeting was to obtain a detailed description of the allegations and to receive copies of the alleger's calculations supporting the allegation *
On September 6, 1990, the inspectors visited the Palisades site*to collect.information regarding the allegations, licensee calculations, and. other supporting'documentatio On September 20, 1990, all of the accumulated information was submitted to the Office of Nuclear Reactor Regulation
(NRR) for review and evaluatio NRR was requested.to
- provide a technical evaluation of the allegation During the course of NRR's review, both the alleger and the licensee werecontqcted to provide additional information when required. -The final NRR safety-evaluations are
- provided as Enclosure 2, "Safety Evaluation *on the Potentia-1 for a Retu,rn* 1:0. Criticality* Following a Large Break LOCA at Palisades," and Enclosure* 3, "Safety
.
Evaluation.Regarding the Post-LOCA Hydrogen Analysis."
The following paragraphs -separate the major allegations into their constituent parts in order to address. each specific concern expressed by the alleger.. At the,end of each.
section, 'an: overall conclusion of the major allegation is provided.* Post LOCA Return to.criticality (partially substantiated)
Synopsis of Allegation:
For a number of years Palisades Nuclear Plant has had problems with in-leakage of primary coolant system (PCS) water into the SITs due to leakage past the SIT check valve In the event of a LOCA,-this relatively unborated water would reach the core first,.and, combined with PCS water remaining in the core following_ the accident, result in a return to criticalit No credit.shouid be given for the control rods in this scenario, as they were never *
shown to insert following a LOC In addition, the licensee analyses incorporates improper assumptions which are designed to maximize the conditions for fuel temperature, but are improper to bound the*condition of
- return to criticalit NRC Review:
The individual portions of the issues discussed above are addressed as follows:
(1)
Allegation:
The SITs are the only method of shutting down the reactor following a. large LOCA, as the control rods were never proven to insert during the LOC NRC Review:
This portion of the allegation i~ a restatement of the licensee's assumptions contained in the Final Safety Analysis Report (FSAR).
As discussed in Enciosure 2, no credit*
was assumed, under worst case conditions, for control rod insertion during a large LOCA. *
Reactor shutdown was to be obtained by the infusion of boron from the SITs and the safety injection systems, which obtain water initially
' (2)
(3)
from the.borated safety injection refueling.water (SIRW) tan It should *pe noted, however, that Palisades is.not unique in its treatment of control rod insertion during a large LOC The* lack of credit for the control rods is a generic assumption at*
pressurized water reactors (PWRs) built by either
.Westinghouse or Combustion Engineerin Allegation:
At Palisades, borated water is delayed from reaching the core due to valve in~ieakage which allows PCS water to enter the tanks and fill the SIT injection line NRC Review:
This portion of* the al.legation was evaluated from the standpoint that if valve in-leakage were occurring, borated water would be delayed from reaching the core, and as such.was substantiate As stated in Enclosure 4, the SITs do not have recirculation capability and a large volume of water is* required to.be drawn from the tanks in order to achieve a representative sample..
Based on.this, the inspectors determined that some stratification was occurring in the SIT Therefore, the lower boron concentration PCS water would be injected first during th~ large break*
LOCA, with a gradual increase.up to the full boron concentration SIT.water~
Allegation:
Palisades has experienced problems with the SIT valves since at.least 1982, *and it has not been correcte This results in PCS in-leakage into the SI No solution has been*
fo.und nor any eval'uation. done of the impact on safety of the *plant in the event of an acciden NRC Review:
This portion of the allegation.was partially substantiate Palisades has experienced problems with the SIT valves and PCS in-leakage. _However, the licensee has developed an enhanced tracking, sampling, and maintenance program to ensure that technical specification (TS) requirements would be me In addition, the NRC performed an evaluation regarding the impact on safety of the plant in the event of an accident when SIT in-leakage were occurrin Based on reviews of plant records, including event reports, deviation reports, and work requests from 1982 until the present, as well as discussions with plant personnel, Palisades has had recurrent
-*
(4)
problems with valve leakage-on the SIT line However, the licensee has* an extensive tracking, sampling, and inaintenapce program to ensure that SIT leakage is* monitored and technical specification requirements maintaine As a result of these programs the licensee.has experienced periods of ti~e when.no in-leakage occurre An increased band for SIT level measurements was. requested by the licensee and approved by the NRC in TS amendment No. 13 An evaluation of the impact of valve leakage on-plant safety was performed by the NRC after the allegations had been made, and is discµssed in Enclosure Allegation:
The licensee is resporisible*to provide. calculations for the_ -worst possible condition accident scenario to ensure that the -
plant would be safely shutdown during such a postulat_ed acciden This was, riot done at-Palisades and, although-the problem existed for a-very long time, it was never pr9perly addresse NRC Review:
This portion of the_allegation was not substantiate CFR 50.46and 10 CFR Part so*, Appendix K, impose certain requirements* for a licensee to use when performing large LOCA accident analyse The licensee correctly followed these reqi.lirements in the performance of its analyse *
These assumptions, while conservative to* show that long term core integrity could be maintained, do not address the alleger's concerns of a return to criticalit However, as noted in enclosure 2, a return to criticality does not necessarily impose additional safety_concerns on the plan The -
potential for Palisades to return to criticality subsequent to a large break LOCA would not result in prompt criticality due to the effect of void distribution expected during reflood and a short-term return to criticality was not likely to significantly impact the_course of LOCA-recovery because of the negative feedback of effects of voiding, the cooling effect of increased steaming, and the imminent shutdown from continued injection-of high boron concentration ECCS wate Therefore, the return to criticality is not considered a significant safety concern or the worst possible accident scenario *
..~-,
(5)
Allegation:
The basic LoCA analysis was performed
. to magnify fuel temperature and its assumptions in
.the.model do not realistically show the actual amount of.PCS water left in the reactor vessel following the blowdown phase of a*r.ocA, because this water is superf iciall_y subtracted.in the
model. to maximize the temperature o*f the fue Also other numerous col)servative assumptions.for maximizing fuel temperature are done, which are not conserV~tive for calculations of boron in the core following*a large brea)c LOC NRC Review: _* This portion of the allegation was substantiated.* The Palisades design basie;.. LOCA.
analysis correctly complied with the requirements of 10 CFR 50.. 46 and 10 CFRPart 50, Appendix These regulatory requirements, which are concerned with peak cladding temperature, iinposed *
conservative assumptions to maximize the likelihood of core.damage in order to show that long term core integrity could be maintaine As noted above, the independent NRC evalua~ion concluded that a return to criticality could not be ruled out, but that such an event would not significantly impact the course of LOCA recover (6). Allegation:* The Palisades large break LOCA analysis did not account for dilution of the Safety Injection Line (7)
-NRC Review:
This portion bf the *allegation was substantiate As.discussed in item (5) above,
.
the licensee's design basis LOCA analysis complied with 10.CFR 50.46 and 10 CFR Part 50, Appendix K, in showing that long term core integrity was maintaine The-se calculations. did not* address the possibilit~ of a return t6.criticality due to insufficient boron being injecte The design
.basis analyses also did*not address the fact that the SIT valve leakage would have caused a large quantity of water at PCS boron concentration to be injected prior to the higher SIT boron concentration The failure to address these concerns in the design_ basis LOCA analysis was* not contrary to regulatory requirements for the analysis, and a realistic boron concentration evaluation, as stated above, did not identify a significant safety concer Alleaation:
The beginning of cycle (BOC)
condition, which is the most limiting, was not considered in the 1982 basis analysis.
(8)
(9)
NRC Review:
This portion of the allegation was substantiate The 1982 analysis was spec~fically performed for the situation.occurring at th~ end of cycle 5, which was an end of cycle (EOC)
conditio However, at the time this analysis was prepared and approved, it was not intended to be applied on a generic basis, but only for the conclusion of cycle A similar analysis was submitted to the NRC to support a TS change request in 199 The NRC never acted on that request or evaluated the licens~e's analysi The request was subsequently withdrawn by the licensee. *
Alleqation:
The critical boron concentration neutronic calculations used interpolations between the transient analysis value This neglected the
'effects of the diluted water in the SIT lines due to valve leakage, which directly affects the outcome of the neutronic calculation NRC Review:
As noted iri the review of sub-allegations 5 and 6 above, the NRC concluded that the licensee's return to criticality analysis continued to use design basis assumptions, which the NRC noted were non-conservative from a criticality perspectiv Therefore, this portion of the allegation was substantiate However, the licensee's analyses were performed in accordance with NRC guidance to address the requirements of 10 CFR 50.46 and 10 CFR Part 50, Appendix Allegation:
During the years, alternatively different tanks (two at a time) had in-leakage problem Therefore, it is conceivable that three or four of the SITs could experience in-leakage at the same.tim NRC Review:
This concern was partially substantiate A review of Palisades records, and conversations with piant personnel confirmed that tank in-leakage has previously occurred on more than one tank at a tim Additionally, because of common'lines between the tanks, the lines lea~ing to all four tanks could be at PCS concentratio However, based on past history and the licensee's proactive efforts to monitor and correct valve leakages, the number and extent of tank leak*age is decreasin Therefore, it is considered unlikely that in-leakage into all four tanks would occur simultaneously now or in the future.
Overall alleaation conclusion:
The allegation was partially substantiate The alleger's concern that the licensee's.SIT boron concentration may not prevent a return to criticality could not be ruled out.*
However, as stated in Enclosure.2, the decision on the
- validity of the allegation does* not affect the *
continued operation, or overall safety of the Palisades plant -because. of the fallowing considerations:. ( 1) the initiating event is a low probability event; (2) the return to criticality requires a failure of all o.f the control rods to insert prior to reflood; (3) a short term return.to non-prompt criticality will not
significantly impact* the course of LOCA. recovery because of the negative feedback *cf effect*s of voiding and the cooling effect of increased steaming, and the imminent shutdown from continued injection of high boron concentration ECCS water;* (4) a prompt critical condition is unlikely due to the effect of the void distribution expected during reflood; and (5) the
licensee's analyses were performed consistent with the requirements of 10 CFR 50.46 and 10 CFE Part 50, Appendix K..
The inspectors have identified concerns with the licensee's design control processes and co~rective actions processe The licensee's Quality Assurance (QA) Program, as described in Topical Report CPC-2A, Revis.ion 10, titted "Quality Assurance Program*
Description for Operational Nuclear Power Pl~nts" specifies the method by which the.licensee implemented the requirements of 10 CFR Part 50, Appendix.Bat the time these issues were identified to the license Paragraph 3.2.'2 of this topical report specifies that
"Errors and deficiencies *in approved design documents, or in design methods (such as computer codes) that could adversely affect structures, systems, and *
components are documente Action is taken to assure that errors and deficiencies are corrected."
. Paragraph 16. 2.*3 of* the topical report specifies that
"For significant conditions adverse to quality, necessary corrective action is promptly determined and recorde corrective action includes determining the cause and extent of the condition, and taking appropriate action to preclude similar problems in the future."*
These concerns will be tracked as an Unresolved Item (255/90020-01 (DRS)), pending further re.view by the NRC staff.
.*
b.
Post-LOCA Hydrogen Generation (partially substantiated)
Synopsis of Allegation:
The lic~nSee severely underestim.ated the amount of aluminum in containment, arid the analyses performed to support Chapter 14.22 of the FSAR was incorrect and contained several error This*could result in the hydrogen conceritration exceeding the 4.1% flammability limit, which would be catastrophi *
NRC Review:
The individual portions of.the issues discussed abov.e are addressed as follows: *
,., \\
~.1. J Allegation:
The.am.cunt cf aluminum i~sµlation inside co~tainment was incorrectly lo NRC Review:
This portion of the allegation was substantiate The alleger calculated ?tn area of 18l,613 square feet of aluminum inside containmen The FSAR reports an area.of 152,462 square fee: The licensee, in its independent review of the alleger's calculation, corrected the FSAR value with two increases:
one of
approximately 27,000 square feet, and one of 11,000 square fee These increases resulted iri a total amount of. aluminum of approximately 190,000 square fee This was larger than the value
.
calculated by the allege However, during the 1990 steam generator replacement outage, much o the aluminum insulation on the steam system piping was replaced with a different (fiberglass) typ This would lower the overall amount of aluminum inside containmen Because the licensee agreed
.that the FSAR value was incorrect, and no accurate total was available, the alleger's calctilated amount of aluminum was used in the NRC calculations, as documented in Enclosure The addition of aluminum or zinc to containment during modifications would have been performed utilizing the requirements of 10 CFR 50.5 These requirements specify that an evaluation be conducted to determine, among other things,
- whether an increase in the probability or consequences of an analyzed accident could occu In this case, the increase in aluminum insulation should have been evaluated to determine its impact on maximum hydrogen concentratio The NRC did not review modifications that would have resulted in changes to the *amount of aluminum or zinc to containment to determine if the associated safety
(2)
(3) !
(4)
evaluations were acceptabl This determination is considered an Unresolved*Item (255/90020-02(DRS)).
Allegation:
No proper correction was administered to the corrosion rates at the beginning of the corrosion process, which were assumed to be* flat *
over.a long period.of time, but.in fact the*
.corrosion rates are very high initially and are changing fast subsequentiall NRC Review:
This portion*of the allegation was partially substantia:ted *.
As stated in.Enclosure 3, the FSAR underpredicted the amounts of aluminum corroded in the first 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />, but after that time, the FSAR values* were more conservative' than those based on time dependent.*rate However, the NRR analyses used the alleger's corrosion rates to calculate the maximum hydrogen.concentration Allegation:
The hydrogen from zinc and galvanized surfaces reaction with water is not fully accounted for in the total hydrogen productio NRC Review:
This portion of the allegation was
,partially substantiate NRR, as documented in Enclosure 3, noted* that the values of the corrosion.of zinc from galvanized surfaces was identical to that given in ORNL-TM-2412, "Design Consideration of Reactor Containment Spray System Part III The Corrosion of Materials in
.Spray Solutions."
NRR also found the corrosion rates of zinc paints used by both the licensee and the alleger to be less than that from recent test program In performing its independent analyses, NRR used the alleger's values and. concluded that, overall, the hydrogen production rate from zinc paints and galvanized surfaces was a relatively insignificant term in the overall'tota Allegation:
The partial pressures and temperatures correction of containment atmosphere do not seem to be included [in the FSAR analysis].
Additionally, no correction on the recombiner intake as a function of temperature and pressure of the containment seems to be include NRC Review: *This portion of the allegation was partially substantiate NRC noted, in Enclosure 3, that the original FSAR analysis was probably outdate However, the licensee's independent review of the alleger's concern used
. ( 5}
(6)
(7)
the COGAP computer program (NUREG/CR-2847., "_COGAP:
A Nuclear Power Plant Containment Hydrogen Control System Evaluation Code"} which properly accounted for containment partial pressures and
temperature The COGAP program, which was
. approved by NRC, was the one also used in the NRC's independent analysis *.
Alleoation:
The analysis should correct for the sprayed volume of containment, which constitutes using the entire volume).
NRC Review:
This portion of the allegation was not substantiate In Enclosure 3, NP~ concluded that the entire free volume of the containment should be used because of the turbulent mixing generated by the break flow jets, containment sprays, and natural convection flow Alleoation:
The differential equation governing production and consumption of hydrogen following the recombiners initiation was never solved and
.the COGAP program was riot utilized eithe NRC Review:
This portion of the allegation was partially substantiate As stated in item (4)
above, and _in Enclosure 3, the original FSAR solution may have been incorrec However, the licensee's independent review *analysis did use the COGAP computer program, which does properly solve the differential equation regarding production and consumption of hydroge *
Alleoation:
Most important is the fact that the Palisades FSAR Section 14.22 allows the plant to reach the control limit of 3.6 volume percent (v/o} hydrogen before initiation of recombiner operatio Because the recombiner, by its nature o~ small hydrogen.intake, does not reduce
_immediately *the hydrogen content of the containment atmosphere, its action is not sufficient to bring the hydrogen.concentration down immediately, and, in fact, the concentration would be still rising to the level much higher than 4.1% flammability limit because of the continuous production of hydrogen from all the other source This, obviously, could be catastrophi NRC Review:
This portion of the allegation was partially substantiate NRR, in Enclosure 3, agreed that the FSAR did state that the hydrogen
- recombiner did not need to be actuated until a
- 1 imi t o.f 3. 6 v/ o hydrogen was reache *However,.
this was contradicted.by the.plant's emergency
operating procedures (EOP) which requir*e one
.. recombiner * to be started whenever containment pressure exceeds 3.7 psi Containment pressure would reach 3._7 psig within: a,few"seconds
.
following a large LOCA.,The staff concluded that the FSAR did requi:re updating in*order to reflect the EOP guidanc~, but that the recombiners would be started within a reasonable time frame following a LOC Furthermore, the independent NRC calculation of post-LdCA *hydrogen *
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.L.l.;tU.*~':flC.11 VQ.L of 2. 9% occurring in approximately 8 days. * * This is f:>Ufficiently below the_ flammability limi.
.
Overall alleoation conclusion:
This*allegation was partially substantiate The staff agreed that the.*
licensee's FSAR.Section 14.22 was outdated and needed to be revised, and noted that the licensee had_ been in the process of updating this section of the FSAR prior to the time-the *allegation was received by the NRc.*
- An.independent analysis of the post-LOCA:hydrogen concentration calculated using the NRC approved COGAP computer program showed the maximum hydrogen concentra ti oh to be 2. 9 % peaking. within 8 * day.
Therefore, the alleger's conclusion that.hydroge flammability limits would be reached, with the potential for a hydrogen burn, was not substantiate Review-of Additional Information The NRC.staff has factored any and all additional technical information received by March 22, 1991, from either the*
alleger or licensee, into its final conclusion Any other cqncerns expressed by the alleger are *undergoing further
review, and will be addressed in separate.correspondenc.
Unresolved Items Unresolved.items are matters about. which more information is required in order to ascertain whether they are acceptable items, items of non-compliance~ or deviation Unresolved items disclosed during the inspection are discussed in Paragraphs 2.a. and 2.b.(l) of this repor Exit Interview Discussions were held with both licensee representatives and the alleger by NRC staff throughout-the inspectio An exit
meeting-was held at the site on March 28, 199 The inspectors summarized the scope and findings in regard to the allegation The inspectors also disctissed the likely informational content of the inspection report~
The licensee did not identify any sucn content as proprietar )
.
13.
UNITED STATES NUCLEAR REGULATORV COMMISSION
- WASHINGTON, D. C. 2055 Encl asure 2
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SAFETY EVALUATION. BY THE OFFICE OF NUCLEAR REACTO.R REGULATION..
INTRODUCTION
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- RELATING TO AN ALLEGATION REGARDING* THE POTENTIAL FOR
. A RETURN TO CRITICALITY FOLLOWING A
. LARGE BREAK LOCA AT PALISADES
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CONSUMERS POWER COMPANY PALISADES NUCLEAR PLANT DOCKET N ~255 By letter of September 20, 1990 (Ref.- 1), Region III requested NRR to assist in re~iew of an allegation regarding LOCA analyses for Palisade In References 1 through 3, the alleger indicated that the assumptions which are intended to assure a conservative estimate* of f 1V,el cladding temperature are, not cons_ervative with respect to the potential for post-LOCA criticalit We have reviewed the allegation based on the information in References 3 and 4 and additional subm.ittals provided by the alleger during the period of October 1990 to January 199 We evaluated the information provided by the licensee in conference calls.
and follow-up submittals to the NR In addition, we have examined LOCA test data to assist in review of the allegatio As a result, we have prepared th~
following evaluatio.0. EVALUATION
The Palisades LOCA analyses are performed using analysis assumptions specified in Appendix K t6 10 CFR 50. These assumptions are defined with the deliberate intention of assuring a conservative calculation of peak cladding temperatur Some of the more important assumptions in this regard include:
(1) use of the Moody model to maximize the blowdown rate, (2) loss of all safety injection.
water to the containment during the bypass period, and (3) an assumed loss of offsite power-with consideration of single failure to minimize the safety injection flo The effect of these assumptions is to minimize the post-blowdown water inventory in the reactor pressure *vessel and thereby prolong the time period during which the core is uncovered. This results in a conservatively high estimate of peak cladding temperature. It also results in the core being reflooded with water only from the highly borated safety injection tanks (SIT) or safety injection system (SIS).
Reactor shutdown is easily achieved under these assumptions (because of the combination of boron and voiding).
No credit is assumed for control rod insertio *910403 SDR ADDCK 05000255 PDR
-2-From a reactivity perspective however, :these same assumptions estimate a less reactive state tha*n would be expected in a more real'i.stic scenari In a
- _realistic scenario significantly_inore primary system water would remain*in the
- vessel following the vessel biowdown. * This water would be cool (relative to*
- normal operating temperature) and would be at the full power boron concentratio.0 This residual water would dilute (in boron concentration) the incoming safety injection water and thereby si'gnificantly alter (in the nonconservative
.
- direction) the reactivity potential of the mixtur In addition, water remaining ih the v~ssel will result in an earlier start of_reflooding since le~s injection water is needed to fi 11 the 1 ower p 1 enu Th1s means that fuel rod temperatures will be )ower (than in the _LOCA analyses) and.will result in les.s boil off
.
during the initial refl6od *st~ge Since boil off leads to an intrease in b6ron concentration of ~he remaining coolant and increas.ed voiding, this too is nonconservative from a reactivity perspectiv The more low boron concentration water *remaining in the reactor vessel the more likely that reflood will ~esult
. in a critical cor Thus, the amount of the. remaining water in the reactor vessel has ~ significant effect on th~ reactor remaining shutdown following a large break LOC *
In reviewing the allegation, we have examined the licensee 1 s post-LOCA criticality analyses documented in References 3 and 5, and additional calculations of October 25~ October 13, and November 27, 199 In these analys~s the licensee has attempted to de~onstr~te that the boron concentration in the vessel water mixture at th~ time of reflood.is greater than the critical boron concentratio Our effort i'ncluded a review of the Palisades plant *primary coolant system (PCS) and ECCS volumes and ~onfigurat1on, model and analysis
assumptions, system respons We also examined the amount of reactivity which cou 1 d be added t.o the sys tern by the cool down of the coo 1 ant during *
depressurizatio The review found that a corildown of 300°F provides the potent i a 1 for a very s i gni fi cant reactivity -Insert ion (severa 1 percent).
We.
also determi.ned that the li.censee 1 s conclusions regarding recriticality depend upon assumptions and outputs from the licensing basis LOCA calculation As discussed above, design basis LOCA analyses minimize the amount of low boron concentration water remaining in the vessel and are therefore nontoni~rvative from a criticality perspectiv Finally, we examined experimental data pertinent to the questio LOFT test
- Ll-2 described in Reference 6 is a large break blowdown test conducted from typical PWR initial temperature and pressur The test result shows that water equivalent to 90 percent of wate_r volume of the lower plenum rem-ains in the reactor* vessel after the blowdow While the applicability of this data to Palisades may.be debated, it nonetheless is a data point based upon simulated PWR conditions.* For a. smaller break LOCA (but large enough to discredit control rod insertion for reactor shutdown) it may be expected-that more water could remain in the reactor vessel after the blOwdown because of a smaller depressutization rat DISCUSSION Based on the above considerations it is the staff 1 s position that a return to critical cannot be exclude However, it is expected that only a prompt critical configuration would be of' concern from a safety perspectiv This is considered unlikely because of the need fo~ a very rapid reactivity insertio **
4.. 0
-3-Experimental data {Ref. 7) indicates that during reflood when t~e coolant comes in qrntact with hot fuel rods, significant voiding occurs~ This has.the beneficial effect of introducing negative reactivity.from the increase in voids, as well as reducing the rate of reflooding (and thus the rate of addition of positive reactivity) due to*the increased resistance of the two phase mixture..
A non-prompt return to critical would not be expected to have significant effect *
on the pl ant recovery fo Tl owing a LOCA because.these same negative feedbacks would limit the power level, and increased steaming would assist in cooling those portions of the core which are.uncovere In addi_tion, the return to power would be short-lived because of the continued boron injection from the*
ECCS.. These factors, in conjunction with the low likelihood.of the initiating event (large break*LOCA), and the consideration that the control rods must also fail to irisert lead to the conclusion th~t continued operation i~ acceptabl CONCLUSION Based on our review of the ~lleger's submittals, the licensee's submittal*s, ~nd the test data discussed in Section 2.0, we conclude that a return to criticalit cannot be exclude This conclusion is based upon the following:
(1) there is the potential for a significant reactivity insertion due to cooldown of the moderator during depressurization; (2) test dat,a indicate that large amounts of (low boron concentration) primary system water may remain following the blowdown
- of the primary system;. and (3)' the licensee's criticality analyses use design basis LOCA models and assumptions.'V(,hich minimize the amount of low boro*n concentration water involved in the core reflood and are thus non-conservative relative to a return to criticality~ However, our decision on the validity of the.allegation in this SER. does not affect the continued operation ot the Palisades plan This i~ based on the following consideration:
(1)
.(2)
(3)
. (4)
(5)
The initiating event (large break LOCA) fs a low probability event; A return to criticality require.s failure of the control rods to insert prior to refl6od;
....
- A short term return-to non-prompt criticality is not likely to significantly impact the course of LOCA recovery _because of the negative feedback of effects of voi d.i ng and the cooling effect of i ncreas'ed steaming (if a notable power generation level is achieved), and the*
imminent shutdown from continued injection of high boron concentration ECCS water; A prompt critical condition is unlikely due to the effect of the void*
distribution expected during refloo The 1 i censee' s analyses have been performed consistent with the*
requirements of 10 CFR 50.46 and Appendix K to 10 CFR 5.0 REFERENCES
- Letter from H. J. Miller~ to Dennis M. Crutchfield, "Request for Assistance Allegations Regarding the Palisades Nu~lear Plant, dated September 20, 199 *
-4-Attachment 1 tif Reference 1 - Transcript of September 4~ 1990 meeting between-th~ alleger, Region III, and NRR (via telephone. ~
Attachment 2 of Reference 1 - Alleger File on Safety Injection Tank Boron Concentratio~.
.
.
4 *. Attachment 5 of Reference 1 - Consumers Power Company letter dated.
August 1, 1988 to the NRC, *concerning a Tech Spec Change Request for Modification.of Peaking Factors and LOCA Lim.it *
.
.
5~
Attachment 8 of Reference 1 - s*afety Injection Tank Boron Analysis from 1982, 1985, 1986, and the 1982 licensing submitta.
iREE-NUREG-1026, "Experimental Data Report for LOFT Nonnuc1ear Test*
.LI-2, 11 January 197.
. JAERI-M-90-236, "Evaluation Report on SCTF Co~e-II Test 52-08 (Effect bf Core Inlet Subcooling on Thermal-Hydraulic Behavior Includ.ing Two-Dimensiona*1 Behavior in P*ressure Vessel During Reflood in PWR-LOCA),
.
January 199L
".:-.*
P_ri nci pal Contributor:
M. Chatterton *
S. Sun
"*
. ~.
Enclosure 3
. SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULAT'rON *
. PLANT SYSTEMS BRANCH..
1.0 *INTRODUCTION POST-LOCA HYDROGEN ANALYSIS PALISADES NUCLEAR PLANT DOCKET NO. 50-255 By memorandum of September 20,_1990 (Reference 1) Region III requested NRR 1 s assistance in.responding to an allegafion regarding the Palisades Nucle~r Pl~n The Plant Systenis Branch of NRR has reviewed the allegation concerning the hydrogen generation fo 11 owi Jig a Loss of Co.ol ant Accident ( LOCA).
The Materials.*
and Chemical E~gineering Branch has provided input (Reference 9} on the-corrosion issu The alleger, who was employed as a design engineer in the transient analysis group for the facility, indicated that the hydrogen analysis in Section 14.22 of Palisades FSAR has many errors, and specifically, identified the following problem areas:
L... *
"t\\,
amount of aluminum insulation, corrosion rates,
.
hydrogen generation from zinc and galvanized surfaces, correction of partial pressures and temperature in containment atmosphere, corre~tion of free volume from effective spray volume, correct i_on of recombi ner intake as a function of temperature and pressure, initiation time of hydrogen recombiner, and improper solution for the differential equation governing production and consumption of. hydrogen fo 11 owi hg the initiation of recombiner....
Particularly, the alleger has a grave concern on the consequences of Item No. 7; the word 11catastrophic 11 was used by the alleger concerning this issue. _The above allegations were mainly based on the results of a calculation performed by the alleger for the facility and documented in Reference The original FSAR analysis was performed in 197 Our initial review indicated that some of the allegations might be valid and could have some impact on plant safet To evaluate the significance of the impact, we req1,.1ested our consultant at Sandia National Laboratories to perform an independent analysis using the computer code, COGA The COGAP code, which is described in Standard Review Plan Section 6.2.5, was developed and has been*
used by the NRC staff to calculate hydrogen concentration following a LOC The results of Sandia's analysis show that a maximum hydrogen concentration of 2.9 volume percent (v/o) will occur 8 days into a LOC PDR ADOCK 05000255 Q
"
-2-2. 0 EVALUATION The.staff reviewed each of the problem areas identified by the alleger and our evaluation is provided below.. The review was based on Regulatory *Guide (RG)
L 7 and Standard Review Plan. (SRP) 6.2.5. *The basic safety concern is whett\\er the containment.hydrogen concentration could exceed 4.0 v/o following a LOC In addition to the review, the staff's consultant at Sandia performed an analy~is to evaluate the overall safety significance of the above al legations. * To prepare a set of cqnservative input for the Sandia's COGAP analysis in response
- to.the allegations, the staff carefu.lly evaluated the FSAR,data, the alleger 1 s
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2.1 The amount of alumfoum insu.lation The total ~luminum surface area documented in Table 14.22-4 of the FSAR is
~52,462 ft. 2The aluminum surface area, calc~lated by the alleger in Reference 2, 1s 183,613 ft.
App~rently,- the number used rn the FSAR is not c9nservative. *
The staff called the licensee on November 27, 1990, to verify the dat The-licensee was aware that the FSAR might need changes and was in the process o re-evaluating the FSAR analysis.. Before the completion of the re-evaluation, the licensee performed a review aQ,alysis (Attachment 15 to Reference 1) 'using the computer. code COGAP and input data from the allege The results of this
. analysis indicated that the maximum hydrogen concentration was less than 4 v/o, but was much higher than was documented in the FSA The licensee used
.this analysis to demonstrate that there had not been any immediate safety concer However, the licensee had not decided whether all the assumptions used by the alleger were correct and whether the FSAR needed to be revise The licensee originally planned to complete its reanalysis by the end of 1990, but the schedule was slipped to the time prior to plant.restart'.
The COGAP computer code, which is described in SRP 6.2.5, has been us-ed*bythe staff to calculate hydrogen concentration following a LOC Therefore, it is acceptabl In reviewing the licensee's COGAP analysis (Attachment 15 to
.
Reference 1), however, we found that several input parameters used by the license were inconsistent with RG 1. Therefore, the staff requested Sandia to...
perform an independent analysi Since the licensee did not provide-any *
additional data on the amount of aluminum, the conservative number (183,613 ft )
from the a1leger was used as an input to Sandia 1 s analysi.2~ Corrosion rate of aluminum Based on Reference 3, the a1leger pointed out that the corrosion rates should be a function of time and that the FSAR assumed the rates to be constant for a long period of tim The most significant factors affecting corrosion rate of aluminum are time, temperature, and pH valu Review of Reference 3 has indicated that there are experimental data on the growth of aluminum corrosion film with tim Further, aluminum corrosion rate is a very sensitive function of temperature and pH
'"
-3-
- values. *.The pH value of the Chemically treated containment bui.lding spray in the Palisades plant is around 7, which is consistent with.the value used in the FSA The corrosion rates in. the fSAR are tempera tu.re dependent, but only three temperature points are. g1ve The staff compared the aluminum corrosion rates between the FSAR and Reference 3..
Further, using the 'temperature values as a function of time, we calculated the.
amount of aluminum corroded as a function of.time. *The results shown in Figure 1 indicate that for the first 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />.. after an accident, the FSAR methodology underpredicts the amounts of alumi'num corroded, but: after that tim~ the FSAR results are.more.conservative relative to the results based on the time deoehdent corrosion rates. of Reference I
.
.
,
In light of the time (8 days) that the maxfmum hyd.rogen concentration occurs,
- .. the FSAR corrosion rates for aluminum probably generate *more _hydrogen than the alleger 1 s corrosion rat~s do~ However, to evaluate the fm~act of the al~egation~
the alleger 1 s alum~num corrosio~ rates (Reference 2) ~ere used as the input to Sandia 1 s analysi In Attachment 15 to Reference 1, the licensee converted the al)eger 1 s corrosion ~ates into the proper units for the COGAP inpu.3 Corrosion of zinc In the Palisades plant zinc can be found on galvanized steei surfaces and in paint Similar to aluminum, \\~e rates of corrosion of ziric depends on teinperature and pH v.al ues.
a)
Corrosion rate of zinc on galvanized surface b)
The data used in the FSAR analysis are identical.to the data in Reference~-
To evaluate the impact of.the allegation, the staff-chose to use the alleger 1 s corrosion rates converted to the proper units (Attachment 15 to Reference 1) as input:data to Sandia's analysi~.
However, after Sandia 1's analysis was completed, the staff discovered an add~tional source of information (Figure 2) on corrosion of zinc on galvanized surfaces from an experimental program performed at Sandia Nati~nal Laboratories (Reference 4).
The staff obtained and evAluated the information and made an engineering judgment that the effect is*
insignificant to its conclusio A reanalysis is not necessary since the time dependent hydrogen generation from zinc, as shown in Reference 2, is a sma 11 fraction of the total, hydrogen generatio Corrosion rate of zinc in paints The information on corrosion of zinc paints comes from a test program performed at Sandia National Laboratories (Reference 7).
The corrosion rates are a function of temperature and pH values. *As shown in Figure 3, the corrosion rates from Reference 7 are significantly higher than t~e constant corrosion rate used in the FSAR, which is test data obtained by the license The large difference is probably from the difference in
.
test conditions; the tests reported in Reference 7 were run at conditions
-4-more severe than those in the,licensee'.s "tes The corrosion rates used by the alleger appear to be between FSAR and.Reference For the same reason discussed above (this_is a relatively insignificant term), the corrosfon rates*of zinc in paints usedin*Sandia's analysis were.taken
- from*.the licensee 1 s revfe\\'f ana lys*i s (Attachment 15 to Refe.rehce 1), wh.i ch
~ere the Same values used for the torrosion rates for zinc on gal~anized surface * 2.4 The effect of partial pressures. and temperature in containment atmosphere and the* effect of pressure and temperature at*. the recombi ner intake The original FSAR analysis was perfofmed in 1976, and the method of analysi~ is probably outdate However, the temperature and pressure effects are included in the COGAP cod The licensee has been using the COGAP code to perform its review analysis~ Therefore, Sandia's analysis and the licensee's review analysis respond the. alleger's-concern in this are However, the FSAR should be revised to incorporate these effects~
2.5 Correction of free volume from effective spray volume The alleger assumed 95 percent of the free volume to be effective for hydroge mixing* because the containment spray covered only 90 percent of the free
- volum The staff does not believe that* this assumption is vali The total free volume should be used becaus~. the turbulent mixing generated by the break
- flow jets, containment sprays, and natural convection flows will assure the hydrogen is we 11 mixed within the entire containment free vo 1 ume, including that outside of the spray volum.6 Initi~tion time of riydrogen recombiner The alleger stated that "Most. important is the fact that the Palisades FSAR Section 14.22 allows plant to reach the control limit of 3.6 v/o hydrogen before initiation of recombiner operation. *Because the recombiner by_ its
- nature of small hydrogen intake does not reduce immediately the hydrogen content of the containment atmosphere, its action is not sufficient to bring the hydrogen concentration down immediately, and in fact the concentration would be still rising to the level much higher than 4.1 % flammability limit because of the continuous production of hydrogen from all the other source This obviously could be catastrophic.
Furthermore,.the alleger infers that initiation of the recombiners might not occur until approximately 800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> following the LOC The staff reviewed the statement in the FSAR and got a somewhat different impression that the recombiner could initiate as late as the time when hydrogen concentration
- reached the limit of 3.6 v/o, not at-800 hour Additionally, the staff reviewed the Palisades Emergency Operating Procedures(EOPs) (Attachme.nt 17'to Reference 1). It is stated in the EOPs that at least one hydrogen recombiner should be in operation if containment pressure exceeds 3.7 psi In a large LOCA the containment pressure would reach 3.7 psig in a few second Additionally, the procedures require operators to perform checks of safety system status using the Safety Function Status Check Sheet on a continuous basis at approximately 15 minutes interval The licensee's review analysis
-5-(Attachment 15 to.Reference 1).assumed: recombiner initiation time of either *
. 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or24 hour In either cas.e, the cakulated. maximum hydrogen concentration was less than 4 v/ The staff believes 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> is a proper assumption* arid ch6se it aithe inpu~ ~o~ Sandia's analys~ Based on our review of the Pa.li sades FSAR and EOPs, the staff determined that the EOPs provide assurance that the hydrogen concentration will not exceed the flammable limit of 4 v/o hydrogen.following a LOC Furthermore 4 v/o hydrogen i~ a lower flammable limit; it is not th~ limit to reach ~ ~atastrophic detoriati6 However, the FSAR should be r~vised to be consistent with the EOP~
for the initiation: time of recombiner Otherwise, the FSAR could mislead the operators and could introduce undesirable risk.7 Improper solution of the differential equation The alleger indicated that the diffe~ential equation governing the production and consumption' of hydrogen following recombiner initiatfon was not solved
. properly in the FSAR analysi Similar to the above discussion in Section 2.4,
. the COGAP code solves the equation properly. *Therefore, the Sandi a analysis and the licens~e 1 s review analysis respond to the alleger's concern in this are The. FSAR should b~ revised t6 reflect the pro~er solution of the differential
equ~tion.0 COGAP Calculation The licensee has performed a COGAP calculation using the alleger 1 s numbers from Reference However, in reviewing the licensee's COGAP calculation and the alleger 1 s analysis, the staff identified several input data that are not consistent with RG 1.7. Our consultant at Sandia National Laboratories performed COGAP calculations with parametric studies and foe.ind that the impact from those deviations could be significan Initially, Sandia's analysis confirmed the licensee's results by using the same set of input as shown in Attachment 15 to Reference Finally, Sandia made pertinent input c.h_anges that will be discussed in the following section Percent of Zirconium Reacted:. A value of 2.0 % was used by the alleger and t.he FSA However, a val1,1e of 1.0 % was used by Sandia based on Table 1 of RG 1. 7, which states tha hydrogen production is 5 times the extent of the maximum calculated reaction under 10 CFR Part 50, Section 50.46, or that amount that would be evolved from a core-wide average depth of reaction into the original cladding of 0.00023 inch, whichever is greater.
Based on Section 14.22.2 of the FSAR the former number is 5 times O.t % (i.e. 1.%),
and based on Attachment 14 to Reference 1 the later number is 0.84.% (less than 1.%).
Therefore, a value of 1.0 % is used in Sandia's analysi Radiolytic Hydrogen Yield:
It was not clear _,what the basis was for the value of *0.3 molecules/100 eV used in the licensee's COGAP analysi Based on Table 1 of RG 1,7, a value of 0.5 is used in Sandia's analysi~.
Containment Volume:
The value of 1.446*106 ft3 was used by the alleger because* it was a 11 eged that only the effective volume of the" spray shoul g be3 u~e As_discus~ed in Sect~on 2.5, the entire free volume of 1.64*10 ft is used in Sandia's analysi *
.
.
- i
. -6.:...
Zinc in paint:
In Attachment 1,5 to Reference 1, the licensee lumped the hydrogen ge.neration froni galvan.ized s.teel and from pain.t together. *The area and the corrosion rate of the galvanized steel were used. In Sandia 1s analysis, the* amount of zinc in paint specified in Table 14.22-4 of:the FSAR
- (11,500 lb with a *thickness of 0.003 inch) was use This may conservativ~ly generate a higher amount of hydrogen than either*the licensee 1 s results or.the alleger 1 s result The licensee indicated that the amount of zinc in paint might be only half of the FSAR valu However, the staff continued to use the FSAR valu *
-
Fission-Product Decay Energy Absorbed (FPDEA):* This is an important consideration because it affects the amount of water that is dissodated into H2 and o Th~ RC(9) *and RG(lO) designation was used in the COGAP manual (Reference 8).
RC(9) -.This is the percent FPDEA absorbed by the core wate The value *used by the licensee in Attachment 15 to Reference 1 was 7.0%.
Based on RG 1.7; a value of 10.0% was us~d in the Sandia analysi RC(lO) - This is the percent of total solid FPDEA absorbed by the sump wate The value used by the licensee was 12.0%~ The licensee explained that the 12% value.used by the alleger corresponded to severe accident condition Based on RG 1. 7, a value of 1:0% was used in the Sandi a 1 s analysis which corresponds to a design basis acciden *
.
.
.
3.1 Calculation Results Sandia performed COGAP calculations which incorporated all of the alleger's cbncerns evaluated in Sections 2.1 thr6ugh 2.7 and the pertinent changes to the licensee's COGAP calculation discussed in Section The maximum hydrogen concentration obtained from Sandia 1 s calculation was 2.9 v/o at 8 day. 0 CONCLUSION The staff determined that portions of the allegation have some meri The staff 1 s consultant at Sandia Laboratories performed an analysis to e\\::aluate the safety significance of the allegatio Based on the results of Sandia' analysis, the calculated hydrogen concentration following a LOCA at Palisades is 2.9 v/o, which does not exceed the 4.0 v/o flammable limit specified in RG Therefore, the impact of the allegations is not as.serious as the
- alleger describe However, the maximum hydrogen concentration from the original FSAR analysis was 2~0 v/ Therefore, the margin of safety documented in the FSAR has been redu-ce The licensee should revise the FSAR to reflect the new data, such as, amount of alumnium insulation, amount of zinc paint, new corrosion rates, recombiner initiation time, new method of analysis using COGAP code, input parameters, and the resulting hydrogen concentratio.0 REFERENCES Memorandum from H; J. MilJer to Dennis M. Crutchfield, dated September 20, 1990, Subject: Request for Assistance Allegations Regarding the Palisades Nuclear Plan *
-7-Attachment 15: licensee's ReView and Ahalysis Package o~ the Alleger's
_Hydrogen Generation Analysis, dated February 5, 1990.*
Atta.chment 17: Emergency Operating Procedures at Palisades Nuclear Plant, Revision 2, Dated July 20, 199 *
2. - Alleger's Hydr~gen Analysis File, SA-88-28, Revision 1, dated May 23~ 1989, Subject: H2 Generation Following.LOCA. -
.
.
.
.
.
.
. ORNL-3541, "Effect of Heat Flux on the Corrosion of Aluminum by Water," -
- o*y,
r
- r..-r-~---*
"
-,..
('~.. ~,..,,
~nA 1 I
i;nnli~h FP-bruarv -1964
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. NUREG/CR-3361, "The EffeCt of Water Chemistry on the Rates of Hydrogen
'Generation from Galvanized Steel Corrosion at Post-LOCA Conditions," by V. M. Loyola.and J.E. Womelsduff, December 198.
ORNL-TM-2412, "Design Consideration of Reactor Containment Spray System Part III, "The Corrosion of Materials in Spray Solutions," by J. t. Griess and A. L. Baccarella, December 196.
The Corrosion Handbook," by H. H. Uhlig, Eleventh Printing,' November 1969. - NUREG/CR-3808, "The Effect of,,post;..LQCA Conditions on a Protective Coating _ -_
(Paint) for the Nuclear Power Industry," by V. M. Loyola and J. E. Womelsduff,
- March 198 *
-
- NUREG/CR-2847, 11COGAP: A Nuclear Power Plant Containment Hydrogen Control System Evaluation Code," by R. G. Gido, January 198.
Memorandum from C. Y. Cheng to Conrad E. McCracken, dated December 27, 1990, Subject: Corrosion of Aluminum and Zinc in Palisades Plant. -
Principal Contributor~ C. Li
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555.
ENCLOSURE 4
- SAFETY EVALUATION BY THE OFFICE OF N.UCLEAR REACTOR REGULATION
.
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. RELATED TO AMENDMENT N0 *.. 136 TO PROVISIONAL OPERATING LICENSE NO. DPR-20 INTRODUCTION CHANGE NO. 1 CONSUMERS POWER COMPANY PALI SADES PLANT DOCKET NO. 50-255 By letter dated November 2, 1990, Consumers Power Company {the licensee)
requested amendment to the Technical Specifications (TSs} appended to
- Provisional Operati~g License No. DPR-20 for the Palisades Pl~nt. The proposed amendment would allow use of the Regulatory Guide 1.97 qualified
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neutron monitoring sy~tem which is being installed during the 1990 refueling outage. Additionally, a change was proposed to the Design Features section to more accurately describe the fixed absorber rod *
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CHANGE NO. 2 By letter dated June 13, 1990, and subsequently revised by letters dated November 9 and December-7, 1990, and January 24, 1991, the licensee requested
- an amendment to revise TS 3.3.1.b., "Emergency Core Cooling System." The *
proposed amendment would reduce the required minimum boron.solution level in
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the Safety Injection Tanks {SIT) from 186 to 174 inches. Additionally, the maximum allowed tank level would be expanded from 198 to 200 inches. This change effectively broadens the operating band at which SIT level must be maintained from 12 to 26 inche *
Two related TS changes were also submitted. First, a new surveillance requirement to check the SIT high and low level alarms was proposed to be included* in TS table 4.1.2. Secondly, the Bases section for lS 3.3~1 has been updated and two TS references have been adde.0. DISCUSSION CHANGE NO. 1 In 1988, the licensee performed a modification which upgraded the sensitivity of the fission chambers used to detect neutron flux. During the most ~ecent
- refueling outage, the l*icensee has completed its Opgrade of the neutron monitoring system ~Y certifying the system is qualified to the criteria of Reg.ulatory Guide 1.97. The changes in the neutron monitoring system performed this outage involved using the existing fission chambers, installing new cables from the fission chambers through two new electric penetrations to 9104090114 910403 PDR ADOCK 05000255 G
-2-preamplifiers (which were relocated to outside containment), and installing new cables from the preamplifiers to the power sources in the control room *.
Additionally, the neutron monitoring channel which supplies the alternate shutdown panel was modified. The alternate shutdown panel had previously.
received neutron monitoring indication from a dedicated, spare fission chambe The new system supplies the panel through an optical isolator associated with the left channel (NI-1/3, of the RG 1.97 qualified instrumen-tatio Eight channels of instrumentation are provided to monitor the neutron flux. The nuclear instrumentation system consists of two start-up channeis, two wide-range logarithmic channels and four power range safety channels. The start-up and wide-range channels share high sensitivity fission chambers while the power ranges channels are completely independent (each power range channels has a separate detector and power supply).
The rate-of-change of power is normally monitored at start-up by two source range monitors which sum inputs from two fission chambers and cover a range of approximately five decades (control room indication uses a scale from 1 to 3xl0e5 cps) *. The two other channels are wide-range units which take signals from fission.chambers and cover a range greater than te~ decades, overlapping the start-up channels by approximately three decades (control room indication uses a scale from lxlOe-8 to 200% power) *
The proposed Technical Specification changes, associated with the above modifications are as follows:
Changes. In the third paragraph from the end of the Basis for Section 3.17, delete reference to the "start-up" range and replace it with
"source" range. Delete reference to the "log" range and replace it with reference to the "wide" rang In Table 3.17.1, ihange Item No. 3 from "Log Range" to
"Wide-Range".
In Table 3.17.4, Item 7, change "Start-up" to *source Range" and in footnote (d) to Table 3.17.4,**change 11 log range" to "wide range".
In Table 3.25.1 Item No. 7, change *start-up" to *source" and
"(N-OOlA)" to "(N~l/3C)".
In Table 3.25.1, Item 14, add "Neutron Monitor System Power" under the Function Colum In Table 4.1.1, Item 2, under the Channel Description Column, delete the word "Logarithmic".
In Surveillance Functional Column, add " Calibrate". In the Frequency Column, add *"R". on the
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Calibrate" lin In the Surveillance Method Column on the 11a 11 line, delete 11both wide-range readings" and insert "Channel indications".
Also in the Surveillance Method column, add " Channel alignment through measurement/adjustment of internal test points".
- In Table 4.1. 3, *Item No. 1 in the _Channel Description Col Limn change 11Start-up 11 to "Source"; in the Surveillance Functions Column add "c. Calibrate"; in the Frequency Column,.add 11 R 11 o the 11c. Calibrate 11 line; and, in the Surveillance Method Column, add 11 Channel alignment through measurement/adjustment of internal test points~*
In Table 4.21.1, item No. 7, under the Channel Description column, change 11Start-up 11 to 11Source 11 and 11(NI-OOlA)
11 to 11(NI-1/3C)
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In the surveillance Method Column, after " Internal Test Signal 11 add, "(Performed under Table 4.1.3, Item 1. b)".
In Section 5.3.2d, change the description rif the mechanically fixed rods from 11 *** mechanically fixed boron rods...
to
~ ** mechanically fixed absorbers rods... ".
CHANGE NO. 2 The four safety injection tanks are part of the safety injection system and are used to flood the core with borated water following a depressurization of the primary coolant syste Three of the four tanks will provide sufficient coolant to recover the core following a Loss of Coolant Acciden The tanks are connected to the Primary Coolant System cold legs through normally open isolation valves.. 'Two check valves prevent primary coolant from entering the tank Current TS maintain the tanks pressurized to at least 200 psig, with a tank Hquid level of at least 186 inches and a maximum level of 198 inches*, and a boron concentration from 1720 to 2000 pp Injection will occur whenever the primary system pressure fails below the combined pressure of the static wate head plus the tank gas pressur The licensee proposed the following changes to the TS; Changes Change Specification 3.3.1.b to read as follows:
"All four Safety Injection Tanks are operable and pressurized to at least 200 psig with a tank liquid level of at least 174 inches and maximum level of 200 inches with a boron concentration of at least 1720 ppm but not more than 2000 ppm.
Change the fourth paragraph of Section 3.3 Basis as follows:
"The limits for the Safety Injection Tank pressure and volume assure the required amount of water injection during an accident and are based on values used for the accident analyses (3, 4).
The minimum 174-inch level corresponds to a volume of 1040 ft3 and the maximum 200-inch level corresponds to a volume of 1176 ft3 *
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-4-Add the following to References:
- (3). FSAR, Section 14.17
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(4) Letter, H. G. Shaw {ANF) to R. J. Gerling (CPCo), *standard Review Plan Chapter 15, Disposition of Events Review for Changes to Technical Specification L 1mits on-Palisades Safety Injection Tank Liquid Levels", April 11, 199 Add Surveillance Function ~c.* To Item 13 on Table 4.1.2 to require performance, at least once per 1~ months, of a functional check on the SiT high and iow ievei aiann *
These changes are considered necessary to reduce the risk of TS violations made possible by periodic surveillante and correctiori of boron concentratio When sampling the SITs to verify boron concentration, it is* necessary to drain the tanks sufficiently to obtain an accurate sample. During this evolution, TS.
Section 3.3.2.a is in effect and limits the non-operability of one tank to one hou Because a significant amount of water must be drained from the tank to obtain a representative sample, the possibility exists that proper level may not be restored within the one hour period. This procedure places demands on the Operations staff which would be minimized if the operating band of the tanks were broadene.0 EVALUATION CHANGE No. 1 The changes to the neutron monitoring system enhance the reliability of the accuracy of the. neutron monitorfog function under accident conditions. The previous system has basically been upgraded to the more stringent requirements of Regulatory Guide 1.97. The upgraded equipment performs the same function
- as the previously installed equipment, and maintains the same degree of
redundanc For consistency with Standard Technical Specifications, the licensee has
- proposed to change the name of the most sensitive neutron monitoring range from "Start-up Range" to "Source-Range". Similarly, the name of the
"Logarithmic Range" has been proposed to be renamed "Wide-Range". These designations are more consistent with standard industry phraseology, and more clearly describe the respective neutron monitoring ranges. Changes A,B,C,D,F,G and H reflect the new termfoolog Changes D and H also correct the designation of the source range neutron monitor which provides indication for alternate shutdown capability {N-OOlA is changed to NI-l/3C). The source of the signal to the alternate shutdown panel source range monitor has been modified. The previous signal (N-OOlA)
originated from an older, dedicated fission chamber. The new signal {NI-1/JC)
1s supplied through an optical isolator associated with the left channel of the new, RG 1.97 qualified instrumentation. The newer system provides for more accurate indication and improved reliabilit **
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. Change E prov;des add;tfonal foformat;on to the TS, to clearly delineate the neutron monitoring system power source. Changes F, G, and H either add additfonal surveillance$ to the TS, or clarify existing surveillance Change I ;s necessary to correct the TS description of the core's mechanically fixed rods. Fixed boron rods are no longer in use at Palisades. A conversion from boron to Gadolinia rods was complet~d in core reloads H, I and J. *Also, with recent fuel cycles converting to reduced leakage designs, additional neutron absorbent rods are being utilized (e.g., stainless steel~ Hafnium}.
Because a variety of materials may be used, the term Rfixed absorber rods" is considered appropriate ~o accommodate different core reload design In surrmary, the proposed TS changes more clearly describe the function and operation of the neutron monitoring system. Additionally, the changed description of the type of neutron absorbing-rod in use at the Palisades Plant corr~cts an error in the Design Features TS section, which was ov~rlooked at the time of the Core reload H submittal. These changes reflect the use of material or equipment which w*ill perform the same functions as existing equipment, and are consideredacceptabl CHANGE NO. 2 Discussions were held with the licensee's Operattons staff to assess the operational restrictions imposed by the current SIT level band. Operational data for the period from 11/88 to 9/90 indicates that the.SITS were sampled 187 time SIT sampling is required monthly by the T The increased sampling frequency {approximately 100 samples above the TS minimum} was required due to known 1nleakage into the SITs from the primary coolant system._ This inleakage slowly lowered SIT boron concentration; therefore,.
additional monitorini of tank concentration was required in order to ensure the minimum TS levels for SIT boron concentration were maintaine The SITs do not have tank recirculation capabilitYi therefore, a relatively large amount of tank water is required by procedure to be purged through the sample lines in order to obtain a representative sample (on the order of 1700 gallons-per sample). This water volume necessitates that the TS Lim;ting Condition for Operation be entered each time a tank is sample Although the most appropriate deterrent in-preventing unnecessary SIT sampling is ensuring the leak-tightness of the SIT boundary valves, widening the operational water level band will assist in maintairiing the tanks within their prescribed limit The licensee contracted with their fuel vendor to evaluate the effect of a slightly reduced or increase~ total liquid volume in the SITs (174 and 202 inches, respectively,, for minimum and maximum allowed SIT levels). The result of the fuel vendor's StGndard Review Plan Chapter 15 disposition of events was that the large break loss of coolant transient is the only event which completely drains the SITs (thereby it is the only event which could be affected by the changes in SIT level limits}.
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-6-Data supplied by the licensee indicates that flow from the intact loop SITs,
. SIT lines, and cold legs keeps the downcomer full for about 30 seconds after the peak cladding temperature (PCT) for the transient is reached. Reduction of the minimum SIT level to 174 inches causes the SITS to empty, in the
- worst case, approximately four seconds earlier than would h~ve oceurred with the previous tank limits. Downcomer level does not fall prior to the time the PCT is reached. Additionally, increasing the maximum SIT level to 202 inches has no impact on the large break LOCA analysis because the SIT flow time would be conservatively extended beyond the time in the limiting analysis. These conclusi~ns apply to all break sizes contained in the February 1990 Paiisades Large Break LOCA A~a1ysis of recor The following factors were also considered:
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The change in the upper SIT limit involves a relatively small increase in the maximum amount of water stored in the SITs (two inches of level);
therefore, the probability of overfilling, containment flooding*, and malfunctions due to seismic events are not significantly increase The LOCA containment analysis, which conservatively does not take credit for SIT injection, shows that peak containment pressure stays below the design pressur A comparison of the SIT operating band volumes in use at several Combustion Engineering plants indicates similar volumes. -Also, the Combustion Engineering Standard TS provides for an operating band of roughly the same volume as that proposed by Palisades (126 and 136 cubic feet, respectively).
The effect of the reduced minimum SIT inventory on the available suction source for Safety Injection during long term recirculation from the containment sump has been considered. The reduction (1885 gal) in required minimum inventory is a very small fraction of the total available inventory (approx. 380,000 gal) considering the vast inventory contribut1on from the four SITs and the Safety Injection *
Refueling Water Tank; and, therefore, will have a negligible effect on the operation of *the safety injection pumps or the containment temperature and pressure respons *The boron concentration in the tanks will be unchanged and the slight reduction in total inventory will not have a significant effect on the sump boron concentration during the recirculation phase of the accident. Additionally, this change will not significantly effect the time before hot leg injection is required to prevent precipitation of bo.ro In su11111ary, the proposed changes to the Technical Specification limits on Palisades SIT levels have been evaluated to ensure that adequate water is available for make-up to the primary coolant system. The analysis shows that when the contents of the SITs are at the proposed lower level, and a large break LOCA occurs, the SITs do not empty until after the peak cladding temperature is reached and until after high and low pressure safety injection
-7-are actuated. The addition of a. surveillance requirement to perform a
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functional check on the SIT high and low level alarms institutes additional TS controls on ensuring that SIT level will be adequately ineasured and maintained. Additionally, the basis section has been updated and the TS*
Reference section appropriately expanded. Therefore, these propcised TS changes are considered acceptabl.0 ENVIRONMENTAL CONSIDERATION This amendment involves a change in a requirement with-respect to the
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area as defined in 10 CFR Part 20 and a change in a surveillance requiremen The staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types,_ of any effluents that*
may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Conlnission has previously issued a proposed finding that this amendment involves no
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significant hazards consideration and there has been no public contnent on such finding. Accordingly, this amendment meets the eligibility criteria for categorical exclusion set fo~th in 10 CFR Section 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental ~impact statement or environmental assessment need be prepared in connection with the issuance of this amendmen. 0 CONCLUSION The staff has concluded, based on the considerations discussed above, that (1) there is reasonable assurance that the hea-lth and safety of the p~blic wi 11 not be enda_ngered by operation in the proposed manner, and ( 2) such activities will be conducted in compliance with the Co1TDTJission 1s regulations, and the issuance of the amendment will not be inimical to the conmon defense and security or to the health and safety of the public. The staff therefore concludes that the proposed changes are acceptabl Principal Contributor: Brian Holian Date: february 15, 1991